[Federal Register Volume 72, Number 166 (Tuesday, August 28, 2007)]
[Rules and Regulations]
[Pages 49352-49566]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 07-3861]
[[Page 49351]]
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Part II
Nuclear Regulatory Commission
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10 CFR Parts 1, 2, 10, et al.
Licenses, Certifications, and Approvals for Nuclear Power Plants; Final
Rule
Federal Register / Vol. 72, No. 166 / Tuesday, August 28, 2007 /
Rules and Regulations
[[Page 49352]]
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NUCLEAR REGULATORY COMMISSION
10 CFR Parts 1, 2, 10, 19, 20, 21, 25, 26, 50, 51, 52, 54, 55, 72,
73, 75, 95, 140, 170, and 171
RIN 3150-AG24
Licenses, Certifications, and Approvals for Nuclear Power Plants
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC) is amending its
regulations by revising the provisions applicable to the licensing and
approval processes for nuclear power plants (i.e., early site permit,
standard design approval, standard design certification, combined
license, and manufacturing license). These amendments clarify the
applicability of various requirements to each of the licensing
processes by making necessary conforming amendments throughout the
NRC's regulations to enhance the NRC's regulatory effectiveness and
efficiency in implementing its licensing and approval processes. The
NRC has considered and resolved the public comments.
DATES: The effective date is September 27, 2007.
FOR FURTHER INFORMATION CONTACT: Nanette V. Gilles, Office of New
Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, telephone 301-415-1180, e-mail [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
A. Development of Proposed Rule
B. Publication of Revised Proposed Rule
II. Overview of Public Comments
III. Reorganization of Part 52 and Conforming Changes in the NRC's
Regulations
IV. Responses to Specific Requests for Comments
V. Discussion of Substantive Changes and Responses to Significant
Comments
A. Introduction
B. Testing Requirements for Advanced Reactors
C. Changes to 10 CFR Part 52
D. Changes to 10 CFR Part 50
E. Change to 10 CFR Part 1
F. Changes to 10 CFR Part 2
G. Changes to 10 CFR Part 10
H. Changes to 10 CFR Part 19
I. Changes to 10 CFR Part 20
J. Changes to 10 CFR Part 21
K. Change to 10 CFR Part 25
L. Changes to 10 CFR Part 26
M. Changes to 10 CFR Part 51
N. Changes to 10 CFR Part 54
O. Changes to 10 CFR Part 55
P. Changes to 10 CFR Part 72
Q. Changes to 10 CFR Part 73
R. Change to 10 CFR Part 75
S. Changes to 10 CFR Part 95
T. Changes to 10 CFR Part 140
U. Changes to 10 CFR Part 170
V. Changes to 10 CFR Part 171
VI. Section-by-Section Analysis
VII. Availability of Documents
VIII. Agreement State Compatibility
IX. Voluntary Consensus Standards
X. Environmental Impact--Categorical Exclusion
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Certification
XIV. Backfit Analysis
XV. Congressional Review Act
I. Background
A. Development of Proposed Rule
On July 3, 2003 (68 FR 40026), the NRC published a proposed
rulemaking that would clarify and/or correct miscellaneous parts of the
NRC's regulations; update 10 CFR part 52 in its entirety; and
incorporate stakeholder comments. On March 13, 2006 (71 FR 12781), the
NRC issued a revised proposed rule that would rewrite part 52, make
changes throughout the Commission's regulations to ensure that all
licensing processes in part 52 are addressed, and clarify the
applicability of various requirements to each of the processes in part
52 (i.e., early site permit, standard design approval, standard design
certification, combined license, and manufacturing license). This
proposed rule superseded the July 3, 2003, proposed rule.
The NRC issued 10 CFR part 52 on April 18, 1989 (54 FR 15372), to
reform the NRC's licensing process for future nuclear power plants. The
rule added alternative licensing processes in 10 CFR part 52 for early
site permits, standard design certifications, and combined licenses.
These were additions to the two-step licensing process that already
existed in 10 CFR part 50. The processes in 10 CFR part 52 allow for
resolving safety and environmental issues early in licensing
proceedings and were intended to enhance the safety and reliability of
nuclear power plants through standardization. Subsequently, the NRC
certified four nuclear power plant designs under subpart B of 10 CFR
part 52--the U.S. Advanced Boiling Water Reactor (ABWR) (62 FR 25800;
May 12, 1997), the System 80+ (62 FR 27840; May 21, 1997), the AP600
(64 FR 72002; December 23, 1999), and the AP1000 (71 FR 4464; January
27, 2006). These design certifications are codified in appendices A, B,
C, and D of 10 CFR part 52, respectively.
The NRC planned to update 10 CFR part 52 after using the standard
design certification process. The proposed rulemaking action began with
the issuance of SECY-98-282, ``Part 52 Rulemaking Plan,'' on December
4, 1998. The Commission issued a staff requirements memorandum (SRM) on
January 14, 1999 (SRM on SECY-98-282), approving the NRC staff's plan
for revising 10 CFR part 52. Subsequently, the NRC obtained
considerable stakeholder comment on its planned action, conducted three
public meetings on the proposed rulemaking, and twice posted draft rule
language on the NRC's rulemaking Web site before issuance of the July
2003 proposed rule. \
B. Publication of Revised Proposed Rule
A number of factors led the NRC to question whether the July 2003
proposed rule would meet the NRC's objective of improving the
effectiveness of its processes for licensing future nuclear power
plants. First, public comments identified several concerns about
whether the proposed rule adequately addressed the relationship between
part 50 and part 52, and whether it clearly specified the applicable
regulatory requirements for each of the licensing and approval
processes in part 52. In addition, as a result of the NRC staff's
review of the first three early site permit applications, the staff
gained additional insights into the early site permit process. The NRC
also had the benefit of public meetings with external stakeholders on
NRC staff guidance for the early site permit and combined license
processes. As a result, the NRC decided that a substantial rewrite and
expansion of the July 2003 proposed rulemaking was desirable so that
the agency may more effectively and efficiently implement the licensing
and approval processes for future nuclear power plants under part 52.
Accordingly, the Commission decided to revise the July 2003
proposed rule and published a revised proposed rule for public comment
on March 13, 2006. This revised proposed rule contained a rewrite of
part 52, as well as changes throughout the NRC's regulations, to ensure
that all licensing and approval processes in part 52 are addressed, and
to clarify the applicability of various requirements to each of the
processes in part 52. In light of the substantial rewrite of the July
2003 proposed rule, the expansion of the scope of the rulemaking, and
the NRC's decision to publish the revised proposed rule for public
comment, the NRC decided that developing responses to comments received
on the July 2003 proposed rule would not be an effective use of agency
resources. The NRC requested that commenters on the July 2003 proposed
rule who believed that their earlier
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comments were not adequately addressed in the March 2006 proposed rule
resubmit their comments.
II. Overview of Public Comments
The public comment period for the March 2006 revised proposed rule
expired on May 30, 2006. The NRC received 19 comment letters from
industry stakeholders, other Federal agencies, and individuals during
the public comment period. The NRC has considered and resolved all of
the public comments received during the comment period and has made
modifications to the rule language, as appropriate. The NRC has
prepared a separate report, entitled Comment Summary Report: 10 CFR
Part 52, Licenses, Certifications, and Approvals for Nuclear Power
Plants, in which it summarizes the public comments received and
discusses the agency's disposition of each comment. This report is
available to the public as discussed in Section VII of the
Supplementary Information of this document. The resolution of
significant public comments is also discussed in Section IV, Responses
to Specific Requests for Comments and, Section V, Discussion of
Substantive Changes and Responses to Significant Comments in this
document.
III. Reorganization of Part 52 and Conforming Changes in the NRC's
Regulations
Since the adoption of 10 CFR part 52 in 1989, the NRC and its
external stakeholders identified a number of interrelated issues and
concerns with the licensing process. One significant concern was that
the overall regulatory relationship between part 50 and part 52 was not
always clear. In the former rules, it was often difficult to tell
whether general regulatory provisions in part 50 apply to part 52. One
example is whether the absence of an exemption provision in part 52
denotes the NRC's determination that exemptions from part 52
requirements are not available, or that these exemptions are controlled
by Sec. 50.12. A related problem is the current lack of specific
delineation of the applicability of NRC requirements throughout 10 CFR
Chapter I to the licensing and approval processes in part 52. For
example, the indemnity and insurance provisions in part 140 were not
revised to address their applicability to applicants for and holders of
combined licenses under subpart C of part 52. Even where part 52
provisions referenced specific requirements in part 50, it was not
always clear from the language of the part 50 requirement how that
requirement applied to the part 52 processes. For example, Sec.
52.47(a)(1)(i) provides that a standard design certification
application must contain the ``technical information which is required
of applicants for construction permits and operating licenses by 10
CFR* * *part 50* * *and which is technically relevant to the design and
not site-specific.''
The language did not explicitly identify the part 50 requirements
that are ``technically relevant to the design.'' Even where a specific
regulation in part 50 is identified as a requirement, the language of
the referenced regulation itself was not changed to reflect the
specific requirements as applied to the part 52 processes. For example,
Sec. 52.79(b) provides that the application must contain the
``technically relevant information required of applicants for an
operating license required by 10 CFR 50.34.'' Other than the fact that
this language shares the problem discussed earlier of what constitutes
a ``technically relevant'' requirement, Sec. 50.34(b) is based upon
the two-step licensing process whereby certain important information is
submitted at the construction permit stage, and then supplemented with
more detailed information at the operating license stage. Thus, it
could be asserted that certain information that must be submitted in
the construction permit application, e.g., the ``principal design
criteria for the facility'' required by Sec. 50.34(a)(3)(i), may be
regarded as not required to be submitted for a combined license
application under the former version of part 52.
Another potential source of confusion is that the different
subparts of part 52 and the appendices on standard design approvals and
manufacturing licenses are not organized using the same format of
individual sections (e.g., ``Scope of subpart,'' followed by
``Relationship to other subparts,'' followed by ``Filing of
application''). Moreover, the organization and textual content of
identically-titled sections differs among the subparts, and with
appendices M, N, O, and Q, which establish additional licensing and
approval processes. While these differences do not constitute an
insurmountable problem to their use and application, it became apparent
to the Commission that adoption of a common format, organization, and
textual content would enhance usability and result in increased
regulatory effectiveness and efficiency.
In the 2003 proposed rule, the NRC proposed several changes that
were intended to address some (but not all) of these issues. However,
based upon comments received on the 2003 proposed rule, the NRC's
experience to date with early site permit applications, interactions
with external stakeholders concerning NRC guidance for combined license
applications, and NRC's screening of 10 CFR Chapter I requirements
following the receipt of public comments on the 2003 proposed rule, the
NRC concluded that the 2003 proposed rule would not adequately address
and resolve these issues.
Accordingly, in the March 13, 2006, proposed rule the NRC took a
more comprehensive approach to addressing these issues by reorganizing
part 52, implementing a uniform format and content for each of the
subparts in part 52, using consistent wording and organization of
sections in each of the subparts, and making conforming changes
throughout 10 CFR Chapter I to reflect the licensing and approval
processes in part 52. The NRC also coordinated and reconciled
differences in wording among provisions in parts 2, 50, 51, and 52 to
provide consistent terminology throughout all of the regulations
affecting part 52. Under the NRC's reorganization of part 52, the
existing appendices O and M on standard design approvals and
manufacturing licenses, respectively, have been redesignated as new
subparts in part 52. Redesignating these appendices as subparts in part
52 has resulted in a consistent format and organization of the
requirements applicable to each of the licensing and approval
processes. In addition, the redesignation clarifies that each of the
licensing and approval processes in these appendices are available to
potential applicants as an alternative to the processes in part 50
(construction permit and operating license) and the existing subparts A
through C of part 52. The Commission does not, by virtue of this
redesignation, either favor or disfavor the processes in the former
appendices M and O of part 52. Rather, the Commission is standardizing
the format and organization of part 52, and clarifying the full range
of alternatives that are available under part 52 for use by potential
applicants. Consistent with the broad scope of part 52, the NRC has
retitled 10 CFR part 52 as ``Licenses, Certifications, and Approvals
for Nuclear Power Plants.''
The NRC has also reorganized and expanded the scope of the
administrative and general regulatory provisions that precede the part
52 subparts by adding new sections on written communications (analogous
to Sec. 50.4), employee protection (analogous to Sec. 50.7),
completeness and accuracy of information (analogous to Sec. 50.9),
exemptions (analogous to Sec. 50.12), combining licenses (analogous to
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Sec. 50.52), jurisdictional limits (analogous to Sec. 50.53), and
attacks and destructive acts (analogous to Sec. 50.13). The NRC
believes that adding the new sections to part 52 rather than revising
the comparable sections in part 50 is more consistent with the general
format and content of the Commission's regulations in each of the parts
of Title 10. The NRC considered whether the numbering of the newly-
added sections to part 52--in particular, the provisions on deliberate
misconduct, employee protection, and completeness and accuracy of
information--should match the numbering of the comparable sections in
part 50. While this may have some benefit, the NRC ultimately decided
not to adopt such a course for several reasons. First, other parts of
the NRC's regulations in 10 CFR Chapter I do not maintain the same
numbering scheme. Rather, it appears that the NRC attempted to maintain
the order in which these sections are listed in each part. Second,
there are other provisions in part 50 for which a comparable provision
needed to be added to the general and administrative provisions in part
52, but for which it would be impossible to maintain the same numbering
(for example, Sec. 50.13 (attacks and destructive acts); Sec. 50.32
(elimination of repetition); Sec. 50.52 (combining licenses)), unless
the substantive provisions of part 52, beginning with Sec. 52.12, were
changed.\1\ Maintaining in part 52 the numbering scheme for some, but
not all, comparable sections from part 50 ultimately would be viewed as
haphazard and arbitrary. Finally, the NRC does not believe that
external stakeholders who must constantly refer to part 52 will be
confused by any difference in numbering of the three sections, given
that there are other comparable provisions for which the numbering is
necessarily different between parts 50 and 52. For these reasons, the
NRC did not attempt to match in the final part 52 rule the numbering of
the comparable sections in part 50.
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\1\ The NRC notes, in this regard that nuclear industry
stakeholders adversely commented on the revised numbering scheme as
set forth in the 2003 proposed part 52 rule. They suggested that the
NRC retain, to the greatest extent posible, the numbering of the
then existing part 52. Inasmuch as Sec. 52.12 is the first
substantive provision of the former party 52, this placed an upper
bound on the number of sections available for general provisions--
that is Sec. 52.0 through 52.11.
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Appendix N, which addresses duplicate design licenses, has been
retained in both part 52 and part 50 to afford future applicants
flexibility and to retain the possibility of achieving regulatory
efficiencies in part 52 combined license proceedings. Since the
preparation of the March 2006 proposed rule, several industry groups
have announced their intention to seek combined licenses utilizing the
same design. In view of this industry development, the NRC believes
that there is potential utility to keeping the option of appendix N
open to potential combined license applicants. Accordingly, the NRC is
retaining in part 52 the procedural alternative provided in appendix N,
and revising its language to make its provisions applicable to combined
licenses using identical designs. Appendix Q, which addresses early
staff review of site suitability issues, is being removed from part 52
but retained in part 50. Appendix Q provides for NRC staff issuance of
a staff site report on site suitability issues with respect to a
specific site for which a potential applicant seeks the NRC staff's
views. The staff site report is issued after receiving and considering
the comments of Federal, State, and local agencies and interested
persons, as well as the views of the Advisory Committee on Reactor
Safeguards (ACRS), but only if site safety issues are raised. The staff
site report does not bind the Commission or a presiding officer in any
hearing under part 2. This process is separate from the early site
permit process in subpart A of part 52. The NRC recognizes the apparent
redundancy between the early review of site suitability issues and the
early site permit process. Accordingly, the NRC is removing appendix Q
from part 52 and retaining it only in part 50.
Inasmuch as the NRC may, in the future, adopt other regulatory
processes for nuclear power plants, the NRC has reserved several
subparts in part 52 to accommodate additional licensing processes that
may be adopted by the NRC. The NRC used a standard format and content
for revising the regulations in the existing subparts and developing
the new subparts that address the former appendices M and O. The
standard format and content was modeled on the existing organization
and content of subparts A and C. Appendix N of part 52, however, has
not been revised in that fashion because of time constraints in
developing the final rule.
Perhaps most importantly, the NRC has reviewed the existing
regulations in 10 CFR Chapter I to determine if the existing
regulations must be modified to reflect the licensing and approval
processes in part 52. First, the NRC determined whether an existing
regulatory provision must, by virtue of a statutory requirement or
regulatory necessity, be extended to address a part 52 process, and, if
so, how the regulatory provision should apply. Second, in situations
where the NRC has some discretion, the NRC determined whether there
were policy or regulatory reasons to extend the existing regulations to
each of the part 52 processes. Most of the conforming changes in this
final rule occur in 10 CFR part 50. In making conforming changes
involving 10 CFR part 50 provisions, the NRC has adopted the general
principle of keeping the technical requirements in 10 CFR part 50 and
maintaining all applicable procedural requirements in part 52. However,
due to the complexity of some provisions in 10 CFR part 50 (e.g., Sec.
50.34), this principle could not be universally followed. A description
of, and bases for, the substantive conforming changes for each affected
part is provided in Section V of this document.
To highlight the relationship between the requirements in part 52
of this final rule and the requirements in existing part 52, the NRC is
making two cross-reference tables available to the public. These tables
can be found on NRC's Agencywide Documents Access and Management System
(ADAMS) at accession number ML062550U0246. Table 1 matches each part 52
requirement in this final rule with its counterpart in the existing
rule. Table 2 is a reverse cross-reference table which identifies the
section of the existing part 52 requirements from which each part 52
requirement in this final rule was derived.
IV. Responses to Specific Requests for Comments
In Section V of the Statements of Consideration for the March 13,
2006, proposed rule, the NRC posed 15 questions for which it solicited
stakeholder comments. In the following paragraphs, these questions are
restated, comments received from stakeholders are summarized, and the
NRC resolution of the public comments is presented.
Question 1: General Provisions. Create new subpart for part 50. In
response to several commenters' concerns about the clarity of the
applicability of part 50 provisions to part 52, the Commission has
added provisions to part 52 (Sec. Sec. 52.0 through 52.11) that are
analogues to comparable provisions in part 50. Another possible way of
addressing the commenters' concerns would be to transfer all the
provisions in part 52 to a new subpart (e.g., subpart M) of part 50,
and retain the existing numbering sequence for the current part 52 with
the addition of a prefix (e.g., proposed
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50.1001 = current 52.1). The Commission is considering adopting this
alternative proposal in the final rule and is interested in whether
stakeholders regard this as a more desirable approach for minimizing
the ambiguity of the relationship between part 50 and part 52.
Commenters' Response: Some commenters stated the clarity of the
regulations would not be enhanced by moving provisions from part 52 to
a new subpart of part 50. The commenters argued that in addition to not
eliminating existing confusion, such a content shift would create new
confusion because current documents referencing part 52 would become
``obsolete.''
NRC Response: The NRC has decided not to transfer provisions from
part 52 to a new subpart in part 50, inasmuch as: (1) no commenter
favored transferring provisions from part 52 to a new subpart in part
50, (2) the approaches are legally equivalent, and (3) nearly 17 years
has passed since the Commission adopted the approach of establishing
early site permits, standard design certifications, and combined
licenses in a new part 52, and a reorganization of the regulations at
this time may engender confusion without any compensating benefits in
clarity, regulatory stability and predictability, or efficiency.
Question 2: Currently, Sec. Sec. 52.17(b) of subpart A of 10 CFR
part 52 requires that an early site permit application identify
physical characteristics that could pose a significant impediment to
the development of emergency plans. An early site permit application
may also propose major features of the emergency plans or propose
complete and integrated emergency plans in accordance with the
applicable standards of Sec. 50.47 and the requirements of appendix E
of 10 CFR part 50. The requirements in Sec. 52.17 do not further
define major features of emergency plans. Section 52.18 of subpart A
requires the Commission to determine, after consultation with the
Federal Emergency Management Agency, whether any major features of
emergency plans submitted by the applicant under Sec. 52.17(b) are
acceptable. Section 52.18 does not provide any further explanation of
the Commission's criteria for judging the acceptability of major
features of emergency plans.
The Commission has concluded, after undergoing the review of the
first three early site permit applications, that Commission review and
acceptance of major features of emergency plans may not achieve the
same level of finality for emergency preparedness issues at the early
site permit stage as that associated with a reasonable assurance
finding of complete and integrated plans. Therefore, the Commission is
considering modifying in the final rule the early site permit process
in proposed subpart A to remove the option for applicants to propose
major features of emergency plans in early site permit applications and
requests public comment on this alternative. The NRC believes that, if
the option for early site permit applicants to include major features
of emergency plans is to be retained, it would be useful to further
define in the final rule what a major feature is and establish a
clearer level of finality associated with the NRC's review and
acceptance of major features of emergency plans. If the option to
include major features of emergency plans is retained in the final
rule, the NRC would define major features of emergency plans as
follows:
Major features of the emergency plans means the aspects of those
plans necessary to: (1) address one or more of the sixteen standards
in Sec. 50.47(b), and (2) describe the emergency planning zones as
required in Sec. Sec. 50.33(g), 50.47(c)(2), and appendix E to 10
CFR part 50.
In addition, the NRC is considering adopting in the final rule the
requirement that major features of emergency plans must include the
proposed inspections, tests, and analyses that the holder of a combined
license referencing the early site permit shall perform, and the
acceptance criteria that are necessary and sufficient to provide
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met, the facility has been
constructed and will operate in conformity with the license, the
provisions of the Atomic Energy Act (AEA), and the NRC's regulations,
insofar as they relate to the major features under review.
The NRC believes that, under this alternative, the level of
finality associated with each major feature that the Commission found
acceptable would be equivalent, for that individual major feature, to
the level of finality associated with a reasonable assurance finding by
the NRC for a complete and integrated plan, including inspections,
tests, analyses, and acceptance criteria (ITAAC), at the early site
permit stage.
Commenters' Response: Several commenters suggested the current
process for addressing major features of emergency plans (EP) in the
early site permit (ESP) be retained without modification. Some
commenters expressed a fear that the loss of this option would result
in a loss of flexibility to achieve ``finality'' without producing a
comprehensive EP. Some commenters identified a need to clarify the
definition of ``major features'' of the EP to make it less restrictive.
Some commenters believed that the approved major features were
acceptable elements of a ``complete and integrated emergency plan that
would be considered later.'' Some commenters believed the information
should not be reviewed again during the COL process, which would
instead focus on (1) the integration of these major features with
information necessary to support the ``reasonable assurance finding,''
and (2) the updating of EP information required by Sec. 52.39(b).
NRC Response: Based on the commenters' feedback, the NRC has
decided to retain the current process for addressing major features of
emergency plans in an ESP without modification. The NRC agrees that it
should clarify the definition of ``major features'' and has done so by
adding the definition suggested by the commenters to Sec. 52.1 in the
final rule. For a detailed discussion of the basis for this change, see
Section V.C.5.b of the Supplementary Information section of this
document which discusses changes to Sec. 52.1, ``Definitions.''
Question 3: As indicated in Section IV, Discussion of Substantive
Changes (in the March 13, 2006, proposed rule), the NRC is proposing to
remove appendix Q to part 52 entirely from part 52 and retain it in
part 50. Currently, appendix Q to part 52 provides for NRC staff
issuance of a staff site report on site suitability issues with respect
to a specific site, for which a person (most likely a potential
applicant for a construction permit or combined license) seeks the NRC
staff's views. The NRC is also considering removing, in the final rule,
the early site review process in appendix Q to part 52 in its entirety
from the NRC's regulations and is interested in stakeholder feedback on
this alternative. One possible reason for removing the early site
review process in its entirety is that potential nuclear power plant
applicants would use the early site permit process in subpart A of part
52, rather than the early site review process as it currently exists in
appendix Q to parts 50 and 52. Also, in cases where a combined license
applicant was interested in seeking NRC staff review of selected site
suitability issues (as appendix Q to part 52 was designed for), the
applicant could request a pre-application review of these issues. The
use of pre-application reviews for selected issues has been
successfully used by applicants for design certification. The NRC is
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especially interested in the views of potential applicants for nuclear
power plant construction permits and combined licenses as to whether
there is any value in retaining the early site review process.
Commenters' Response: Some commenters expressed concern about the
loss of flexibility to assess site suitability that would result from
the deletion of appendix Q from parts 50 and 52. These commenters
believed that appendix Q to parts 50 and 52 (in conjunction with
subpart F of 10 CFR part 2) was important for allowing ``critical path
issues'' to be reviewed prior to submission of a combined license (COL)
application in instances where prior completion of an ESP was not
feasible. Some commenters argued for the efficiency of appendix Q to
parts 50 and 52 and subpart F of part 2 because only applicant-selected
issues would be reviewed during these processes. Some commenters
recommended changes be made to specifically allow ESP and COL
applicants to reference an early site review conducted in accordance
with appendix Q or subpart F. The commenters stated that the NRC should
not delete the option for a part 52 applicant to reference a review
performed under appendix Q to 10 CFR part 52.
NRC Response: After considering these comments the NRC has decided
to go forward with removal of appendix Q from part 52 in the final
rule.
However, the NRC agrees that Sec. 2.101(a-1) and subpart F of part
2 should be modified to allow applicants for early site permits and
combined licenses under part 52 to take advantage of those provisions.
Both Sec. 2.101(a-1) and subpart F of part 2 have been revised in the
final rule, albeit somewhat differently than the approach recommended
by the commenter. Inasmuch as the revisions are to the Commission's
rules of procedure and practice, the Commission may adopt them in final
form without further notice and comment, under the rulemaking
provisions of the APA, 5 U.S.C. 553(b)(A). The Commission believes that
sufficient flexibility will be retained for future combined license
applicants with the preservation of the provisions in Sec. 2.101(a-1)
and subpart F of part 2 and that there is little value in also
retaining the provisions in appendix Q.
Question 4: Under subpart F of part 52 of the proposed rule, the
NRC proposes to require approval of, and extend finality to, the final
design for a reactor to be manufactured under a manufacturing license.
While the NRC will also review the acceptability of the manufacturing
license applicant's organization responsible for design and
manufacturing, as well as the quality assurance (QA) program for design
and manufacturing, the proposed rule does not provide a regulatory
structure for further extending the scope of NRC review and issue
finality to the manufacturing process itself. The NRC is considering
extending regulatory review approval, and consequently expand issue
finality, to the manufacturing itself in the final rule. There are two
models that the Commission is considering adopting if it were to move
in this direction. The first would be an analogue to the subpart C of
part 52 combined license process, whereby the NRC would review and
approve manufacturing ITAAC to be included in the manufacturing
license. During the manufacturing of each reactor, the NRC would verify
at the manufacturing location whether the ITAAC have been conducted and
the acceptance criteria met. A NRC finding of successful completion of
all the ITAAC would preclude any further inspection of the
acceptability of the manufacture of the reactor at the site where the
manufactured reactor is to be permanently sited and operated. The NRC's
inspections and findings for the combined license or operating license
would be limited to whether the reactor had been emplaced in undamaged
condition (or damage had been appropriately repaired) and all interface
requirements specified in the manufacturing license had been met. The
NRC believes that it has authority to issue a manufacturing license
under Section 161.h of the AEA.
The other model that the NRC could adopt would be a combination of
the approval processes used by the Federal Communications Commission
(FCC) and Federal Aviation Administration (FAA) in approving the
manufacture of electronic devices and airplanes. The NRC's
manufacturing license would approve: (1) the design of the nuclear
power reactor to be manufactured; (2) the specific manufacturing and
quality assurance/quality control processes and procedures to be used
during manufacture; and (3) tests and acceptance criteria for
demonstrating that the reactor has been properly manufactured. To be
completely consistent with the FCC and FAA models, the NRC would issue
a manufacturing license only after a prototype of the reactor had been
constructed and tested to demonstrate that all performance requirements
(i.e., compliance with NRC requirements and manufacturer's
specifications) can be met by the design to be approved for
manufacture.
The NRC requests public comment on whether the manufacturing
license process in proposed subpart F of part 52 should be further
extended in the final rule to provide an option for NRC approval of the
manufacturing, and if so, which model of regulatory oversight, i.e.,
the combined license ITAAC model or the FCC/FAA approval model, should
be used by the NRC. The NRC also seeks public comment on whether an
opportunity for hearing is required by the AEA in connection with a NRC
determination that the manufacturing ITAAC have been successfully
completed.
Commenters' Response: Some commenters requested that applicants for
manufacturing licenses be allowed, but not required, to use ITAAC to
ensure that an ``as-manufactured plant conforms to the important design
characteristics specified in the application for the manufacturing
license.'' Some commenters stated that a manufacturing license for
evolutionary designs should be subject to proposed Sec. 50.43(e) and
should not require a prototype. Some commenters stated that
manufacturing licenses should not be subject to more stringent
requirements than design certifications.
NRC Response: The NRC has decided to defer consideration of this
alternative on ITAAC, for several reasons. First, one commenter's
proposal to allow ITAAC for assuring that the as-manufactured reactor
``conforms to the important design characteristics specified in the
application for the manufacturing license,'' raises questions about
what those ``important design characteristics'' might be, and why the
ITAAC would be so narrowly limited. The Commission did not receive any
in-depth comments presenting arguments one way or the other on the
feasibility of developing such ITAAC, and the potential legal
implications of, and technical considerations with respect to, such a
finding by the manufacturer. Moreover, it is clear that any regulatory
process that the Commission may adopt in rulemaking would require
further opportunity for public comment, and therefore could not be
adopted in a final part 52 rulemaking without substantial delay. In
light of the lack of any near-term interest by any entity in obtaining
a manufacturing license, the Commission has decided not to adopt any
provisions for ITAAC governing approval of manufacturing in the final
part 52 rule. However, the Commission would address these issues in a
timely fashion if raised in a rulemaking
[[Page 49357]]
petition which demonstrated near-term interest in an application for a
manufacturing license.
The Commission agrees with the commenters'' suggestions that
manufacturing licenses for evolutionary designs should be subject to
new Sec. 50.43(e), and that under those provisions a prototype would
not be prerequisite to issuance of a manufacturing license for an
evolutionary design. Further discussion is provided below in Testing
Requirements for Advanced Reactors.
Question 5: Currently, part 52 allows an applicant for a
construction permit to reference either an early site permit under
subpart A of part 52 or a design certification (DC) under subpart B of
part 52. Specifically, Sec. 52.11 states that subpart A of part 52
sets out the requirements and procedures applicable to NRC issuance of
early site permits for approval of a site or sites for one or more
nuclear power facilities separate from the filing of an application for
a construction permit or combined license for such a facility.
Similarly, Sec. 52.41 states that subpart B of part 52 sets out the
requirements and procedures applicable to NRC issuance of regulations
granting standard design certification for nuclear power facilities
separate from the filing of an application for a construction permit or
combined license for the facility. However, the current regulations in
10 CFR part 50 that address the application for and granting of
construction permits do not make any reference to a construction permit
applicant's ability to reference either an early site permit or a
design certification. Also, the NRC has not developed any guidance on
how the construction permit process would incorporate an early site
permit or design certification, nor has the nuclear power industry made
any proposals for the development of industry guidance on this subject.
The NRC has not received any information from potential applicants
stating an intention to seek a construction permit for the construction
of a future nuclear power plant. In addition, the NRC recommends that
future applicants who want to construct and operate a commercial
nuclear power facility use the combined license process in subpart C of
part 52. Therefore, the NRC is considering removing from part 52, in
the final rule, the provisions allowing a construction permit applicant
to reference an early site permit or a design certification and is
interested in stakeholder feedback on this alternative.
Commenters' Response: Some commenters stated the deletion of
provisions allowing a construction permit applicant to reference an ESP
or DC was ill-advised given the untested nature of the COL process and
the resulting need to retain ``regulatory flexibility'' to deal with
unexpected issues. As a contingency plan to buffer against difficulties
with COL process, the commenters proposed the addition of a provision
in part 50 to specify that a construction permit applicant could
reference a DC without the inclusion of ITAAC. The commenters suggested
that in these instances, ``the operating license proceeding would need
to find under 10 CFR 50.57(a)(1) that construction of the facility has
been substantially completed, in conformity with the construction
permit and the application as amended, the provisions of the Act, and
the rules and regulations of the Commission.'' Commenters stated that
standard design should be final and not open to review in the
construction permit and operating licenses proceeding. Commenters
requested a construction permit applicant be able to reference an ESP
in the same way as would a COL applicant.
NRC Response: Based on some of the commenters' responses to this
question and further consideration of the issue, the NRC has decided
not to make any changes in the final rule to delete provisions allowing
a construction permit applicant to reference an early site permit or a
design certification. The NRC has also decided not to add any
additional provisions to part 50 or part 52 to address a construction
permit applicant's ability to reference either a design certification
or an early site permit. The NRC believes it is unlikely that such a
construction permit application will be submitted, and the NRC will
handle any such applications on a case-by-case basis. If such an
application were submitted, there are many process issues that would
need to be carefully considered and would need to be discussed with the
applicant and other stakeholders. In particular, the previously
certified designs all used design acceptance criteria in lieu of
detailed design information. A process for completing that design
information without using ITAAC would have to be developed.
Question 6: The NRC is considering revising Sec. 52.103(a) in the
final rule to require the combined license holder to notify the NRC of
the licensee's scheduled date for loading of fuel into a plant no later
than 270 days before the scheduled date, and to advise the NRC every 30
days thereafter if the date has changed and if so, the revised
scheduled date for loading of fuel. The initial notification would
facilitate timely NRC publication of the notice required under Sec.
52.103(a) and NRC staff scheduling of inspection and audit activities
to support NRC staff determinations of the successful completion of
ITAAC under Sec. 52.99. The proposed updating would also facilitate
NRC staff scheduling of those inspection and audit activities,
Commission completion of hearings within the time frame allotted under
Sec. 52.103(e), and any Commission determinations on petitions as
provided under Sec. 52.103(f). The NRC requests public comment on the
benefits and impacts (including information collection and reporting
burdens) that would occur if the proposed requirements were adopted.
Commenters' Response: Some commenters agreed with this concept.
However, they do not support a rule change because they believe a rule
change is not necessary. Rather, they believe that the concept should
be implemented via guidance rather than a rule change. Additionally,
following the initial notification, a licensee should be required to
submit a follow-up 30-day notification only if the schedule in the
prior notification has changed. It would be unnecessarily burdensome to
require a licensee to submit notifications every 30 days stating that
the schedule has not changed.
NRC Response: The NRC has decided to amend Sec. 52.103(a) in the
final rule to ensure that the combined license holder will notify the
NRC of its scheduled date for initial loading of fuel into a plant no
later than 270 days before the scheduled date, and will notify the NRC
of updates to its schedule every 30 days thereafter. The notification
will facilitate timely NRC publication of the notice required under
Sec. 52.103(a), completion of hearings within the time frame allotted
under Sec. 52.103(e), and completion of any Commission determinations
on petitions filed under Sec. 52.103(f). The NRC believes that the
update notifications when the schedule has not changed will not be
burdensome. Additional discussion on this issue is provided in Section
V.C.8.b of the supplementary information in this final rule.
Question 7: As discussed in Section IV.C.6.f of the March 13, 2006,
proposed rule, the NRC is proposing to modify Sec. 52.79(a) to add
requirements for descriptions of operational programs that need to be
included in the final safety analysis report (FSAR) to allow a
reasonable assurance finding of acceptability. This proposed amendment
is in support of the Commission's direction to the staff in SRM-SECY-
02-0067 dated September 11, 2002, ``Inspections, Tests, Analyses, and
Acceptance Criteria for Operational
[[Page 49358]]
Programs (Programmatic ITAAC),'' that a combined license applicant was
not required to have ITAAC for operational programs if the applicant
fully described the operational program and its implementation in the
combined license application. In this SRM, the Commission stated:
[a]n ITAAC for a program should not be necessary if the program
and its implementation are fully described in the application and
found to be acceptable by the NRC at the COL stage. The burden is on
the applicant to provide the necessary and sufficient programmatic
information for approval of the COL without ITAAC.
Accordingly, the NRC is proposing in the final part 52 rulemaking
to add requirements to Sec. 52.79 that combined license applications
contain descriptions of operational programs. In doing so, the
Commission has taken into account NEI's proposal to address SRM-SECY-
04-0032 in its letter dated August 31, 2005 (ML052510037). However, the
NRC is concerned that there may be operational program requirements
that it has not captured in its proposed Sec. 52.79. Therefore, the
NRC is requesting public comment on whether there are additional
required operational programs that should be described in a combined
license application that are not identified in proposed Sec. 52.79. If
additional required operational programs are identified, the Commission
is considering adding them to Sec. 52.79 in the final rule.
Commenters' Response: Some commenters believed that requirements
for operational programs were sufficient as proposed, and that no
additional operational programs needed to be described in the COL
application.
NRC Response: The NRC does not agree that no additional operational
programs need to be described in a COL application. During the
preparation of the final rule, the NRC discovered that several of the
operational programs listed in SECY-05-0197 (October 28, 2005) were not
addressed in proposed Sec. 52.79. To ensure the list of requirements
for the contents of applications is complete, the NRC is adding several
new provisions to address operational programs in the final rule.
Specifically, the NRC is adding requirements to Sec. 52.79 for COL
applicants to include a description of: (1) the process and effluent
monitoring and sampling program required by appendix I to 10 CFR part
50 [Sec. 52.79(a)(16)(ii)]; (2) a training and qualification plan in
accordance with the criteria set forth in appendix B to 10 CFR part 73
[Sec. 52.79(a)(36)(ii)]; (3) a description of the radiation protection
program required by Sec. 20.1101 [Sec. 52.79(a)(39)]; (4) a
description of the fire protection program required by Sec. 50.48
[Sec. 52.79(a)(40)]; and (5) a description of the fitness-for-duty
program required by 10 CFR part 26 [Sec. 52.79(a)(44)]. During the
preparation of the final rule, the NRC also noticed that it had not
completely implemented the Commission's direction regarding the
treatment of operational programs in a COL application because it had
failed to add requirements to address program implementation in its
revisions to Sec. 52.79(a). Therefore, in the final rule, the NRC has
added requirements to address the implementation of all operational
programs required to be described in a COL application. This is
consistent with the Commission's direction to the staff in SRM-SECY-02-
0067 (September 11, 2002, ML022540755) that a combined license
applicant was not required to have ITAAC for operational programs if
the applicant fully described the operational program and its
implementation in the combined license application.
Question 8: Backfitting--reproduce backfitting requirements in part
52. The NRC notes that the backfitting provisions applicable to various
part 52 processes are contained in both part 50 and part 52 and,
therefore, the proposed language for Sec. 50.109 cross-references to
applicable provisions of part 52, which may be confusing. The NRC is
considering adopting in the final rule an alternative which would
remove from Sec. 50.109 the backfitting provisions applicable to the
licensing and approval processes in part 52, and place them in part 52.
There are two possible approaches for doing so: the first would be for
the NRC to establish a general backfitting provision in part 52
applicable exclusively to the licensing and approval processes in part
52. Under this approach, each licensing and approval process in part 52
would be the subject of a backfitting section in a new subpart of part
52 (e.g., Sec. 52.201 for standard design approvals, etc.). The
existing backfitting provisions applicable to early site permits and
design certification would be transferred to the relevant sections in
the new subpart. The second approach would be to ensure that each
subpart of part 52 contains the backfitting provisions applicable to
the licensing or approval process in that subpart. The NRC is
considering adopting these alternative approaches in the final rule and
requests public comment on whether either of these administrative
approaches is preferable to the approach in the proposed rule.
Commenters' Response: Some commenters stated that NRC's alternative
approach to addressing backfitting was unnecessary to clarify the
application of the backfit rule to part 52 actions. Commenters stated
that the proposed rule included adequate references to Sec. 50.109 and
in the various subparts of part 52, making replication of this language
elsewhere unnecessary. If the NRC deemed the inclusion of such
information necessary, several commenters suggested each subpart in
part 52 include its own standards for backfitting to avoid confusion.
NRC Response: The NRC has decided to revise Sec. 50.109 to include
the conforming changes necessary to reflect part 52, rather than
adopting a backfitting provision in part 52, because no commenter
favored the alternative approach of adopting a backfitting provision in
part 52, and both approaches are legally equivalent.
Question 9: The Commission is considering adopting in the final
part 52 rulemaking an alternative to the re-proposed rule's approach
for addressing new and significant environmental information with
respect to matters addressed in the ESP environmental impact statement
(EIS) which require supplementation.\2\ As a separate matter, the
Commission is also considering adopting in the final part 52 rulemaking
an analogous requirement for addressing new information necessary to
update and correct the emergency plan approved by the ESP, the ITAAC
associated with EP, or the terms and conditions of the ESP with respect
to emergency preparedness, or new information materially changing the
Commission's determinations on emergency preparedness matters
previously resolved in the ESP. To implement either or both of these
alternatives, the Commission is also evaluating whether several
additional concepts should be adopted in the final rulemaking. The two
alternatives, as well as the additional implementing concepts, are
described below. The Commission emphasizes that it may, with respect to
the alternative addressing updating environmental information and
emergency preparedness information, adopt either or both alternatives
in the final part 52
[[Page 49359]]
rulemaking, in place of or in addition to the proposed rule's
alternative of conducting the updating in each combined license
proceeding. Under the option where multiple alternatives for updating
environmental and emergency preparedness information would be allowed,
the Commission proposes that the decision be left to the combined
license applicant as to which alternative to pursue. Commenters are
requested to address: (1) the advantages and disadvantages of adopting
each alternative for updating environmental and emergency preparedness
information in an ESP proceeding as opposed to the proposed rule's
alternative of conducting the updating in each combined license
proceeding; (2) whether the Commission should only allow updating of
environmental and emergency preparedness information in an ESP
proceeding or in a COL proceeding, but not both; and (3) if the
Commission allows updating in either an ESP proceeding or in a COL
proceeding, whether it should be an option for the COL applicant to
decide which update process to pursue. The Commission believes it may
allow COL applicants the option of deciding whether to update
environmental and emergency preparedness information in either an ESP
proceeding or in a COL proceeding in order to afford the COL applicant
the determination which approach best satisfies their business and
economic interests.
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\2\ The scope of environmental information that must be
supplemented is limited to the matters which were addressed in the
original EIS for the ESP. Thus, for example, if the ESP applicant
chose not to address need for power (as is allowed under Sec.
52.18), the combined license applicant need not address need for
power in its environmental report (ER) to update the ESP EIS, and
the NRC need not determine whether there is new and significant
information with respect to need for power as part of the updating
of the ESP EIS.
---------------------------------------------------------------------------
Environmental Matters Resolved in ESP
The Commission is considering requiring a combined license
applicant planning to reference an ESP to submit a supplemental
environmental report for the ESP. The supplemental environmental report
must address whether there is any new and significant environmental
information with respect to the environmental matters addressed in the
ESP EIS. Based upon this information, the NRC will prepare a draft
supplemental environmental assessment (EA) or EIS setting forth the
agency's proposed determinations with respect to any new and
significant information. In accordance with existing practice and
procedure, the draft supplemental EA or EIS will be issued for public
comment. After considering comments received from the public and
relevant Federal and State agencies, the NRC will issue a final
supplemental EA or EIS. Once the final supplemental EA or EIS is
issued, the ESP finality provisions in proposed Sec. 52.39 would apply
to the matters addressed in the supplemental EA or EIS, and those
matters need not be addressed in any combined license proceeding
referencing the ESP. Thus, for example, if a new and significant
environmental issue, for example, a newly-designated endangered
species, is addressed in the supplemental ESP EIS, the matter would be
resolved for all combined licenses referencing the ESP (unless, of
course, there is new and significant information identified at the time
of a subsequent referencing combined license with respect to that
endangered species). There would be no updating of environmental
information necessary in the combined license proceeding. The
Commission considers this approach for updating the ESP as meeting the
Agency's obligations under the National Environmental Policy Act
(NEPA), without imposing undue burden on the ESP holder and the NRC
through continuous or periodic updating, and preserving the distinction
between the ESP and any referencing combined license proceeding. Since
an ESP may be referenced more than once, this approach would provide
for issue finality of the updated information and preclude the need for
reconsideration of the same environmental issue in successive combined
license proceedings referencing the ESP. The Commission requests public
comment on this proposal, which would likely involve changes to
Sec. Sec. 52.39, 51.50(c), 51.75, and 51.107 (and possibly conforming
changes in parts 2, 51, and 52).
Emergency Preparedness Information Resolved in ESP
The Commission is separately considering requiring a combined
license applicant referencing an ESP to provide to the NRC new EP
information necessary to correct inaccurate information in the ESP
emergency plan, EP ITAAC, or the terms and conditions of the ESP with
respect to EP. Based upon the EP information submitted by the combined
license applicant, the NRC will, as necessary, approve changes to the
ESP emergency plan, the EP ITAAC, or the terms and conditions of the
ESP with respect to EP. Once the Commission has resolved the EP
updating matters, these matters would be accorded finality under Sec.
52.39. There would be no separate updating necessary in the combined
license proceeding. Thus, for example, if an EP ITAAC in an ESP were
changed by virtue of this updating process, the changed ITAAC for EP
would be applicable to any combined license referencing the ESP whose
ITAAC have not yet been satisfied (i.e., the amended EP ITAAC would not
be applicable to a combined license where the Commission has made the
Sec. 52.103(g) finding with respect to that EP ITAAC). The NRC's
consideration of such EP information would be considered to be part of
the ESP proceeding, and any necessary changes with respect to EP would
therefore be deemed to be changes within the scope of the ESP. The
Commission considers this proposal as a means for updating the ESP with
respect to EP information in a timely fashion, without imposing undue
burden on the ESP holder and the NRC through continuous or periodic
updating, while preserving the distinction between the ESP and any
referencing combined license proceeding.
Since an ESP may be referenced more than once, this approach would
provide for issue finality of the updated information and preclude the
need for reconsideration of the same issue in successive combined
license proceedings referencing the ESP. The Commission requests
comment whether this approach should be adopted by the Commission in
the final rulemaking, which will likely involve changes to Sec. 52.39
(and possible conforming changes in Sec. 50.47, 50.54, and 10 CFR part
50, appendix E).
ESP Updating in Advance of Combined License Application Submission
To minimize the possibility that the ESP updating process may
adversely affect a combined license proceeding referencing that ESP,
the Commission proposes to require the combined license applicant
intending to reference an ESP to submit its application to update the
ESP with respect to EP and/or environmental information no later than
18 months before the submission of its combined license application.
The Commission believes that the 18-month lead time is sufficient to
complete the NRC's regulatory consideration of the updating, such that
the combined license applicant will be able to prepare its application
to reflect the updated ESP. The Commission also recognizes that there
may be increased regulatory complexity under this approach, as well as
the possibility that resources may be unnecessarily expended if the
potential combined license applicant ultimately decides not to proceed
with its application. The Commission requests public comment on whether
the 18-month lead time is appropriate, whether the time should be
decreased or increased, or whether the Commission should simply require
that the ESP update application be filed no later than simultaneously
with the filing of the combined license application. Based upon the
public comments, the Commission will adopt one of these
[[Page 49360]]
alternatives, if it decides that updating of environmental and/or EP
matters should be accomplished in an ESP proceeding, as opposed to the
combined license proceeding in which the ESP is referenced.
Expanding the Scope of Resolved Issues After ESP Issuance
The Commission is also considering whether the final rule should
include provisions addressing how the ESP holder may request, at any
time after the issuance of the ESP, that additional issues be resolved
and given finality under Sec. 52.39. For example, the holder of the
ESP which does not include an approved emergency plan, may wish to
submit complete emergency plans for NRC review and approval. Such a
request is not explicitly addressed in either the current or re-
proposed subpart A to part 52, although it would be reasonable to treat
that request as an application to amend the ESP.
The Commission requests public comment on whether the Commission
should adopt in the final rule new provisions in subpart A to part 52
that would explicitly address requests by the ESP holder to amend the
early site permit to expand the scope of issues which are resolved and
given issue finality under Sec. 52.39. The Commission is also
considering whether, as part of the ESP updating process discussed
previously, the ESP holder/combined license applicant should be allowed
to request an expansion of issues which are resolved and given issue
finality.
If the Commission were to allow an ESP holder/combined license
applicant to expand the scope of resolved issues in the ESP update
proceeding, the Commission believes that the 18-month time period for
filing the updating application in the ESP proceeding may be
insufficient, and is considering adopting in the final rule a 24-month
(2-year) period for filing the ESP updating application, where the ESP
holder/combined license applicant seeks to expand the scope of resolved
issues. The Commission seeks public comment on whether, in such cases,
the Commission should require in the final rule an 18- or 24-month
period, or some other period, for submitting its ESP updating
application.
Approval in ESP of Process and Criteria for Updating ESP After Issuance
The Commission requests public comment whether the Commission
should adopt in the final rulemaking provisions affording the ESP
applicant the option of requesting NRC approval of procedures and
criteria for identifying and assessing new and significant
environmental information, and/or new information necessary to update
and correct the emergency plan approved by the ESP, the ITAAC
associated with emergency preparedness (EP), or the terms and
conditions of the ESP with respect to emergency preparedness, or
otherwise materially changing the Commission's determinations on
emergency preparedness matters previously resolved in the ESP. These
procedures and criteria, if approved as part of the ESP issuance, could
be used by any combined license applicant referencing the ESP to
identify the need to update the ESP with respect to environmental and/
or emergency preparedness information. There would be no need for the
NRC to review the adequacy of the ESP holder/combined license
applicant's process and criteria for determining whether new
information is of such importance or significance so as to require
updating; the NRC review could thereby be focused solely on whether the
ESP holder's updated information, or determination that there is no
change in either an environmental or emergency preparedness matter, was
correct and adequate. Under this proposal, Sec. 52.17 and/or Sec.
51.50(b) would be amended to incorporate such a process for ``pre-
approval'' of ESP updating procedures and criteria.
While NRC approval of updating procedures and criteria would be
reflected in the ESP, the Commission does not believe that the ESP
itself must contain the procedures and criteria in order to be accorded
finality under Sec. 52.39. An ESP holder/combined license applicant
need not comply with any or all of the updating process and criteria,
and would be free to use (and justify) other procedures or criteria in
the ESP updating proceeding. Naturally, there would be no finality
associated with such departures from the ESP-approved procedures and
criteria.
The Commission does not believe that either subpart A of part 52 or
an ESP with the contemplated approved updating procedures and criteria
should contain a ``change process'' akin to Sec. 50.59, allowing the
ESP holder to make changes to the approved updating procedures and
criteria without NRC review and approval. Any change (other than
typographic and administrative corrections) should require an amendment
to the ESP. However, the Commission seeks public comment on whether a
different course should be adopted in the final rule.
The Commission recognizes that any NRC-approved procedures and
criteria for updating environmental and/or emergency preparedness
information in an ESP updating process as described previously, would
be equally valid for updating such information under the updating
provisions in the re-proposed rule. The Commission requests comments on
whether, if the Commission adopts in the final rulemaking the re-
proposed rule's concept of updating in the combined license proceeding,
the Commission should provide the ESP applicant with the option of
seeking NRC approval of the procedures and criteria for updating
environmental and/or emergency preparedness information in a combined
license proceeding which references the ESP.
Public Participation in ESP Updating Process
The Commission is considering two ways for allowing public
participation in the updating process, if the updating alternative is
adopted in the final rule. One approach would be to allow interested
persons to challenge the proposed updating by submitting a petition,
analogous to that in proposed Sec. 52.39(c)(2), which would be
processed in accordance with Sec. 2.206. This approach would be most
consistent with the existing provisions in Sec. 52.39, inasmuch as
updating of an ESP is roughly equivalent to a request that the terms
and conditions of an ESP be modified. A consequence of this approach is
that the potential scope of matters which may be raised is not limited
to those ESP matters which the ESP holder/combined license applicant
and the NRC conclude must be updated.
The other approach that the Commission may adopt is to treat any
necessary updating as an amendment to the ESP, for which an opportunity
to request a hearing is provided. This approach would limit the scope
of the hearing to those matters for which an amendment is required.
Where the ESP holder does not request an amendment on the basis that no
updating is necessary with respect to a matter, an interested person
could not intervene with respect to that matter. A consequence of this
approach is that, under the Commission's regulations in 10 CFR part 2
and its current practice, a hearing granted on any amendment
necessitated by the updating process would be more formalized than a
hearing accorded under the Sec. 2.206 petition process. The Commission
requests public comment on the approach that the Commission should
adopt, together with the reasons for the commenter's recommendation.
Commenters' Response: Several commenters believed an ESP holder
should not be required to update the
[[Page 49361]]
information in the ESP application. These commenters stated that the
proposal to require updating would add an unnecessary additional level
of review (and possibly hearings) with little or no additional benefit
(i.e., the COL applicant would still be under the obligation to update
the information provided by the ESP holder). Some commenters contended
that an updating requirement would only serve to erode the finality and
certainty provided by the ESP, thereby defeating one of the purposes of
an ESP. These commenters also believed that an updated requirement
would run counter to NRC regulations. Some commenters stated that while
the ESP is in effect, the NRC cannot change or impose new requirements,
including emergency planning requirements, unless it determines that a
modification is necessary either to bring the permit or the site into
compliance with the NRC's regulations and orders applicable and in
effect at the time the permit was issued, or to assure adequate
protection of the public health and safety or the common defense and
security. Some commenters argued that the proposed 18-month updating
requirement may not be feasible. A commenter gave the following
example, ``under the NRC's current schedule for the existing ESP
applications for North Anna and Grand Gulf, the ESPs will not be issued
until 2007, shortly before the planned COL applications for those
sites. This would result in insufficient time for the updating
envisioned by the NRC, and it would be unfair to those applicants to
require them to delay their COL applications to accommodate the
updating process. Additionally, the proposed updating process would be
inconsistent with Sec. 52.27(c), which permits a COL application to
reference an ESP application.''
Several commenters agreed with NRC's proposal to provide the ESP
holder with the option of requesting an ESP amendment in order to
resolve issues that were not addressed at the ESP stage or to achieve
finality on updated information. These commenters also suggested that a
COL applicant should be able to reference an application for an ESP
amendment that is pending approval by the NRC similar to the process
that already exists in 10 CFR 52.27(c).
Several commenters expressed the belief that a COL applicant should
be able to make changes or updates to ESP emergency planning
information without NRC approval in accordance with the criteria in 10
CFR 50.54(q) just as the remaining safety information can be revised
under Sec. 50.59 once it has been reviewed and approved. These
commenters also stated that this revised information should not be
considered as an ``amendment'' submitted under Sec. 50.90 for review
and approval, but rather should be considered to be information
equivalent to that provided under Sec. 50.71(e) for information.
NRC Response: Upon consideration of the public comments on this
subject, the NRC has decided not to require updating of ESP information
prior to receipt of a COL application referencing the ESP. The NRC is
retaining the proposed rule structure for dealing with new EP and
environmental information at the COL stage. The NRC believes this
structure will provide for the most effective and efficient use of NRC
and applicant resources. The NRC is, however, making revisions to the
final rule to allow for voluntary changes to an ESP by the ESP holder
through the license amendment process. Specifically, the NRC is making
revisions to Sec. Sec. 50.90 and 50.92 to include ESPs within the
scope of these requirements. The NRC is also adding a new provision to
Sec. 52.39 to allow ESP holders to make changes to the ESP, including
changes to the SSAR, under the license amendment process. These changes
will provide ESP holders with additional flexibility to resolve issues
that were not addressed in the original ESP review and to achieve
finality on new information. The NRC does not believe it is necessary
to add rule language to address the situation where a COL applicant
references an ESP for which there is an amendment review pending before
the NRC. The NRC will address these situations on a case-by-case basis.
Question 10: The Commission is considering adopting in the final
part 52 rulemaking a new provision in Sec. 50.71 that would require
combined license holders to update the PRA [probabilistic risk
assessment] submitted with the combined license application
periodically throughout the life of the facility on a schedule similar
to the schedule for final safety analysis report (FSAR) updates (i.e.,
at least every 24 months) or, alternatively, on a schedule to coincide
with every other refueling outage. Updates would be required to ensure
that the information included in the PRA contains the latest
information developed. The PRA update submittal would be required to
contain all the changes necessary to reflect information and analyses
submitted to the Commission by the licensee or prepared by the licensee
pursuant to Commission requirement since the submittal of the original
PRA, or as appropriate, the last update to the PRA under this section.
The submittal would be required to include the effects of all changes
made in the facility or procedures as reflected in the PRA; all safety
analyses and evaluations performed by the licensee either in support of
approved license amendments or in support of conclusions that changes
did not require a license amendment in accordance with Sec.
50.59(c)(2) or, in the case of a license that references a certified
design, in accordance with Sec. 52.98(c); and all analyses of new
safety issues performed by or on behalf of the licensee at Commission
request. The Commission requests stakeholder feedback on whether such a
requirement should be added to the Commission's regulations and, if so,
what is an appropriate update schedule.
Commenters' Response: Several commenters noted that the proposed
rule did not include a frequency for updating the PRA. These commenters
noted that the Commission stated that PRA scope and methods should be
addressed in guidance, not in regulations (SRM on SECY-05-0203). These
commenters stated that they believed that PRA update frequency should
also be addressed in guidance rather than regulations. These commenters
indicated a frequency of once every two operating cycles would be
reasonable and consistent with existing requirements in 10 CFR
50.69(e).
Additionally, some commenters stated the plant-specific PRA used to
support a COL application that references a design certification would
essentially be the design certification PRA. These commenters expressed
the belief that the plant-specific PRA would be updated to be
consistent with the PRA scope and quality standards 6 months before the
COL was issued as plant-specific design and as-built information was
developed during construction. Some commenters argued that this would
allow (1) an updated plant-specific PRA that was representative of the
as-built plant to be completed, and (2) an updated plant-specific PRA
that would be available prior to fuel load for NRC audit and to support
plant operations. These commenters suggested that the update of the
plant-specific PRA during construction was a matter suitable for
guidance.
Some commenters expressed confusion over the NRC proposal to
require PRA updates to reflect safety analyses and evaluations
performed by the licensee, and analyses of new safety issues performed
by or on behalf of the licensee at the NRC's request. These commenters
stated that new analyses
[[Page 49362]]
and evaluations were often performed using design-basis assumptions
that may not be appropriate for a PRA. These commenters suggested that
only new analyses that impact the PRA warrant consideration, and
requested guidance and examples be developed regarding the information
that should be considered when updating the plant-specific PRA.
NRC Response: As discussed in further detail in Section V.D.6.b of
this document, the Commission is adopting requirements to require
maintenance of a PRA, and periodic upgrades every 4 years, by a COL
holder beginning at the time of initial operation. These PRAs and
upgrades are not required to be submitted to the NRC, but instead
should be maintained by the licensee for NRC inspection.
Question 11: In a letter dated July 5, 2005, the Nuclear Energy
Institute (NEI) submitted comments on the proposed rule for the AP1000
design certification. Many of those comments have generic applicability
to the three pre-existing design certification rules (DCRs) in
appendices A through C of 10 CFR part 52. In the final AP1000
rulemaking (January 27, 2006; 71 FR 4464), the Commission adopted some
of the NEI-recommended changes, while rejecting others (71 FR 4465-
4468). For those changes that were adopted in the final AP1000 design
certification, the Commission indicated that it would consider making
the same changes to the existing design certifications in appendices A
through C. For those changes that were not adopted in the final AP1000
design certification, the Commission stated that it would reconsider
the issues in the part 52 rulemaking, and if the Commission changes its
position and the change is adopted, the Commission would make the
change for all four design certifications, including the AP1000.
The Commission is considering amending the appropriate sections in
each DCR based on the comments below. The Commission considers most of
NEI's proposed changes to be consistent with proposed Sec.
52.63(a)(1); in particular, the Commission believes that the proposed
changes would satisfy the ``reduces unnecessary regulatory burden''
criterion in proposed Sec. 52.63(a)(1)(iii). The few remaining
changes, constituting editorial clarifications or corrections
reflecting the Commission's original intent, are not subject to the
existing change restrictions in Sec. 52.63(a)(1). Accordingly, the
Commission believes that it has authority to incorporate some or all of
the NEI-proposed changes into appendices A through D in the final part
52 rulemaking.
The Commission also requests comments on whether some of NEI's
proposed changes accepted in the AP1000 design certification and
proposed for inclusion in appendices A through C should not be included
in those appendices in the final part 52 rulemaking because they are
unnecessary, or because they would not meet one or more of the change
criteria in proposed Sec. 52.63(a)(1). The Commission is also
assessing whether NEI's proposed changes which were not adopted in the
AP1000 final rulemaking should be adopted in the final part 52
rulemaking for all four design certifications, including the AP1000.
The Commission is particularly interested in whether there are reasons,
other than those presented by NEI, for adopting those changes, as well
as commenter's views on the Commission's reasons for rejecting the NEI
proposals as stated in the final AP1000 design certification
rulemaking.
a. NEI recommended modification of the generic technical
specification definition in Section II.B to clarify that bracketed
information is not part the DCRs for purposes of the change processes
in Section VIII.C, and an exemption is not required for plant-specific
departures from bracketed information. The Commission stated in the
section-by-section analysis for the AP1000 DCR (71 FR 4464) that some
generic technical specifications and investment protection short-term
availability controls contain values in brackets. The values in
brackets are neither part of the DCR nor are they binding. Therefore,
the replacement of bracketed values with final plant-specific values
does not require an exemption from the generic technical specifications
or investment protection short-term availability controls. The
Commission believes that including this guidance in each DCR is not
necessary. The Commission requests comment on whether there are
countervailing considerations that favor inclusion of this provision in
the DCRs.
b. NEI recommended modification of the Tier 2 definition in Section
II.E to clarify that bracketed information in the investment protection
short-term availability controls is not part of Tier 2 and thus not
subject to the Section VIII.B change controls. The Commission stated in
the section-by-section analysis for the AP1000 DCR (71 FR 4464) that
some generic technical specifications and investment protection short-
term availability controls contain values in brackets. The values in
brackets are neither part of the DCR nor are they binding. Therefore,
the replacement of bracketed values with final plant-specific values
does not require an exemption from the generic technical specifications
or investment protection short-term availability controls. The
Commission believes that including this guidance in each DCR is not
necessary. The Commission requests comment on whether there are
countervailing considerations that favor inclusion of this provision in
the DCRs.
c. NEI recommended modification of the requirement in Section
VIII.C.2 to delete the phrase ``or licensee'' because that phrase
conflicted with the requirement in Section VIII.C.6. The Commission
believes that generic technical specifications should not apply to
holders of a combined license because the license will include plant-
specific technical specifications. Therefore, the Commission is
considering amending each of the DCRs to delete the phrase ``or
licensee'' from Section VIII.C.2 and requests public comment on this
approach.
d. NEI recommended modification of the requirement in Section
VIII.C.6 to delete the last portion, which states ``changes to the
plant-specific technical specifications will be treated as license
amendments under 10 CFR 50.90.'' NEI stated that this sentence is not
necessary because it is redundant with Sec. 50.90. It is not necessary
to include a provision in each DCR stating that a license amendment is
necessary to make changes to technical specifications in order to
render this a legally-binding requirement inasmuch as Section 182.a of
the AEA requires that technical specifications be part of each license.
The Commission believes that clarity and understanding by the reader is
enhanced by repeating this statutory requirement in each DCR. The
Commission requests comment on whether there are countervailing
considerations that favor non-inclusion of this provision in the DCRs,
and may decide to remove this provision in the final part 52
rulemaking.
e. NEI recommended modification of the requirement in Section X.A.1
to require the design certification applicant to include all generic
changes to the generic technical specifications and other operational
requirements in the generic DCD. The Commission believes that inclusion
of changes to the generic technical specifications and other
operational requirements will enhance the generic DCD and facilitate
its use by referencing applicants. The Commission is considering
amending each of the DCRs to include the generic technical
specifications and other operational requirements in the generic
[[Page 49363]]
DCD and requests public comment on this approach.
f. NEI recommended modification of the requirement in Sections
IV.A.2 and IV.A.3 to be consistent with respect to inclusion of
information in the plant-specific DCD, or explain the difference
between ``include'' (IV.A.2) and ``physically include'' (IV.A.3). The
Commission is considering amending each of the DCRs to use the same
term in both provisions, and requests public comment on this approach.
g. NEI recommended modification of the definition in Section II.E.1
to exclude the design-specific probabilistic risk assessment (PRA) and
the evaluation of the severe accident mitigation design alternatives
(SAMDA) from Tier 2 information. The Commission believes that the PRA
and SAMDA evaluations do not need to be included in Tier 2 information
because they are not part of the design basis information. The
Commission is considering amending each of the DCRs to modify the
definition of Tier 2, and requests public comment on this approach.
h. NEI recommended modification of the requirement in Section III.E
to use ``site characteristics'' consistently, instead of ``site-
specific design parameters.'' The Commission intends to use the term
``characteristics'' to refer to actual values and ``parameters'' to
refer to postulated values. The Commission has proposed amending
Section III.E of each DCR to use ``site characteristics,'' and requests
public comment on this approach.
i. NEI recommended modification of Section IV.A.2 to clarify the
use of ``same information'' and ``generic DCD'' in that requirement.
The Commission has proposed amending Section IV.A.2 of each DCR to use
the phrase ``same type of information'' to avoid confusion, and
requests public comment on this approach.
j. NEI recommended modification of the requirement in Section
VIII.B.6.a to delete the sentence ``The departure will not be
considered a resolved issue, within the meaning of Section VI of this
appendix and 10 CFR 52.63(a)(4),'' in order to be consistent with the
requirement in Section VI.B.5 of the DCRs. The Commission believes that
departures from Tier 2* information should not receive finality or be
treated as resolved issues within the meaning of section VI.B of the
DCRs. The Commission requests comment on whether departures from Tier
2* information should be considered a resolved issue, and may decide to
remove this provision from each DCR.
k. NEI recommended modification of Section VIII.C.3 to require the
NRC to meet the backfit requirements of 10 CFR 50.109 in addition to
the special circumstances in 10 CFR 2.758(b) (which has now been
designated as Sec. 2.335) in order to require plant-specific
departures from operational requirements. The Commission believes that
plant-specific departures should not have to meet the backfit
requirement for generic changes. The Commission will have to
demonstrate that special circumstances, as defined in Sec. 2.335, are
present in order to require a plant-specific departure. The Commission
requests comment on whether there are countervailing considerations
that would favor modification of this provision in the DCRs.
l. NEI recommended modification of the requirement in Section
VIII.C.4 to include a requirement that operational requirements that
were not completely reviewed and approved by the NRC should not be
subject to any Tier 2 change controls, e.g., exemptions. However, NEI
previously proposed that requested departures from Chapter 16 by an
applicant for a COL require an exemption (62 FR 25808; May 12, 1997).
The Commission believes that the requirement for an exemption applies
to technical specifications and operational requirements that were
completely reviewed and approved in the design certification rulemaking
(see 62 FR 25825). The Commission requests comment on whether
departures from technical specifications and operational requirements
that were not completely reviewed and approved should also require an
exemption.
m. NEI recommended modification of the requirement in Section
VIII.C.4 to delete the sentence ``The grant of an exemption must be
subject to litigation in the same manner as other issues material to
the license hearing,'' in order to be consistent with the requirement
in Section VI.B.5 of the DCRs. The Commission believes that exemptions
from operational requirements should not receive finality or be treated
as resolved issues (refer to Section VI.C of the DCRs). The Commission
requests comment on whether exemptions from operational requirements
should be considered a resolved issue, and may decide to modify this
provision in each DCR.
n. NEI recommended modification of the requirement in Section
IX.B.1 to better distinguish between NRC staff ITAAC conclusions under
proposed Sec. 52.99(e) and the Commission's ITAAC finding under
proposed Sec. 52.103(g). The Commission believes that individual DCRs
should not address the scope of the NRC staff's activities with respect
to ITAAC verification. This is a generic matter that, if it is to be
addressed in a rulemaking, is more appropriate for inclusion in subpart
C of part 52 dealing with combined licenses. The Commission requests
comment on whether there are countervailing considerations that favor
clarification of this provision in the DCRs.
o. NEI recommended modification of the language in Section IX.B.3
to make editorial changes for clarity, e.g., ``ITAAC will expire'' vs.
``their expiration will occur.'' The Commission believes that the
original rule language is acceptable. The Commission requests comment
on whether there are countervailing considerations that favor
clarification of this provision in the DCRs.
p. NEI recommended modification of the language in Sections X.B.1
and X.B.3 to clarify references to the design control documents, e.g.,
``plant-specific'' vs. ``generic.'' The Commission agrees that the
references to plant-specific and generic DCD should be clarified in
Sections X.B.1 and X.B.3 to ensure that the requirements in these
sections are properly implemented by applicants referencing the design
certification rules. The Commission requests public comment on this
prospective modification.
Commenters' Response: Several commenters recommended the NRC
incorporate the NEI recommendations on the AP1000 rule, cited specific
NEI recommendations (71 FR 12834-12836), and made additional
suggestions and clarifications.
Regarding NEI recommendations (a) and (b), several commenters
suggested it would be sufficient if the statements of considerations
for the final rule provided the requested clarification, rather than
the rule itself.
Regarding NEI recommendation (f), several commenters supported the
use of the term ``include'' rather than ``physically include'' for
requirements in Section IV of the design certification rules concerning
content of COLAs. These commenters also requested clarification on the
permissible method of incorporating the generic DCD into the plant-
specific DCD portion of the COL application's final safety analysis
report (FSAR), because the current NRC position has apparently ``led to
considerable confusion'' among COL preparers. These commenters noted
that in the statements of consideration accompanying the AP1000 final
rule, NEI recommended a change to the Definitions (Section III.B of
that rule, 71 FR 4466). These commenters stated the NRC staff disagreed
with this
[[Page 49364]]
recommendation, saying that ``the generic DCD should also be part of
the FSAR, not just incorporated by reference, in order to facilitate
the NRC staff's review of any departures or exemptions.'' Some
commenters believed that this NRC position was in conflict with the
former Sec. 52.79(b), which states that the COL application's FSAR
``may incorporate by reference the final safety analysis report for a
certified standard design,'' and with Sec. 50.32, which provides for
incorporation by reference to eliminate repetitive information. Some
commenters argued that although the wording had been altered, the
ability to incorporate by reference was preserved in proposed
Sec. Sec. 52.79 (b) and (c), respectively. These commenters claimed
this interpretation of incorporation was validated by NRC staff during
the Draft Regulatory Guide (DG)-1145 workshops. These commenters stated
support for this interpretation and requested the NRC explicitly
describe that either approach is acceptable.
In discussing NEI recommendation (j), several commenters mentioned
Section VIII.B.6.a of the design certification rules, which states that
an applicant who references the design certification rule must obtain
NRC approval for departures from Tier 2* information in the generic
DCD. Some commenters believed that this section states the departure is
not considered to be a resolved issue under Section VI of the design
certification rules. Some commenters indicated this was inconsistent
with Section VI.B.5 of the design certification rules, which states
that license amendments are considered to be resolved. These commenters
expressed support for the revision of Section VIII.B.6. of the design
certification rules to make it consistent with Section VIII.B.5 of the
design certification rules. These commenters stated that departures
from Tier 2* information that are reviewed and approved by the NRC in
the combined license proceeding should have finality for the plant in
question.
With respect to NEI recommendation (k), several commenters
expressed concern that Section VIII.C.3 of the design certification
rules ``inappropriately'' allowed the NRC to make changes to
operational requirements in the DCD without satisfying the backfit
requirements in Sec. 50.109. These commenters stated that the
operational requirements in the design certification proceeding should
be afforded the protection of the backfit rule. Some commenters
supported a revision to Section VIII.C.3 of the design certification
rules to include a reference to Sec. 50.109 for these changes.
In the discussion of NEI recommendations (l) and (m), several
commenters mentioned Section VIII.C.4 of the design certification
rules, which states a COL applicant must request an exemption from the
NRC if the applicant wants to depart from the generic technical
specifications or other operational requirements. These commenters
described this requirement as ``unduly burdensome.'' These commenters
noted that the operational requirements do not have finality under
Section VI.C of the design certification rules, and that no basis
existed for applying such a change control process to a COL applicant
seeking to change operational requirements. Some commenters cited
Section VIII.B.5 of the design certification rules, which states a COL
applicant may depart from final design-related provisions in the design
certification rule using a ``Sec. 50.59-like'' process, and argued
that imposing an exemption process with respect to operational
provisions was not required. Some commenters recommended Section
VII.C.4 be amended to state that a departure from an operational
requirement does not require an exemption.
Several commenters mentioned information from NEI's September 30,
2003, response to the 2003 part 52 notice of proposed rulemaking. These
commenters expressed support for the need to add a basic definition of
``departure'' to the DCRs to be consistent with adding the definition
of ``departure from a method of evaluation,'' and stated that both
should be based on Regulatory Guide 1.187. The commenters stated, ``The
basic definition of `change or departure' should precede the definition
of departure from a method of evaluation.'' Some commenters recommend
adding the new definition as paragraph II.G and renaming the final two
paragraphs as II.H and II.I.
NRC Response: In response to Question 11.a, the NRC has decided
that modification of the generic technical specification definition in
Section II.B of the DCRs is not necessary. As stated in the section-by-
section analysis for the AP1000 DCR (71 FR 4475; January 27, 2006):
Some generic technical specifications and investment protection
short-term availability controls contain values in brackets [ ]. The
brackets are placeholders indicating that the NRC's review is not
complete, and represent a requirement that the applicant for a
combined license referencing the AP1000 DCR must replace the values
in brackets with final plant-specific values. The values in brackets
are neither part of the design certification rule nor are they
binding. Therefore, the replacement of bracketed values with final
plant-specific values does not require an exemption from the generic
technical specifications or investment protection short-term
availability controls.
The NRC believes that the above guidance resolves NEI's concern
regarding bracketed information in the generic technical
specifications.
Regarding Question 11.b, the NRC has decided that modification of
the Tier 2 definition in Section II.E of the DCRs is not necessary. The
NRC believes that the previously mentioned guidance resolves NEI's
concern regarding bracketed information in the investment protection
short-term availability controls located in the Tier 2 information.
Regarding Question 11.c, the NRC agrees with NEI's recommendation
and has decided to delete the phrase ``or licensee'' from Section
VIII.C.2 of the DCRs because the generic technical specifications will
not apply to holders of a combined license.
Regarding Question 11.d, the NRC has decided not to modify the rule
language in Section VIII.C.6 of the DCRs, which states that ``changes
to the plant-specific technical specifications will be treated as
license amendments under 10 CFR 50.90.'' The Commission believes that
this statement provides clarity to this requirement.
Regarding Question 11.e, the NRC agrees with NEI's recommendation
and has decided to modify the requirement in Section X.A.1 of the DCRs.
The Commission believes that the inclusion of changes to the generic
technical specifications and other operational requirements in the
generic design control document (DCD) will enhance the DCD and
facilitate its use by referencing applicants.
Regarding Question 11.f, the NRC has decided to modify Section IV
of the DCRs to consistently use the term ``include'' rather than
``physically include'' as recommended by NEI.
Several commenters also requested clarification on the permissible
method of incorporating the generic DCD in the plant-specific DCD
portion of the COL application's final safety analysis report (FSAR),
because the NRC position has apparently ``led to considerable
confusion'' among COL preparers. The NRC is requiring COL applicants
that reference the DCRs in appendices A through D of part 52 to include
the generic DCD in the application's FSAR, in order to facilitate the
NRC staff's review of any departures or exemptions. Simply
incorporating the generic DCD by reference into the FSAR is not
[[Page 49365]]
sufficient because of the manner in which these existing DCDs were
submitted to the NRC. Therefore, Section IV.A.2 of the DCRs overrides
Sec. Sec. 50.32 and 52.79(d). The NRC is hopeful that future DCRs will
not have to use this special requirement.
Regarding Question 11.g, the NRC agrees with NEI's recommendation
and has decided to modify the definition of Tier 2 in Section II.E.1 of
the DCRs to exclude the design-specific probabilistic risk assessment
(PRA) and the evaluation of the severe accident mitigation design
alternatives (SAMDAs). The NRC believes that the PRA and SAMDA
evaluations do not need to be included in Tier 2 because they are not
part of the design basis information. Also, the revised Section II.E.1
is now consistent with the requirements in the new Sec. 52.80
regarding PRA and SAMDA evaluations.
Regarding Question 11.h, the NRC agrees with NEI's recommendation
to use ``site characteristics'' instead of ``site-specific design
parameters'' in Section III.E of the DCRs. This modification of the
rule language in Section III.E was made in the proposed rule and,
therefore, no change was made to the final rule.
Regarding Question 11.i, the NRC agrees with NEI's recommendation
to clarify the rule language in Section IV.A.2.a of the DCRs and adopts
the phrase ``same type of information'' to avoid confusion. An
applicant for a combined license must submit, as part of its
application, a plant-specific DCD that contains the same type of
information and uses the same organization and numbering as the generic
DCD. This organization will facilitate the NRC staff's review of the
plant-specific DCD. The NRC recognizes that the plant-specific DCD will
not contain the exact, same information as the generic DCD because the
plant-specific DCD will be modified and supplemented by the applicant's
exemptions, departures, and COL action items.
Regarding Question 11.j, the NRC does not agree with NEI's request
to modify the requirement in Section VIII.B.6.a of the DCRs. The
Commission decided during the initial design certification rulemakings
that departures from Tier 2* information (by an applicant) would not
receive finality or be treated as a resolved issue within the meaning
of Section VI of the DCR. This provision applies to applicants for a
combined license and the new information is subject to litigation in
the same manner as other plant-specific issues in the licensing
hearing. Also, Tier 2* information has the same safety significance as
Tier 1 information and would have received the Tier 1 designation,
except that NRC decided to provide more flexibility for this type of
information.
Regarding Question 11.k, the NRC does not agree with NEI's
recommendation to modify Section VIII.C.3 of the DCRs. NEI requests
that the NRC meet the backfit requirements in Sec. 50.109 in addition
to the special circumstances in Sec. 2.335 in order to require plant-
specific departures from operational requirements. In the original
design certification rulemakings, the Commission decided on different
standards for changes made under Section VIII.C (see Section VI.C and
62 FR 25805; May 12, 1997). The Commission has decided that plant-
specific departures should not have to meet the backfit requirements in
Sec. 50.109.
Regarding Question 11.l, the NRC does not agree with NEI's
recommendation to modify Section VIII.C.4 of the DCRs. The requirement
in Section VIII.C.4 for an applicant to request an exemption applies to
generic technical specifications and operational requirements that were
comprehensively reviewed and finalized in the design certification
rulemaking (see 62 FR 25825; May 12, 1997). Because this guidance is
already set forth in the section-by-section discussion for the DCRs,
the NRC has decided that changes to the rule language are not
necessary.
Regarding Question 11.m, the NRC does not agree with NEI's
recommendation to delete the last sentence from Section VIII.C.4 of the
DCRs. This sentence applies to applicants for a combined license and
the new information is subject to litigation in the same manner as
other plant-specific issues in the licensing hearing. The Commission
believes that exemptions from operational requirements should not
receive finality or be treated as resolved issues (refer to Section
VI.C of the DCRs).
Regarding Question 11.n, the NRC does not agree with NEI's
recommendation to modify Section IX.B.1 of the DCRs. The NRC has
decided that individual DCRs should not address the scope of the NRC
staff's activities with respect to ITAAC verification. This is a
generic matter that was addressed in Sec. 52.99(e).
Regarding Question 11.o, the NRC does not agree with NEI's request
to clarify the phrase ``their expiration will occur'' in Section IX.B.3
of the DCRs. The NRC has decided that the original rule language is
acceptable.
Regarding Question 11.p, the NRC agrees with NEI's recommendation
to clarify references to the DCDs in Sections X.B.1 and X.B.3 of the
DCRs. The references to plant-specific and generic DCD were revised in
Sections X.B.1 and X.B.3 to ensure that the requirements in these
sections will be properly implemented by applicants and licensees that
reference the design certification rules.
Question 12: The Commission is considering adopting in the final
part 52 rulemaking a new provision that would either require combined
license applicants to submit a detailed schedule for the licensee's
completion of ITAAC or require the combined license holder to submit
the schedule for ITAAC completion. Delaying submission of the schedule
would allow the combined license holder to develop the schedules based
on more accurate information regarding construction schedules and would
allow the schedule to be submitted at a time when it would be most
useful to the NRC for planning purposes. The Commission could require
that applicants submit the schedule within a specified time prior to
scheduled COL issuance--for example, 3 months prior to COL issuance or
within some time period (e.g., 6 months or 1 year) after COL issuance.
In addition, the Commission is considering an additional element to
this provision that would require that the licensee submit an update to
the ITAAC schedule within 12 months after combined license issuance and
that the licensee update the schedule every 6 months until 12 months
before scheduled fuel load, and monthly thereafter until all ITAAC are
complete. The Commission is considering adopting these requirements to
support the NRC staff's inspection and oversight with respect to ITAAC
completion, and to facilitate publication of the Federal Register
notices of successful completion of ITAAC as required by proposed Sec.
52.99(e). The Commission requests stakeholder comment on whether such a
provision, with or without the update element, should be added to the
Commission's regulations and which time frame for submission of the
schedule would be most beneficial.
The Commission is also considering adopting a provision that would
establish a specific time by which the licensee must complete all ITAAC
to allow sufficient time for the NRC staff to verify successful
completion of ITAAC, without adversely affecting the licensee's
scheduled date for fuel load and operation. The Commission considers
``60 days prior to the schedule date for initial loading of fuel'' to
be a
[[Page 49366]]
reasonable time period by which all ITAAC must be completed. However,
the Commission requests comments on whether this time period would
provide too much or too little time prior to scheduled fuel load.
Alternatively, the Commission is considering a 30-day or a 90-day time
period prior to scheduled fuel load. The 30-day option would allow more
flexibility for the licensee to complete ITAAC late in construction but
would require immediate action on the part of the NRC (to determine if
the final ITAAC were completed successfully and, if so, for the
Commission to make its finding under Sec. 52.103(g)) so as not to
delay scheduled fuel load. The 90-day option would reduce licensee
flexibility to complete ITAAC late in construction but would ensure
that the NRC had ample time to make its determination on the final
ITAAC for Commission review of all ITAAC under Sec. 52.103(g). The
Commission requests stakeholder comment on whether a provision
requiring completion of ITAAC within a certain time period prior to
scheduled fuel load should be added to the Commission's regulations.
Commenters' Response: Several commenters believed it was
unnecessary to include a requirement for either the COL applicant or
the COL holder to submit a detailed schedule for ITAAC completion
because a COL applicant could provide only a progressively less
accurate estimated completion schedule. Some commenters stated that the
COL holder would have schedules at the site, and those schedules would
be available for NRC review. Some commenters believed that COL holders
would interact and coordinate with the NRC to ensure that NRC had
sufficient information to schedule its inspection activities for ITAAC,
making a regulatory requirement for submission of a schedule
unnecessary. In addition, these commenters noted that a COL applicant/
holder would likely consider detailed schedule information to be
proprietary information, which would make its submission inappropriate.
Several commenters also stated it was ``wrong'' to require
completion of ITAAC in a set time period prior to fuel loading and
operation. These commenters indicated that a COL holder would likely
complete several ITAAC within 30 days of fuel loading and argued that
the NRC should not abrogate responsibility by imposing a mandatory
delay on licensees. Some commenters stated the importance of the NRC
providing the appropriate level of inspections and reviews to prevent
delays in fuel load and emphasized the high cost (stated to be on the
order of $1,000,000 per day) of such delay. Some commenters suggested
the NRC should be in a position to make a Sec. 52.103(g) finding
promptly following the completion of the last ITAAC.
NRC Response: The NRC has decided to amend Sec. 52.99 to require
licensees to submit their schedules for completing the inspections,
tests, or analyses in the ITAAC. The NRC has added a new paragraph (a)
in Sec. 52.99 that requires a licensee to submit to the NRC, no later
than 1 year after issuance of the combined license or at the start of
construction as defined in 10 CFR 50.10, whichever is later, its
schedule for completing the inspections, tests, or analyses in the
ITAAC. Licensees are required to submit updates to the ITAAC schedule
every 6 months thereafter and, within 1 year of its scheduled date for
initial loading of fuel, licensees must submit updates to the ITAAC
schedule every 30 days until the final notification is provided to the
NRC under Sec. 52.99(c)(1). Although commenters did not believe that a
requirement for submission of a schedule was necessary, the NRC
believes it is necessary to ensure that the NRC has sufficient
information to plan all of the activities necessary for the NRC to
support the Commission's determination as to whether all of the ITAAC
have been met prior to initial operation. In the event that licensees
consider their schedule information to be proprietary, they can request
that the schedule be withheld from public disclosure under Sec. 2.390.
If an applicant claims that its construction schedule information
submitted to the NRC is proprietary, and requests the NRC to withhold
that information under the Freedom of Information Act (FOIA), the NRC
will consider that request under the existing rules governing FOIA
disclosure in 10 CFR 2.309(a)(4).
The NRC has also decided to amend Sec. 52.99(c) which requires the
licensee to notify the NRC that the prescribed inspections, tests, and
analyses in the ITAAC have been or will be completed and that the
acceptance criteria have been met. The NRC is revising Sec.
52.99(c)(1) in the final rule to more closely follow the language of
Section 185b. of the AEA and to clarify that the notification must
contain sufficient information to demonstrate that the prescribed
inspections, tests, and analyses have been performed and that the
prescribed acceptance criteria have been met. The NRC is adding this
clarification to ensure that combined license applicants and holders
are aware that (1) it is the licensee's burden to demonstrate
compliance with the ITAAC and (2) the NRC expects the notification of
ITAAC completion to contain more information than just a simple
statement that the licensee believes the ITAAC has been completed and
the acceptance criteria met. The NRC expects the notification to be
sufficiently complete and detailed for a reasonable person to
understand the bases for the licensee's representation that the
inspections, tests, and analyses have been successfully completed and
the acceptance criteria have been met. The term ``sufficient
information'' requires, at a minimum, a summary description of the
bases for the licensee's conclusion that the inspections, tests, or
analyses have been performed and that the prescribed acceptance
criteria have been met. The NRC plans to prepare regulatory guidance,
in consultation with interested stakeholders, to explain how the
functional requirement to provide ``sufficient information'' with
regard to ITAAC submittals could be met.
The NRC is also revising Sec. 52.99(c) by adding a new paragraph
(c)(2) requiring that, if the licensee has not provided, by the date
225 days before the scheduled date for initial loading of fuel, the
notification required by paragraph (c)(1) of this section for all
ITAAC, then the licensee shall notify the NRC that the prescribed
inspections, tests, or analyses for all uncompleted ITAAC will be
performed and that the prescribed acceptance criteria will be met prior
to operation (consistent with the Section 185.b requirement that the
Commission, ``prior to operation,'' find that the acceptance criteria
in the combined license are met). The notification must be provided no
later than the date 225 days before the scheduled date for initial
loading of fuel. It is the licensee's burden to demonstrate that it
will comply with the ITAAC and it must provide sufficient information
to demonstrate that the prescribed inspections, tests, or analyses will
be performed and the prescribed acceptance criteria for the uncompleted
ITAAC will be met. The term ``sufficient information'' requires, at a
minimum, a summary description of the bases for the licensee's
conclusion that the inspections, tests, or analyses will be performed
and that the prescribed acceptance criteria will be met. In addition,
``sufficient information'' includes, but is not limited to, a
description of the specific procedures and analytical methods to be
used for performing the inspections, tests, and analyses and
determining that the acceptance criteria have been met.
Paragraph (e) has been revised to require that the NRC make
available to
[[Page 49367]]
the public the notifications to be submitted under Sec. 52.99(c)(1)
and (c)(2), no later than the Federal Register notice of intended
operation and opportunity for hearing on ITAAC under Sec. 52.103(a). A
conforming change is included in Sec. 2.105(b)(3) to require that the
Sec. 52.103(a) notice reference the public availability of the Sec.
52.99(c)(1) and (2) notifications. The NRC is requiring that the
paragraph (c)(2) notification be made 225 days before the date
scheduled for initial loading of fuel, in order to ensure that the
licensee notifications are publicly available through the NRC document
room and online through the NRC Web site at the same time that the
Sec. 52.103(a) notice is published in the Federal Register. The NRC's
goal is to publish that notice 210 days before the date scheduled for
fuel loading, but in all cases the Sec. 52.103(a) notice would be
published no later than 180 days before the scheduled fuel load, as
required by Section 189.a(1)(B) of the AEA.
Commenters did not support addition of a requirement on completion
of ITAAC in a set time period prior to fuel load and the NRC has not
included a provision requiring the completion of all ITAAC by a certain
time prior to the licensee's scheduled fuel load date. Instead, the NRC
has decided to modify the concept slightly by requiring the licensee to
submit, with respect to ITAAC which have not yet been completed 225
days before the scheduled date for initial loading of fuel, additional
information addressing whether those inspections, tests, and analyses
will be successfully completed and the acceptance criteria met before
initial operation. In the case where the licensee has not completed all
ITAAC by 225 days prior to its scheduled fuel load date, the NRC
expects the information that the licensee submits related to
uncompleted ITAAC to be sufficiently detailed such that the NRC can
determine what activities it will need to undertake to determine if the
acceptance criteria for each of the uncompleted ITAAC have been met,
once the licensee notifies the NRC that those ITAAC have been
successfully completed and their acceptance criteria met. In addition,
the NRC is adopting the requirements in paragraphs (c)(1) and (c)(2) to
ensure that interested persons will have sufficient information to
address the Atomic Energy Act, Section 189.a(1), threshold for
requesting a hearing with respect to both completed and as-yet
uncompleted ITAAC. The NRC plans to prepare regulatory guidance
providing further explanation of what constitutes ``sufficient
information'' that must be submitted under paragraphs (c)(1) and (c)(2)
demonstrating that the inspections, tests, or analyses for ITAAC have
been or will be completed and the acceptance criteria for the ITAAC
have been or will be met. The NRC expects that any contentions
submitted by prospective parties regarding uncompleted ITAAC would
focus on any inadequacies of the specific procedures and analytical
methods described by the licensee under paragraph (c)(2), in the
context of the findings called for by Sec. 52.103(b)(2).\3\
---------------------------------------------------------------------------
\3\ Inasmuch as the ITAAC themselves have already been approved
by the NRC and their adequacy may not be challenged except under the
provisions of Sec. 52.103(f), a contention which alleges the
deficiency of the ITAAC is not admissible under Sec. 52.103(b).
---------------------------------------------------------------------------
The NRC notes that, even though it did not include a provision
requiring the completion of all ITAAC by a certain time prior to the
licensee's scheduled fuel load date, the NRC will require some period
of time to perform its review of the last ITAAC once the licensee
submits its notification that the ITAAC has been successfully completed
and the acceptance criteria met. In addition, the Commission itself
will require some period of time to perform its review of the staff's
conclusions regarding all of the ITAAC and the staff's recommendations
regarding the Commission finding under Sec. 52.103(g). Therefore,
licensees should structure their construction schedules to take into
account these time periods. The NRC staff intends to develop regulatory
guidance on the licensee's completion and NRC verification of ITAAC and
will provide estimates of the time it expects to take to verify
successful completion of various types of ITAAC. The NRC expects that
such guidance, along with frequent communication with licensees during
construction, will provide licensees with adequate information to plan
initial fuel loading and related activities.
Question 13: ML Hearings. As discussed in Section IV.F.6 of the
March 13, 2006, proposed rule, the Commission proposes, as a matter of
policy and discretion, that the Commission hold a ``mandatory'' hearing
(i.e., a hearing which, under NRC requirements in 10 CFR part 2, is
held regardless of whether the NRC receives any hearing requests or
petitions to intervene) in connection with the initial issuance of
every manufacturing license. The Commission believes that Section
189.a.(1)(A) of the AEA does not require that a hearing be held in
connection with the initial issuance of a manufacturing license.
Nonetheless, there are several reasons for the Commission to require by
rule, as a matter of discretion, a mandatory hearing. A manufacturing
license may be viewed as analogous to a construction permit--a
regulatory approval for which Section 189 of the AEA specifically
requires that a hearing be held. Even though the Commission's
regulations did not address the hearing requirements for manufacturing
licenses, the Commission noticed a ``mandatory'' hearing in connection
with the only manufacturing license application ever received by the
Agency. Offshore Power Systems (Floating Nuclear Power Plants), 38 FR
34008 (December 10, 1973). Accordingly, proposed Sec. Sec. 2.104 and
52.163 require that a mandatory hearing be held in each proceeding for
initial issuance of a manufacturing license. However, the Commission
recognizes that there may be countervailing considerations weighing
against Commission adoption of a rulemaking provision mandating that a
hearing be held in connection with the initial issuance of every
manufacturing license where there has been no stakeholder interest in a
hearing. If there is no stakeholder interest in a hearing, transparency
and public confidence would not appear to be relevant considerations in
favor of holding a mandatory hearing. Considerations of regulatory
efficiency and effectiveness would be paramount, and would weigh
against holding of a mandatory hearing. The Commission requests
comments on whether the Commission should exercise its discretion to
provide by rule an opportunity for hearing, rather than a mandatory
hearing, and the reasons in favor of providing an opportunity for
hearing as opposed to holding a mandatory hearing. Based upon the
public comments, the Commission may adopt a final rule which deletes
Sec. 2.104(f), revises Sec. 2.105 (governing the content of a Federal
Register notice of proposed action where a mandatory hearing is not
held under Sec. 2.104) to add, as appropriate, references to issuance
of manufacturing licenses, and revised Sec. 52.163 to provide an
opportunity for hearing rather than a mandatory hearing in connection
with the initial issuance of a manufacturing license.
Commenters' Response: Several commenters stated there was no need
to require mandatory hearings for manufacturing licenses, or that the
need for such hearings was unclear. These commenters expressed the
belief that such hearings were not an appropriate method for reviewing
and resolving
[[Page 49368]]
technical issues. Some commenters advised that the decision to request
a hearing be left to either the NRC staff or stakeholders.
NRC Response: As stated in the statement of considerations for the
March 13, 2006, proposed rule, the NRC acknowledges that hearings on
initial issuances of manufacturing licenses are not required by the AEA
(71 FR 12814). The NRC also agrees with the general premise of the
commenters that adjudicatory hearings may not be the best approach for
resolving technical design issues--especially in uncontested
proceedings. Indeed, the NRC removed the opportunity for adjudicatory-
style hearings for design certifications as part of the 2004 changes to
10 CFR part 2 (January 14, 2004; 69 FR 2182). The primary
responsibility for determining the safety of an application is with the
NRC staff, and not the presiding officer. This is true regardless of
whether the proceeding is contested or uncontested. Public confidence
would not seem to be enhanced in any significant manner by the holding
of a hearing where there is no request that the NRC hold a hearing.
Accordingly, the NRC has decided not to adopt in the final part 52 rule
a requirement for a ``mandatory'' hearing in connection with issuance
of manufacturing licenses.
Question 14: As discussed in Section IV.C.5.g of the statements of
consideration of the March 13, 2006, proposed rule, the proposed rule
would amend the special backfit requirement in 10 CFR 52.63(a)(1) to
provide the Commission with the ability to make changes to the design
certification rules (DCRs) or the certification information in the
generic design control documents that reduce unnecessary regulatory
burdens. The underlying rationale for this provision also forms the
basis for amending the Tier 2 change process in the three DCRs
(appendices A, B, and C of part 52) to incorporate the revised change
criteria in 10 CFR 50.59.
The Commission is considering adopting an additional provision
[Sec. 52.63(a)(1)(iv)] in the final rule that would allow amendments
of design certification rules to incorporate generic resolutions of
design acceptance criteria (DAC) or other design information without
meeting the special backfit requirement in the current Sec.
52.63(a)(1). The applicants for the current DCRs requested use of DAC
in lieu of providing detailed design information for certain areas of
their nuclear plant designs, for example, instrumentation and control
systems. Under the proposed requirements, a generic change to design
certification information would have to meet the special backfit
requirement of Sec. 52.63(a)(1) or reduce an unnecessary regulatory
burden while maintaining protection to public health and safety and the
common defense and security. The Commission adopted this special
backfit requirement to restrict changes and to require that everyone
meet the same backfit standard for generic changes, thereby ensuring
that all plants built under a referenced DCR would be standardized. By
allowing a DCR amendment to include generic resolutions of DAC or other
design information, the Commission would enhance its goals for design
certification, for example, early resolution of all design issues and
finality for those issue resolutions, which would avoid repetitive
consideration of design issues in individual combined license
proceedings.
There are currently three ways of resolving generic design issues:
(1) the combined license applicant that references a DCR could submit
plant-specific resolutions in its application, which could result in
loss of standardization; (2) a vendor could submit generic resolutions
in topical reports that, if approved, could but would not be required
to be referenced in a combined license application; or (3) the
Commission could exempt itself from the special backfit requirement in
Sec. 52.63(a)(1) and amend the DCR to incorporate a generic
resolution, which could result in multiple rulemakings to revise each
DCR to incorporate each generic resolution. The Commission intends that
any review of a proposed generic resolution would be performed under
the regulations that are applicable and in effect at the time that the
approval or amendment is completed.
Therefore, the NRC is requesting public comments on: (1) whether a
provision should be added to Sec. 52.63(a)(1) to allow generic
amendments to design certification information that meet applicable
regulations in effect at the time that the rulemaking is completed; and
(2) whether the generic resolutions should be incorporated into a DCR
without meeting a backfit requirement, which would provide for
completion of the design certification information and facilitate
standardization, or whether an application for a generic amendment
should be required to meet a backfit requirement (e.g., Sec. 50.109).
Commenters' Response: Some commenters stated that revisions to NRC
regulations should include the current 10 CFR 52.63, which they
believed should allow the original design certification applicant (or
its successor) to obtain amendments to the design certification rule.
These commenters believed current regulations prevented any amendment
to a design once the design has been certified by rule (10 CFR
52.63(a)(1)). Some commenters stated that the design certification
applicant should be able to petition the NRC for, and obtain, an
amendment to the design certification rule to incorporate
``beneficial'' changes to the design certification, including: (1)
Design changes that would result in significant improvements in safety;
(2) design changes that would result in significant improvements in
efficiency, reliability and/or economics; (3) design changes that
result from continuing engineering or design work or are required
because of lack of availability of components specified in the original
design certification; and (4) design changes necessary to correct minor
errors in the original design certification. Some commenters also
suggested that where proposed changes involved changes to Tier 2, the
design certification applicant should be able to make such changes
using a Sec. 50.59-like change process. One commenter noted that
changes to allow an amendment to the final design certification could
potentially simplify COL applications, reduce NRC staff resource
burden, and help assure standardization across the industry.
NRC Response: The NRC has decided to include an amendment process
in the final rule that: (1) Reduces unnecessary regulatory burden and
maintains protection to public health and safety and common defense and
security; (2) provides the detailed design information necessary to
resolve selected design acceptance criteria; (3) corrects material
errors in the certification information; (4) substantially increases
overall safety, reliability, or security of a facility and the costs of
the change are justified; or (5) contributes to increased
standardization of the certification information, without meeting the
special backfit requirement in Sec. 52.63(a)(1)(ii). These amendments
will apply to all plants that have referenced or will reference the
DCR. The NRC believes that these amendments will enhance
standardization by further completing or correcting the certification
information. A detailed discussion of the amendment process is provided
in Section V.C.7.g of the Supplementary Information of this document.
Question 15: In Section IV.J of the supplementary information of
the March 13, 2006, proposed rule, the NRC
[[Page 49369]]
outlines key principles regarding its proposal for reporting
requirements that implement Section 206 of the Energy Reorganization
Act, as amended, for part 52 licenses, certifications, and approvals.
The NRC discusses that the beginning of the ``regulatory life'' of a
referenced license, standard design approval, or standard design
certification under part 52 occurs when an application for a license,
design approval, or design certification is docketed. The NRC also
cautions, however, that this does not mean that an applicant is without
Section 206 responsibilities for pre-application activities because
there are two aspects to the reporting requirements, namely, a
``backward looking'' or retrospective aspect with respect to existing
information, and a ``forward looking'' or prospective aspect with
respect to future information. For an early site permit applicant, the
retrospective obligation is that the early site permit holder and its
contractors, upon issuance of the early site permit, must report all
known defects or failures to comply in ``basic components,'' as defined
in part 21. Under the proposed part 21 requirements presented in the
proposed rule, the early site permit holder and its contractors are
required to meet these requirements upon issuance of the early site
permit. Accordingly, applicants should procure and control safety-
related design and analysis or consulting services in a manner
sufficient to allow the early site permit holder and its contractors to
comply with the above described reporting requirements of Section 206,
as implemented by part 21. A similar argument applies to design
certification applicants. Although the Commission has not proposed an
explicit requirement imposing part 21 on applicants for an early site
permit or design certification in the proposed rule, it is considering
adopting such a requirement in the final part 52 rulemaking because, as
a practical matter, the NRC has to require these applicants to
implement a part 21 program before approval of the early site permit or
design certification. Therefore, providing explicit part 21
requirements for applicants would clarify the Commission's intent. The
Commission requests stakeholder comment on whether it should, in the
final rule, impose part 21 reporting requirements on applicants for
early site permits and design certifications.
Commenters' Response: Several commenters were opposed to the
proposed changes to part 21. Some commenters stated part 21 had been in
existence for almost 30 years, during which it was never applied to
applicants. They complained that they were not aware, and the NRC had
not made them aware, of problems that would warrant a change. The
commenters noted that applicants take measures to ensure that they were
made aware of any errors and deficiencies identified by contractors and
suppliers for work performed on commercial nuclear projects, because
applicants eventually become holders, and licensees and want equipment
to operate correctly. Several commenters were also concerned that the
proposal was contrary to the Energy Reorganization Act (ERA), which was
the basis for part 21. They believed it would be inappropriate and
contrary to the ERA to apply part 21 to applicants. They stated part 21
was established to implement Sec. 206 of the ERA, which applies to
``licensees'' and vendors, suppliers, and contractors of licensees, not
to ``applicants.'' These commenters cited 10 CFR 21.2, stating that the
existing regulations of part 21 apply only to entities licensed to
possess, use, or transfer radioactive material within the United
States, or to construct, manufacture, possess, own, operate, or
transfer within the United States, any production or utilization
facility or fuel storage facility. The commenter believed applicants
did not fall within the scope of Sec. 206 of the ERA, and it was
inconsistent with the Act to expand the scope of Sec. 21.2 to include
applicants.
Some commenters also noted that it had been the standard practice
for a construction permit (CP) applicant to specify part 21
requirements in its procurement contracts for a plant prior to issuance
of the construction permit. Some commenters agreed with this practice
because part 21 was applicable to such contracts once the CP was issued
by the NRC, and expected that this ``good practice'' would be
implemented by COL applicants as well. From a ``practical
perspective,'' the commenters believed this negated the need to expand
part 21 to applicants.
Some commenters argued that the obligations for applicants to
provide information to the NRC under proposed Sec. 52.6(a) was broader
than the obligation in part 21, and would require applicants to update
and correct their applications to account for the types of defects and
noncompliances covered by part 21. These commenters stated the industry
had no objection to proposed Sec. 52.6(a), which should therefore
eliminate the need to apply part 21 to applicants.
NRC Response: The Commission proposed part 21 reporting
requirements on applicants for early site permits, design
certifications, and standard design approvals in the proposed rule. A
detailed discussion on the Commission's rationale for imposing these
requirements in the final rule is provided in Section V.J of the
supplementary information of this document.
V. Discussion of Substantive Changes and Responses to Significant
Comments
A. Introduction
The changes to 10 CFR Chapter I are further discussed by part.
Changes to parts 52 and 50 are discussed first, followed by changes to
other parts in numerical order. Within each part, general topics are
discussed first, followed by discussion of changes to individual
sections as necessary. In addition to the substantive changes, rule
language was revised to make conforming administrative changes (e.g.,
identification of regulations containing information collection
requirements in Sec. 52.11), correct typographic errors, adopt
consistent terminology (e.g., ``makes the finding under Sec.
52.103(g)''), correct grammar, and adopt plain English. These changes
are not discussed further.
B. Testing Requirements for Advanced Reactors
This rule amends Sec. Sec. 50.43, 52.47, 52.79, and 52.157 to
achieve clarity and consistency in the testing requirements for
advanced reactor designs and plants. This amendment requires applicants
for a combined license, operating license, or manufacturing license
that use new safety features but do not reference a certified advanced
reactor design to also perform the design qualification testing
required of certain applicants for design certification. If a combined
license application references a certified design, the necessary
qualification testing will have been performed under Sec. 52.47(c)(2).
The codification of testing requirements in the original Sec. 52.47
was a principal issue during the development of 10 CFR part 52 (see
Section II of 54 FR 15372; April 18, 1989). The requirement to
demonstrate the performance of new safety features for nuclear power
plants that differ significantly from evolutionary light-water reactors
or that use simplified, inherent, passive, or other innovative means to
accomplish their safety functions (advanced reactors), were included in
10 CFR part 52 to ensure that these new safety features will perform as
predicted in the applicant's safety analysis report, to provide
sufficient data to validate analytical codes, and that the effects of
systems
[[Page 49370]]
interactions are acceptable. The design qualification testing
requirements may be met with either separate effects or integral system
tests; prototype tests; or a combination of tests, analyses, and
operating experience. These requirements implement the Commission's
policy on proof-of-performance testing for all advanced reactors and
its goal of resolving all safety issues before authorizing
construction.
Some commenters stated that it is unnecessary to apply
qualification testing requirements to combined license applicants. The
Commission does not agree because, when it reformed the licensing
process for new nuclear plants with the issuance of part 52, the
Commission required applicants to demonstrate that new safety features
will perform as predicted in the final safety analysis report. Although
the focus of the NRC at that time was on applications for design
certification, the Commission intended that testing to qualify new
design features (proof-of-performance testing) would be required for
all advanced reactors, including custom designs (see Question 6 at 51
FR 24 646; July 8, 1986). Furthermore, it would make no sense for the
Commission to require qualification testing for design certification
applicants (so-called paper designs) and not require testing for
applications to build and operate an advanced nuclear power plant.
Therefore, the NRC has implemented its intent in adopting part 52 to
resolve issues early and its policy on advanced reactors that it is
necessary to demonstrate the performance of new or innovative safety
features through design qualification testing for all advanced nuclear
reactor designs or plants (including nuclear reactors manufactured
under a manufacturing license).
This amendment also includes a requirement in Sec. 50.43(e)(2) for
licensing a prototype plant, as defined in Sec. Sec. 50.2 and 52.1, if
the plant is used to meet the testing requirements in Sec.
50.43(e)(1). The new Sec. 50.43(e) states that, if a prototype plant
is used to comply with the qualification testing requirements, the NRC
may impose additional requirements on siting, safety features, or
operational conditions for the prototype plant to compensate for any
uncertainties associated with the performance of the new or innovative
safety features in the prototype plant.
Some commenters stated that it would be inappropriate to establish
or impose prototype testing on combined license applicants. Although
the Commission stated that it favors the use of prototypical
demonstration facilities and that prototype testing is likely to be
required for certification of advanced non-light-water designs (see
Advanced Reactor Policy Statement at 51 FR 24646; July 8, 1986, and the
statement of consideration for 10 CFR part 52, 54 FR 15372; April 18,
1989), this rule does not require the use of a prototype plant for
qualification testing. Rather, this rule provides that if a prototype
plant is used to qualify an advanced reactor design, then additional
conditions may be required for the licensed prototype plant to
compensate for any uncertainties with the unproven safety features.
Also, the prototype plant could be used for commercial operation.
C. Changes to 10 CFR Part 52
1. Use of Terms: Site Characteristics, Site Parameters, Design
Characteristics, and Design Parameters in Sec. Sec. 52.1, 52.17, 52.U0
, 52.39, 52.47, 52.54, 52.79, 52.93, 52.157, 52.158, 52.167, 52.171,
and Appendices A, B, and C to Part 52
The NRC is revising 10 CFR part 52 to clarify the use of the terms,
site characteristics, site parameters, design characteristics, and
design parameters, in order to ensure that the NRC's requirements
governing applications for and issuance of early site permits, design
approvals, design certifications, combined licenses, and manufacturing
licenses are expressed in clear and unambiguous terms. This final rule
adds or revises these terms where necessary to reflect this
clarification. Corresponding changes are made to Sec. Sec. 52.17,
52.24, 52.39, 52.47, 52.54, 52.79, 52.93, 52.157, 52.158, 52.167,
52.171, and Section III.E of appendices A, B, and C to part 52.
The NRC is also adding definitions of the terms design
characteristics, design parameters, site characteristics, and site
parameters to Sec. 52.1 to clarify the use of these terms. Design
characteristics are defined as the actual features of a reactor. Design
characteristics are specified in a standard design approval, a standard
design certification, a combined license application, or a
manufacturing license. Design parameters are defined as the postulated
features of a reactor or reactors that could be built at a proposed
site. Design parameters are specified in an early site permit. Site
characteristics are defined as the actual physical, environmental and
demographic features of a site. Site characteristics are specified in
an early site permit or in a final safety analysis report for a
combined license. Site parameters are defined as the postulated
physical, environmental and demographic features of an assumed site.
Site parameters are specified in a standard design approval, standard
design certification, or a manufacturing license.
In addition, the NRC is revising Sec. 52.79 to include a
requirement that a combined license application referencing a certified
design must contain information sufficient to demonstrate that the
design of the facility falls within the site characteristics and design
parameters specified in the early site permit. Former Sec. 52.79
included a requirement that a combined license application referencing
an early site permit contain information sufficient to demonstrate that
the design of the facility falls within the parameters specified in the
early site permit. The NRC interprets parameters to mean the site
characteristics and design parameters as defined in Sec. 52.1. The NRC
is making similar changes to Sec. Sec. 52.39 and 52.93. The need for
these changes became evident during NRC's review of the pilot early
site permit applications. Because the NRC is relying on certain design
parameters specified in the early site permit applications to reach its
conclusions on site suitability, these design parameters will be
included in any early site permit issued. The NRC believes that these
changes, in the aggregate, will provide sufficient clarification on the
use of the terms in question.
As the NRC completes its review of the first early site permit
applications and prepares for the submittal of the first combined
license application, it is focusing on the interaction among the early
site permit, design certification, and combined license processes. The
NRC believes that its review of a combined license application that
references an early site permit will involve a comparison to ensure
that the actual characteristics of the design chosen by the combined
license applicant fall within the design parameters specified in the
early site permit. NRC review of a combined license application that
references a design certification will involve a comparison to ensure
that the actual characteristics of the site chosen by the combined
license applicant fall within the site parameters in the design
certification. Similarly, if a combined license applicant references
both an early site permit and a design certification, the NRC will
review the application to ensure that the site characteristics in the
early site permit fall within the site parameters in the referenced
design certification and that the actual characteristics of the
certified design fall within the design parameters in the early site
permit. For these
[[Page 49371]]
reasons, the NRC believes it is important to make the changes described
above in order to clarify these terms and their use in part 52
licensing processes.
2. Issuance of Combined and Manufacturing Licenses (Sec. Sec. 52.97
and 52.167)
Current Sec. 50.50 sets forth the NRC's authority to include
conditions and limitations in permits and licenses issued by the NRC
under part 50. Similar language delineating the NRC's authority in this
regard is also set forth in Sec. 52.24 for early site permits, but is
not included in part 52 with respect to either combined licenses or
manufacturing licenses. There are two possible ways of addressing this
omission: Sec. 50.50 could be revised to refer to combined licenses
and manufacturing licenses, or provisions analogous to Sec. 50.50
could be added to the appropriate sections in part 52 for combined
licenses and manufacturing licenses. Inasmuch as the NRC's inclusion of
appropriate conditions in combined licenses is not a technical matter
per se but rather a matter of regulatory authority, the most
appropriate location for this provision appears to be in part 52.
Inclusion of these provisions in appropriate portions of part 52 would
be consistent with the provision applicable to early site permits in
Sec. 52.24. Accordingly, the NRC is adding the language in Sec.
52.97(c) for combined licenses, and Sec. 52.167(b) for manufacturing
licenses, which are analogous to Sec. 50.50.
3. NRC Staff Information Requests
Section 52.47(a)(3) of the 1989 part 52 rulemaking provided that
the NRC staff would advise the design certification applicant on
whether there was any additional information beyond that required to be
submitted by that section, that must be submitted. The March 2006
proposed rule included analogous provisions (Sec. Sec. 52.17(d),
52.79(a)(42), 52.137(a)(27), and 52.157(p)) for each of the other
licensing and regulatory approval processes in part 52. Upon further
consideration in response to a comment on the March 2006 proposed rule,
the Commission has decided that these provisions are redundant to Sec.
2.102(a), which provides the NRC staff with overall authority to
request information to support their review of an application.
Accordingly, Sec. Sec. 52.17(d), 52.79(a)(42), 52.137(a)(27), and
52.157(p) of the proposed rule have not been adopted in the final rule,
and Sec. 52.47(a)(3) is removed from part 52.
4. Changes to a Design Certification, Departures, Variances, Exemptions
External stakeholders have expressed confusion over the years in
public meetings and in written comments submitted under various
circumstances with respect to the meaning of the terms, change to a
design certification, departures, variances, and exemptions. To clarify
the meaning of these terms, the Commission provides the following
explanation of these terms.
a. Change to a Design Certification
A change to a design certification is a generic change to the
design certification information which is approved by the Commission in
a standard design certification rule under subpart B of part 52. In the
four design certifications currently approved by the Commission, the
design certification information which is approved by the Commission is
either ``certified information'' and is designated as ``Tier 1,'' or is
``approved'' and is designated as ``Tier 2.'' The term ``generic,''
means that if the Commission makes a change to the design
certification, Sec. 52.63(a) requires that the change
(``modification'' under Sec. 52.63(a)(3)) be applied to each plant
referencing the design certification rule.
A change to a design certification may be distinguished from a
departure or variance by understanding that a change is generic.
Therefore, a change to a design certification is:
(1) Requested by the original design certification applicant in
accordance with 10 CFR 2.811 (see 10 CFR 2.800(c)), or by any other
member of the public, in a petition for rulemaking under 10 CFR 2.802;
(2) Applies to all past nuclear power reactors (including
manufactured reactors) whose applications have referenced the design
certification, as well as future reactors referencing the design
certification rule; and
(3) Requires the Commission provide an exemption to the applicant,
if the proposed change is inconsistent with the one or more of the
Commission's regulations.
b. Departure
A departure as a plant-specific ``deviation'' from design
information in either a standard design certification or a
manufacturing license. For a design certification, a departure is a
deviation from the certification information which is certified by the
Commission in a standard design certification rule (for the current
four design certification rules in appendices A through D of part 52,
the certification information is ``Tier 1'' information). For a
manufacturing license, a departure is a deviation from any design
information approved in the manufacturing license, including technical
specifications, site parameters and design characteristics, and
interface requirements.\4\ A departure may be distinguished from a
change to a standard design certification rule (i.e., a change to Tier
1 or Tier 2 information in a design certification rule) or a change to
the design approved in a manufacturing license by recalling that a
departure is plant-specific. Therefore, a departure:
---------------------------------------------------------------------------
\4\ As discussed in the section-by-section discussion for Sec.
52.171, a departure requested by a holder of a combined license
referencing a manufactured reactor must be in the form of a license
amendment, but the criteria for determining the request will be the
exemption criteria in Sec. 52.7 even though the departure itself
may not involve an exemption.
---------------------------------------------------------------------------
Concerns certified design information or manufacturing
license information.
Is requested by the applicant/licensee referencing a
design certification or the use of a manufactured reactor.
Applies only to the design of the nuclear power reactor
referencing the design certification or the manufactured reactor for
which a departure is sought by the applicant/licensee.
Requires the applicant/licensee to obtain an exemption
from the referenced design certification if the proposed departure is
inconsistent with one or more of the Commission's regulations. The
exemption would be granted under the provisions of Sec. 52.7 (which
references the same criteria for the granting of exemptions that are
set forth in Sec. 50.12).
c. Variance
A variance is a plant-specific ``deviation'' from one or more of
the site characteristics, design parameters, or terms and conditions of
an early site permit, or from the site safety analysis report. A
variance to an early site permit is analogous to a departure to a
standard design certification, in that it is plant-specific. Therefore,
a variance:
(1) Concerns information addressed in an early site permit;
(2) Is requested by the applicant referencing an early site permit;
(3) Applies only to the construction permit or combined license
referencing the early site permit; and
(4) Requires the applicant to also obtain an exemption from the
Commission's regulations if the proposed variance is inconsistent with
one or more of the Commission's regulations.
[[Page 49372]]
d. Exemption
An exemption is a Commission-granted dispensation from compliance
with one or more of the Commission's rules and regulations which would
otherwise apply to an entity, a license, permit or other approval such
as a standard design certification rule. Exemption from the
requirements in part 26, or from the requirements in any particular
design certification rule would be provided under Sec. 52.7. Exemption
from an underlying technical requirement in part 50 would be provided
under Sec. 50.12. This would be true even in the course of Commission
adoption of a design certification rule. For example, if the design
certification did not, at the time of final rulemaking, comply with a
technical requirement in part 50, the Commission would provide an
exemption to that requirement as part of the final design certification
rulemaking. Moreover, if the nature of the technical requirement is
such that a subsequent applicant referencing the design certification
would need an exemption from compliance with the requirement as applied
to the applicant, then the Commission would include the exemption in
the design certification rule itself.
5. General Provisions
a. Section 52.0, Scope; Applicability of 10 CFR Chapter I Provisions
The Commission is redesignating former Sec. 52.1, Scope, as Sec.
52.0, Scope; applicability of 10 CFR Chapter I provisions, in order to
add additional sections in the General Provisions portion of part 52.
As discussed elsewhere, the Commission has decided general provisions,
common to all substantive parts in 10 CFR Chapter I, should be added to
part 52. To provide enough section numbers, it is necessary to
redesignate former Sec. 52.1 as Sec. 52.0.
Paragraph (a) of Sec. 52.0 is derived from the text of former
Sec. 52.1, but is revised to include standard design approvals and
manufacturing licenses within the scope of part 52, and to remove
references to Section 104.b of Atomic Energy Act of 1954 (AEA), thereby
providing that licenses issued under part 52 are licenses issued under
Section 103 of the AEA. After passage of the 1970 amendments to the
AEA, all licenses for commercial nuclear power plants with construction
permits issued after the date of the amendments were required to be
issued as Section 103 licenses. The NRC interprets the 1970 amendment
as requiring combined licenses under Section 185 to be issued as
Section 103 licenses.\5\ Accordingly, the NRC is revising the scope of
part 52 to limit its applicability to licenses issued under Section 103
of the AEA.
---------------------------------------------------------------------------
\5\ This may be an academic distinction, in light of the Energy
Policy Act of 2005, Pub. L. No. 109-58, which removed the need for
antitrust reviews of new utilization facilities.
---------------------------------------------------------------------------
Paragraph (b) of Sec. 52.0 is a new provision that makes clear
that the regulations in 10 CFR Chapter I apply to a holder of, or
applicant for an approval, certification, permit, or license issued
under part 52 and that any license, approval, certification, or permit,
issued under 10 CFR part 52 must comply with these regulations. The
need for this paragraph was determined as a result of the July 3, 2003
(68 FR 40026) proposed rule on part 52. In that proposed rule, the
Commission proposed a new Sec. 52.5 listing all of the licensing
provisions in 10 CFR part 50 that also apply to all of the licensing
processes in 10 CFR part 52. This proposal responded to a letter dated
November 13, 2001, from the Nuclear Energy Institute (NEI), which
stated:
The industry proposes that additional General Provisions be
added to Part 52 in addition to an appropriate provision on Written
Communications. This approach is preferable to including cross-
references in Part 52 to Part 50 general provisions because these
provisions typically must be tailored to apply appropriately to the
variety of licensing processes in Part 52.
Section 52.5, as proposed in 2003, would have clarified that the
general provisions in 10 CFR part 50 were also applicable to the new
licensing processes for early site permits, standard design
certifications, and combined licenses in part 52 (as well as the
licensing and approval processes in appendices M, N, O, and Q which
were added to part 52 by the 1989 part 52 rulemaking). Although the
general provisions in part 50 did not specifically refer to the
additional licensing processes in 10 CFR part 52 (and no changes to the
language of those general provisions was proposed), the Commission
believed that proposed Sec. 52.5 would make clear that a holder of, or
applicant for an approval, certification, permit, or license issued
under part 52 must also comply with those general provisions.
However, few commenters on the July 2003 proposed rule believed
that the proposed Sec. 52.5 would provide greater clarity. On the
contrary, some commenters indicated that Sec. 52.5 was overly broad
and would impose burdensome and seemingly inappropriate new
requirements on applicants for design certifications that were
unwarranted.
Accordingly, in the March 2006 proposed rule, the Commission
proposed a different approach, viz., making conforming changes to all
of the regulations in 10 CFR Chapter I to specify their applicability
to the relevant part 52 regulatory processes, and to add proposed Sec.
52.0(b) to make clear that the regulations in 10 CFR Chapter I apply to
the relevant part 52 regulatory processes, and holders and applicants
under part 52. The Commission did not receive any comments calling into
question the legality of this approach, or otherwise questioning the
clarity of the proposed regulatory language. Accordingly, the
Commission is adopting this approach in the final part 52, including
Sec. 52.0(b).
As discussed elsewhere in this document, the NRC is retaining
appendix N in part 52, and revising this appendix to apply to part 52
combined licenses. The provisions of appendix N to part 52 concern
applicants for combined licenses under part 52. Therefore, the
applicability language in Sec. 52.0, by referring to ``licenses''
under part 52, need not specifically refer to appendix N to part 52.
b. Section 52.1, Definitions
Section 52.1 (formerly, Sec. 52.3) is revised by adding
definitions for decommission, license, licensee, major feature of the
emergency plans, manufacturing license, modular design, prototype
plant, and standard design approval. A definition of decommission,
which is identical to that in 10 CFR part 50, is added to part 52
because the final part 52 rulemaking addresses decommissioning of
nuclear power reactors with combined licenses under part 52.
Definitions of license and licensee are added to facilitate the use of
these terms throughout part 52. These definitions were derived from the
definitions in Sec. 2.4, but were modified to reflect the regulatory
processes in part 52. The definitions of these terms in part 2 are
modified to be consistent with the definitions in part 52, and the
definitions of these terms are added in part 50, to ensure consistency
among parts 2, 50, and 52. Definitions of manufacturing license and
standard design approval are added to part 52 so that each of these
part 52 license types are defined.
A definition of modular design is added to explain the type of
modular reactor design which is the subject of the second sentence of
Sec. 52.103(g). That provision is added to part 52 to facilitate the
licensing of nuclear plants, such as the Modular High Temperature Gas-
Cooled Reactor (MHTGR) and Power Reactor Innovative Small Module
[[Page 49373]]
(PRISM) designs, consisting of three or four nuclear reactors in a
single power block with a shared power conversion system. During the
period that the power block is under construction, the NRC could
separately authorize operation for each nuclear reactor when each
reactor and all of its necessary support systems were completed. In
view of the several definitions of ``modular reactor'' which are used
within the nuclear industry, the Commission intends to avoid future
disputes regarding the intended applicability of Sec. 52.103(g) by
defining the term, modular design, for purposes of part 52.
The definition of major feature of the emergency plans is being
added in the final rule, based on commenters' responses to Question 2
in Section V of the Supplementary Information of the 2006 proposed
rule, to clarify what is meant by this term as it is used in Sec. Sec.
52.17, 52.18, 52.39, and 52.79. The definition states that a major
feature of the emergency plans means an aspect of those plans necessary
to: (1) address in whole or part, one or more of the sixteen standards
in Sec. 50.47(b), or (2) describe the emergency planning zones as
required in Sec. 50.33(g). The goal of the ``major features'' option
in Sec. 52.17(b) is an NRC finding that the proposed major features
are acceptable as elements of a complete and integrated emergency plan
that would be considered later, when the early site permit is
referenced in a license application. This is not the same level of
finality as the ``reasonable assurance'' finding that would be made in
connection with the approval of a completed and integrated plan.
However, the NRC would not re-review, at the COL stage, information
that provided the basis for the NRC approval of major features in an
ESP but would address integration of approved major features with the
balance of emergency planning information provided in the COL
applications necessary to support the NRC's reasonable assurance
finding; and updated emergency planning information required by Sec.
52.39(b).
A definition of prototype plant is added to explain the type of
nuclear power plant that the NRC is addressing in Sec. Sec. 52.43,
52.47(b), 52.79, and 52.157. A prototype plant is a licensed nuclear
reactor test facility that is similar to and representative of either
the first-of-a-kind or standard nuclear plant design in all features
and size, but may have additional safety features. The purpose of the
prototype plant is to perform testing of new or innovative safety
features for the first-of-a-kind nuclear plant design, as well as being
used as a commercial nuclear power facility.
c. Section 52.2, Interpretations; and Sec. 52.4, Deliberate Misconduct
The former section on interpretations in Sec. 52.5 is retained and
redesignated without change as Sec. 52.2. The former section on
deliberate misconduct in Sec. 52.9 is retained and redesignated
without change as Sec. 52.4.
d. Section 52.3, Written Communications; Sec. 52.5, Employee
Protection; Sec. 52.6, Completeness and Accuracy of Information; Sec.
52.7, Specific Exemptions; Sec. 52.8, Combining Licenses; Sec. 52.9,
Jurisdictional Limits; and Sec. 52.10, Attacks and Destructive Acts
Section 52.3, Written communications, which is essentially
identical with the current Sec. 50.4, is added to address the
requirements for correspondence, reports, applications, and other
written communications from applicants, licensees, or holders of a
standard design approval to the NRC concerning the regulations in part
52.
Section 52.5, which is largely identical with the current Sec.
50.7, is added to make clear that discrimination against an employee
for engaging in certain protected activities concerning the regulations
in part 52 is prohibited. This section differs from its part 50
counterpart, in that the Commission has added a provision on
coordination with the requirements in 10 CFR part 19.
Section 52.6, which is identical with the current Sec. 50.9, is
added to require that information provided to the Commission by a
licensee, a holder of a standard design approval, and an applicant
under part 52, and information required by statute or by the NRC's
regulations, orders, or license conditions to be maintained by a
licensee, holder of a standard design approval, and applicant under
part 52 (including the applicant for a standard design certification
under part 52 following Commission adoption of a final design
certification rule) be complete and accurate in all material respects.
The Commission has corrected an error in the proposed rule version of
paragraph (a) of Sec. 52.6. In the proposed rule, the first sentence
began, ``Information provided to the Commission by a licensee
(including a construction permit holder, and a combined license holder)
* * *.'' In the final rule, this phrase has been corrected to read,
``Information provided to the Commission by a licensee (including an
early site permit holder, a combined license holder, and a
manufacturing license holder) * * *.'' This provision applies to
licenses issued under part 52 and not to licenses issued under part 50.
Section 52.7, which is essentially identical with current Sec.
50.12, is added to address the procedure and criteria for obtaining an
exemption from the requirements of part 52. Although part 50 contains a
provision (Sec. 50.12) for obtaining specific exemptions, Sec. 50.12
by its terms applies only to exemptions from part 50. Although it would
be possible to revise Sec. 50.12 so that its provisions apply to
exemptions from part 52, this is inconsistent with the general
regulatory structure of 10 CFR, wherein each part is treated as a
separate and independent regulatory unit. The NRC notes that the
exemption provisions in Sec. 52.7 are generally applicable to part 52,
and do not supercede or otherwise diminish more specific exemption
provisions that are in part 52.
Section 52.8, which combines into a single section regulatory
provisions which are addressed in separate regulations in part 50, is
added to clarify that these regulatory provisions also apply to part 52
licenses.
Paragraph (a) of Sec. 52.8, which is analogous to Sec. 50.31, is
added to make clear that an applicant for a license under part 52 may
combine in one application, several applications for different kinds of
licenses under various regulations in 10 CFR Chapter I. Section 50.31
currently provides that an applicant may combine in one application,
several applications for different kinds of licenses under various
regulations in 10 CFR Chapter I. The plain reading of this language,
given that this provision is located in part 50, is that a part 50
application may contain in one application other applications for
different licenses in other parts of 10 CFR Chapter I. Thus, Sec.
50.31 would not appear to allow a part 52 application (as for a
combined license) to combine in one application other applications for
different license in other parts of 10 CFR Chapter I. Accordingly,
paragraph (a) of Sec. 52.8 of the final rule makes clear that a part
52 application may be combined with applications for different licenses
in other parts of 10 CFR Chapter I. This provision was not included in
the March 2006 proposed rule, inasmuch as the NRC determined the
desirability of including in part 52 a provision analogous to Sec.
50.31 only after the publication of the March 2006 proposed rule.
Paragraph (b) of Sec. 52.8, which is analogous to Sec. 50.32, is
added to make clear that an applicant for a license, standard design
certification, or design approval under part 52 may incorporate by
reference in its application information contained in other documents
provided to the Commission,
[[Page 49374]]
but must clearly specify the information to be incorporated. This
provision was also not included in the March 2006 proposed rule,
inasmuch as the NRC determined the desirability of including in part 52
a provision analogous to Sec. 50.32 only after the publication of the
March 2006 proposed rule.
Paragraph (c) of Sec. 52.8, which is analogous to Sec. 50.52, is
added to clarify the Commission's authority under Section 161.h of the
AEA to combine NRC licenses, such as a special nuclear materials
license under part 70 for the reactor fuel, with a combined license
under part 52. Analogous to the situation with respect to Sec. 50.31,
the language in Sec. 50.52 would not appear to allow the Commission to
combine into a single part 52 license, other non-part 52 licenses.
Inasmuch as these changes to Sec. 52.8 constitute revisions to the
Commission's rules of procedure and practice, the Commission may adopt
them in final form without further notice and comment, under the
rulemaking provisions of the APA, 5 U.S.C. 553(b)(A).
Section 52.9, which is identical with Sec. 50.53, is added to
clarify that NRC licenses issued under part 52 do not authorize
activities which are not under or within the jurisdiction of the United
States; an example would be the construction of a nuclear power reactor
outside the territorial jurisdiction of the United States which uses a
design identical to that approved in a standard design certification
rule in part 52.
Section 52.10 is added because there is no specific provision in
part 52 specifying that the Commission's longstanding determination
with respect to the lack of need for design features and other measures
for protection of nuclear power plants against attacks by enemies of
the United States, or the use of weapons deployed by United States
defense activities, applies to part 52 applicants. The Commission's
determination, which was upheld by the U.S. Court of Appeals for the
D.C. Circuit, see Siegel v. Atomic Energy Commission, 400 F.2d 778
(D.C. Cir 1968), is currently codified for part 50 applicants in Sec.
50.13. Although it would be possible to revise Sec. 50.13 so that its
provisions apply to applications under part 52, this would be
inconsistent with the overall regulatory pattern of 10 CFR Chapter I,
whereby each part is treated as a separate and independent regulatory
unit. Moreover, any changes to Sec. 50.13 might erroneously be viewed
as changes to the Commission's substantive determination on this
matter. For these reasons, the Commission is adding new Sec. 52.10 to
part 52, which is essentially identical with Sec. 50.13. Inclusion of
this provision in part 52 makes clear that applications for combined
licenses, manufacturing licenses, design certification rulemakings,
standard design approvals, and amendments to these licenses,
rulemakings, and approvals under part 52 need not provide design
features or other measures for protection of nuclear power plants
against attacks by enemies of the United States, or the use of weapons
deployed by U.S. defense activities. In adding Sec. 52.10, the
Commission emphasizes that it is not changing in any way, nor is it
intending to revisit in this rulemaking, the Commission's determination
with respect to the lack of need for design features or other measures
for protection of nuclear power plants against attacks by enemies of
the United States, or the use of weapons deployed by U.S. defense
activities. The Commission is simply making it clear that its
longstanding determination applies to applications under part 52 just
as it applies to applications under part 50.
6. Subpart A, Early Site Permits
a. Emergency Preparedness Requirements for Early Site Permit Applicants
The NRC is amending Sec. Sec. 52.17(b), 52.18, and 52.39 to
address changes to emergency preparedness requirements for early site
permit applicants. The NRC is amending Sec. 52.17(b)(1), which
requires that an early site permit application identify physical
characteristics unique to the proposed site that could pose a
significant impediment to the development of emergency plans. The NRC
is adding a sentence to require that, if physical characteristics that
could pose a significant impediment to the development of emergency
plans are identified, the application must identify measures that
would, when implemented, mitigate or eliminate the significant
impediment. The NRC believes this addition is necessary to clarify the
NRC's expectations in cases where a physical characteristic exists that
could pose a significant impediment to the development of emergency
plans. Simply identifying these physical characteristics alone does not
provide the NRC with enough information to determine if these
characteristics are likely to pose a significant impediment to the
development of emergency plans. Similarly, the Commission is amending
Sec. 52.18 to require that the Commission determine whether the
information required of the applicant by Sec. 52.17(b)(1) shows that
there is no significant impediment to the development of emergency
plans that cannot be mitigated or eliminated by measures proposed by
the applicant [emphasis added].
The NRC is amending Sec. Sec. 52.17(b)(2)(i), 52.17(b)(2)(ii), and
52.18 to clarify that any emergency plans or major features of
emergency plans proposed by early site permit applicants must be in
accordance with the applicable standards of 10 CFR 50.47 and the
requirements of appendix E to part 50. These changes clarify the
standards applicable to emergency preparedness information supplied
with an early site permit application. The NRC is also amending
Sec. Sec. 52.17(b)(1), (b)(2), and (b)(4) to indicate that the
emergency preparedness information supplied in the early site permit
application must be included in the site safety analysis report. This
change is necessary for consistency with past practice and with the
requirements for combined license applicants in Sec. 52.79(a) that
require emergency preparedness information to be included in the final
safety analysis report. Note that the proposed rule only included these
changes in Sec. 52.17(b)(2). In the final rule, the NRC is making the
additional conforming changes in Sec. Sec. 52.17(b)(1) and (b)(4).
The NRC is adding new Sec. 52.17(b)(3) to require that any
complete and integrated emergency plans submitted for review in an
early site permit application must include the proposed inspections,
tests, and analyses that the holder of a combined license referencing
the early site permit shall perform, and the acceptance criteria that
are necessary and sufficient to provide reasonable assurance that, if
the inspections, tests, and analyses are performed and the acceptance
criteria met, the facility has been constructed and would operate in
conformity with the license, the provisions of the AEA, and the NRC's
regulations. The NRC is making these amendments for consistency with
the requirements in subpart C of part 52 regarding the review of
emergency plans and to provide additional finality to ESP holders. The
NRC believes that its review of complete and integrated plans included
in an early site permit application should be no different than its
review of emergency plans submitted in a combined license application,
given that the NRC must make the same findings in both cases, namely,
that the plans submitted by the applicant provide reasonable assurance
that adequate protective measures can and will be taken in the event of
a radiological emergency. The NRC will
[[Page 49375]]
not be able to make the required finding without the inclusion of
proposed ITAAC in an early site permit application that includes
complete and integrated emergency plans. In the final rule, the NRC has
added an allowance that major features of an emergency plan submitted
under paragraph (b)(2)(i) of Sec. 52.17 may include proposed ITAAC.
This will give an applicant that has proposed major features additional
opportunities to achieve finality on major features in cases where
ITAAC can be included to address implementation aspects of the major
feature.
b. Section 52.13, Relationship to Other Subparts
The title of Sec. 52.13 is revised from ``Relationship to subpart
F of 10 CFR part 2 and appendix Q of this part,'' to ``Relationship to
other subparts,'' to reflect the revised scope of this section, which
has been refocused on part 52.
c. Section 52.16, Contents of Applications; General Information and
Sec. 52.17, Contents of Applications; Technical Information
The NRC is adding Sec. 52.16 to include the general content
requirements from Sec. 52.17(a)(1).
The title of Sec. 52.17 is revised to read, ``Contents of
applications; technical information.'' In response to several comments
on the proposed rule, the NRC is including a general grandfathering
provision in Sec. 52.17(a) that states, ``For applications submitted
before September 27, 2007, the rule provisions in effect at the date of
docketing apply unless otherwise requested by the applicant in
writing.'' This revision reflects the Commission's belief that ESPs
currently under review or issued prior to the effective date of the
final part 52 rule should not be required to be modified by this rule.
Section 52.17(a)(1) is amended to state that the early site permit
application must specify the range of facilities for which the
applicant is requesting site approval (e.g., one, two, or three
pressurized-water reactors). This new language provides a clearer and
more complete statement of the applicant's proposal with respect to the
facilities which may be located under the early site permit. This
facilitates NRC review, as well as providing adequate notice to
potentially-affected members of the public and State and local
governmental entities. The NRC assumes that an applicant for an early
site permit may not know what type of nuclear plant may be built at the
site. Therefore, the application must specify the postulated design
parameters for the range of reactor types, the numbers of reactors,
etc., to increase the likelihood that approval of the site will resolve
issues with respect to the actual plant or plants that the combined
license or construction permit applicant decides to build. In a letter
dated November 13, 2001 (comment 27 on draft proposed rule text), NEI
stated, ``The proposed change is too limited. To address the required
assessment of major SSCs [structures, systems, and components] that
bear on radiological consequences and all items 52.17.a.1.i-vii (sic.),
industry recommends new Sec. 52.17a.2.'' The NRC disagrees with NEI's
proposal to have a separate provision for applicants who have not
determined the type of plant that they plan to build at the proposed
site. The NRC expects that some applicants for an early site permit may
not have decided on a particular type of nuclear power plant,
therefore, Sec. 52.17(a)(1) was revised to address this situation.
The NRC is amending Sec. 52.17(a)(1) to eliminate all references
to Sec. 50.34. The references to Sec. 50.34(a)(12) and (b)(10) are
removed because these provisions require compliance with the earthquake
engineering criteria in appendix S to part 50 and are not requirements
for the content of an application. The reference to Sec.
50.34(b)(6)(v), which requires plans for coping with emergencies, is
also being removed. All requirements related to emergency planning for
early site permits are addressed in Sec. 52.17(b) and other plans for
coping with emergencies will be addressed in a combined license
application. Finally, the reference to the radiological consequence
evaluation factors identified in Sec. 50.34(a)(1) is being removed and
the requirements are included in Sec. 52.17(a)(1). The NRC is
modifying the existing requirement for early site permit applications
to describe the seismic, meteorological, hydrologic, and geologic
characteristics of the proposed site to add that these descriptions
must reflect appropriate consideration of the most severe of the
natural phenomena that have been historically reported for the site and
surrounding area and with sufficient margin for the limited accuracy,
quantity, and time in which the historical data have been accumulated.
This addition is to ensure that future plants built at the site would
be in compliance with general design criterion 2 from appendix A to
part 50 which requires that structures, systems, and components
important to safety be designed to withstand the effects of natural
phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami,
and seiches without loss of capability to perform their safety
functions. The design bases for these structures, systems, and
components are required to reflect appropriate consideration of the
most severe of the natural phenomena that have been historically
reported for the site and surrounding area, with sufficient margin for
the limited accuracy, quantity, and time in which the historical data
have been accumulated.
The NRC is adding several requirements to Sec. 52.17(a)(1). A
requirement is added to Sec. 52.17(a)(1)(x) that applications for
early site permits include information to demonstrate that adequate
security plans and measures can be developed. This requirement is
inherent in current Sec. 52.17(a)(1) which states that site
characteristics must comply with 10 CFR part 100. Section 100.21(f)
states that site characteristics must be such that adequate security
plans and measures can be developed. A new Sec. 52.17(a)(1)(xi) is
added to require early site permit applications to include a
description of the quality assurance program applied to site activities
related to the future design, fabrication, construction, and testing of
the structures, systems, and components of a facility or facilities
that may be constructed on the site. This change was made for
consistency with changes to Sec. 50.55 and appendix B to part 50. A
discussion of these changes can be found in this section under the
heading ``Appendix B to Part 50.''
An additional requirement is added to Sec. 52.17(a)(1) that is
taken from Sec. 50.34(h), and that the NRC believes should be
applicable to early site permits. Section 52.17(a)(1)(xii) requires
that early site permit applications include an evaluation of the site
against the applicable sections of the standard review plan (SRP)
revision in effect 6 months before the docket date of the application.
The SRP requirement currently exists for applicants for construction
permits, operating licenses, and combined licenses. The NRC also
believes it should be applicable to applicants for early site permits
because they are partial construction permits that can be referenced in
applications for construction permits or combined licenses and because
it will facilitate the NRC's review of the early site permit
application.
The NRC is not requiring applicants to evaluate their site against
the applicable sections of Regulatory Guide (RG) 1.206, ``Combined
License Applications for Nuclear Power Plants.'' However, the NRC
believes that the applicable portions of RG 1.206 can provide useful
guidance to ESP applicants in preparing their
[[Page 49376]]
applications and that use of this guidance will facilitate the NRC's
review.
The NRC is making a change to Sec. 52.17(a)(1) based on several
comments on the proposed rule. The NRC is deleting the requirement in
proposed Sec. 52.17(a)(1)(x) that required ESP applicants to address
impacts on operating units of constructing new units on existing sites,
as well as include a description of the managerial and administrative
controls to be used to assure that the limiting conditions of operation
for existing units will not be exceeded. The NRC is deleting this
requirement because it was contrary to the industry-NRC understanding
documented in correspondence in 2003 regarding ESP Topic ESP-19 [see
NEI letter dated May 14, 2003 (ML031920U0 6), and NRC letter dated
August 11, 2003 (ML031490478)] and because the COL applicant is in the
best position to provide such information, since it will have final
information regarding the facility design and construction plans. The
NRC may include a condition in early site permits that would require
the permit holder to notify the operating plant licensee prior to
conducting any activities authorized under Sec. 52.25. These controls
should be sufficient to evaluate construction activities at a site with
an existing operating unit. The NRC has deleted this provision from
subpart A in the final rule. COL applicants will, however, continue to
be required to meet this provision under Sec. 52.79(a)(31).
The NRC is moving the environmental provisions in former Sec.
52.17(a)(2) to Sec. 51.50(b). Revised Sec. 52.17(a)(2) simply states
that an early site permit application must contain a complete
environmental report as required by 10 CFR 51.50(b). A discussion of
the final rule provisions related to the NRC's environmental review at
the ESP stage can be found in the Supplementary Information section
that discusses changes to 10 CFR part 51.
The NRC is amending Sec. 52.21 to reflect clarifications provided
in part 51 that an early site permit applicant has the flexibility of
either addressing the matter of alternative energy sources in the
environmental report supporting its early site permit application, or
deferring consideration of alternative energy sources to the time that
the early site permit is referenced in a licensing application. These
changes to Sec. 52.21 clarify that the NRC's EIS need not address the
need for power or alternative energy sources (and therefore these
matters may not be litigated) if the early site permit applicant
chooses not to address these matters in its environmental report.
The NRC is amending Sec. 52.17(c) to clarify that if the applicant
wants to request authorization to perform limited work activities at
the site after receipt of the early site permit, the application must
contain an identification and description of the specific activities
that the applicant seeks authorization to perform. This request by the
early site permit applicant would be separate from, but not in addition
to, a request to perform activities under 10 CFR 50.10(e)(1). The
submittal of this descriptive information will enable the NRC staff to
perform its review of the request, consistent with past practice, to
determine if the requested activities are acceptable under Sec.
50.10(e)(1). If an applicant for a construction permit or combined
license references an early site permit with authorization to perform
limited work activities at the site and subsequently decides to request
authorization to perform activities beyond those authorized under Sec.
52.U0 (c), those additional activities will have to be requested
separately under Sec. 50.10(e)(1). Some minor changes were made to the
rule language in Sec. 52.17(c) in the final rule to remove references
to information being included in either the site safety analysis report
or the environmental report. The NRC concluded that it is preferable to
include both the list of proposed activities and the redress plan as a
separate document in the application, outside of both the site safety
analysis report and the environmental report. The NRC's conclusion is
based on the fact that the requirements in Sec. 50.10(e) address both
safety and environmental issues. Additional changes were made to
Sec. Sec. 51.50, 52.79(a), and 52.80 to implement this concept.
d. Section 52.24, Issuance of Early Site Permit
The NRC is revising Sec. 52.24 to clarify the information that the
NRC must include in the early site permit when it is issued. Section
52.24 is also being amended to be more consistent with the parallel
provision in Sec. 50.50, Issuance of licenses and construction
permits, by requiring the NRC to ensure that there is reasonable
assurance that the site is in conformity with the provisions of the
AEA, and the NRC's regulations; that the applicant is technically
qualified to engage in any activities authorized; and that issuance of
the permit will not be inimical to the common defense and security or
to the health and safety of the public.
Section 52.24 is being amended to provide that the early site
permit must state the site characteristics and design parameters, as
well as the ``terms and conditions,'' of the early site permit, rather
than the ``conditions and limitations'' as was formerly provided. The
change provides consistency with Sec. 52.39(a)(2), and in particular
Sec. 52.39(a)(2)(iii) of the former regulations, which also refers to
``site parameters'' (corrected to ``site characteristics'' in the final
rule) and ``terms and conditions.'' Section 52.24(c) is being added to
require that the early site permit state the activities that the permit
holder is authorized to perform at the site. This change is consistent
with the revision to Sec. 52.17(c) where the applicant must specify
the activities that it is requesting authorization to perform at the
site under Sec. 50.10(e)(1).
The NRC is revising paragraph (b) of this section based on public
comments. Paragraph (b) states that the early site permit shall specify
the site characteristics, design parameters, and terms and conditions
of the early site permit the NRC deems appropriate. Paragraph (b)
further states that, before issuance of either a construction permit or
combined license referencing an early site permit, the Commission shall
find that any relevant terms and conditions of the early site permit
have been met. The NRC is revising this paragraph to add a provision
that any terms or conditions of the early site permit that could not be
met by the time of issuance of the construction permit or combined
license, must be set forth as terms or conditions of the construction
permit or combined license. This provision is needed to address terms
or conditions of the early site permit that are related to activities
that will not take place until after issuance of the construction
permit or combined license, such as construction activities. A similar
change is being made to Sec. 52.79(b)(3).
e. Section 52.27, Duration of Permit
Section 52.27 provides for the duration of an early site permit.
The NRC did not propose any changes to this section in the proposed
rule. However, in the final rule, the NRC is making several revisions.
First, the NRC is revising former Sec. 52.27(b)(1) [final Sec.
52.27(b)]. This paragraph states that an early site permit continues to
be valid beyond the date of expiration in any proceeding on a
construction permit application or a combined license application that
references the early site permit and is docketed before the date of
expiration of the early site permit, or, if a timely application for
renewal of the permit has been filed, before the Commission has
determined whether to
[[Page 49377]]
renew the permit, consistent with the ``Timely Renewal'' doctrine of
the Administrative Procedure Act. This section is changed in the final
rule by deleting the term, ``filing,'' and substituting the term,
``docketing.'' The NRC believes that timely renewal protection should
only be provided to those applications which are of sufficient quality
to be docketed. This is consistent with the requirement in Sec.
2.109(b) requiring filing of a ``sufficient'' application for renewal
of operating licenses as a prerequisite for the applicability of the
timely renewal protection. Inasmuch as the changes to former Sec.
52.72(b)(1) constitute revisions to the NRC's rules of procedure and
practice, the NRC may adopt them in final form without further notice
and comment, under the rulemaking provisions of the APA, 5 U.S.C.
553(b)(A).
The NRC is also making revisions to Sec. 52.27 based on public
comments. The NRC is deleting proposed Sec. 52.27(b)(2) because it was
inconsistent with proposed Sec. 52.39(d) and the NRC's intention that
the early site permit be subsumed into the construction permit or
combined license once the construction permit or combined license is
issued. To make this intention clear, the NRC is also adding new Sec.
52.27(d) in the final rule. This provision states that upon issuance of
a construction permit or combined license, a referenced early site
permit is subsumed, to the extent referenced, into the construction
permit or combined license. By ``subsumed'' the NRC means that the
information that was contained in the early site permit site safety
analysis report (SSAR) becomes part of the referencing combined license
final safety analysis report upon issuance of the combined license in
the same manner as if the combined license applicant had not referenced
an early site permit. The NRC is including the phrase ``to the extent
referenced,'' to indicate that it is not all of the information
submitted in the early site permit application that is subsumed into
the combined license, but, only that information that is contained in
the SSAR and identified by the applicant as being referenced in the
combined license application. This subsumption of the early site permit
into the referencing license affects the way changes to the early site
permit information will be handled because it breaks the tie to the
finality provisions in Sec. 52.39. After issuance of the construction
permit or combined license, Sec. 52.39 no longer applies to the early
site permit information and such information will be covered by the
same finality provisions as the rest of the information in the FSAR
(with the exception of any referenced design certification
information), as outlined in Sec. 52.98 (e.g., in accordance with
Sec. Sec. 50.54, 50.59, etc.).
f. Section 52.28, Transfer of Early Site Permit
Section 52.28 is being added to state that transfer of an early
site permit from its existing holder to a new applicant would be
processed under Sec. 50.80, which contains provisions for transfer of
licenses. In a letter dated November 13, 2001 (comment 19 on draft
proposed rule text), the NEI recommended that a new section be added to
part 52 to clarify the process for transfer of an early site permit.
The NRC has determined that a new section is not necessary because an
early site permit is a partial construction permit and, therefore, is
considered to be a license under the AEA. The NRC believes that the
procedures and criteria for transfer of utilization facility licenses
in 10 CFR 50.80 (and the procedures in subpart M of part 2 for the
conduct of any hearing) should apply to the transfer of an early site
permit. Changes that the NRC has made to Sec. 50.80 in the final rule
to address comments made regarding requirements for transfer of an
early site permit can be found in Section V.D.8.a of the supplementary
information of this document.
g. Section 52.33, Duration of Renewal
Section 52.33 has been revised in the final rule to clarify that
the renewal period for an early site permit includes any remaining
years on the early site permit then in effect before renewal. This
change was made to be consistent with the NRC's regulations concerning
renewal of nuclear power plant operating licenses as specified in Sec.
54.31 of this chapter.
h. Section 52.37, Reporting of Defects and Noncompliance; Revocation,
Suspension, Modification of Permits for Cause
Section 52.37 is removed because this provision only contains a
cross-reference to 10 CFR part 21 and Sec. 50.100, and the NRC is
making conforming changes to those requirements to account for
requirements for early site permits.
i. Section 52.39, Finality of Early Site Permit Determinations
The NRC is revising Sec. 52.39 to address the finality of an early
site permit. While some of the changes are conforming or clarifying,
others represent a change from the finality provisions in the former
Sec. 52.39. Paragraph (a)(2) of the former rule distinguishes among
issues alleging that: (1) a ``reactor does not fit within one or more
of the site parameters,'' which are to be treated as valid contentions
(paragraph (a)(2)(i)); (2) a ``site is not in compliance with the terms
of an early site permit,'' which are to be subject to hearings under
the provisions of the Administrative Procedure Act (paragraph
(a)(2)(ii)); and (3) the ``terms and conditions of an early site permit
should be modified,'' which are to be processed in accordance with 10
CFR 2.206(a)(2)(iii). With the benefit of hindsight and experience
gained in reviewing the first three early site permit applications, the
NRC believes that all issues concerning a referenced early site permit
may be characterized as:
(1) Questions regarding whether the site characteristics, design
parameters, or terms and conditions specified in the early site permit
have been met;
(2) Questions regarding whether the early site permit should be
modified, suspended, or revoked; or
(3) Significant new emergency preparedness or environmental
information not considered on the early site permit.
Questions about the referencing application demonstrating
compliance with the early site permit are fundamentally questions of
compliance with the early site permit. They do not attack the
underlying validity of the permit. For example, if a person questions
whether the design characteristics of the nuclear power facility that
the referencing applicant proposes to construct on the site falls
within the design parameters specified in the early site permit, it is
a matter of compliance with the early site permit. These compliance
matters are specific to the proceeding for the referencing application,
and the NRC concludes that a question about whether the referencing
application complies with the early site permit may be viewed as
question/material to the proceeding and appropriate for consideration
in the referencing application proceeding (assuming that all relevant
Commission requirements in 10 CFR part 2, such as standing and
admissibility, are met).
The NRC also regards new emergency preparedness information
submitted in the referencing application that substantially alters the
bases for a previous NRC conclusion or constitutes a sufficient basis
for the Commission to modify or impose new terms and conditions related
to emergency preparedness as an issue material to the
[[Page 49378]]
proceeding and appropriate for consideration as a contention in the
referencing application proceeding (assuming that all relevant
Commission requirements in 10 CFR part 2, such as standing and
admissibility, are met). This is a change to the standard that was
provided in the proposed rule for new emergency preparedness
information and is based on public comments. The proposed rule standard
for litigation of emergency preparedness matters was ``new or
additional information * * * which materially affects the Commission's
earlier determination on emergency preparedness, or is needed to
correct inaccuracies in the emergency preparedness information approved
in the early site permit.'' Because the final rule language suggested
by the commenters is the definition that the NRC gave for information
that could ``materially affect'' the Commission's earlier decision, as
indicated in the supplementary information section of the 2006 proposed
rule, the NRC believes it appropriate to use this language in the final
rule itself. The NRC has decided to drop the language that referred to
information ``needed to correct inaccuracies'' because the language, by
itself, could have allowed litigation of issues not significant to
safety. The NRC believes that the final rule language encompasses all
significant emergency preparedness matters that should be subject to
litigation.
Any significant environmental issue that was not resolved in the
early site permit proceeding, or any issue involving the impacts of
construction and operation of the facility that was resolved in the
early site permit proceeding for which significant new information has
been identified may also be the subject of a contention during the
proceeding on the referencing application. The NRC is also making a
change to this standard in the final rule based on public comment. The
standard in the final rule more closely reflects the NRC's obligation
under NEPA to address new and significant information in a COL that
references an early site permit. Additional discussion of this subject
can be found in the discussion of changes in 10 CFR part 51, in the
supplementary information section of this document.
Because new emergency planning or environmental information, if
any, will be identified only at the time a license application
referencing the early site permit is submitted to the NRC, the NRC
believes it is appropriate to address these issues in the proceeding on
the referencing application. Other questions regarding whether the
permit should be modified, suspended, or revoked will be challenges to
the validity of the early site permit. These challenges may be framed
in many different ways, e.g., a Commission error at the time of
issuance; or actual changes to the site have occurred since issuance of
the permit that render some aspect of the permit irrelevant or
inadequate to protect public health and safety or common defense and
security. The Commission's process for challenges to the validity of a
license is contained in 10 CFR 2.206. Accordingly, the Commission
concludes that challenges to the validity of an early site permit
should be processed in accordance with Sec. 2.206. In the Commission's
view, a variance is not fundamentally a challenge to the validity of
the early site permit, because it requests dispensation from compliance
with some aspect of the permit whose validity remains undisputed.
Therefore, the Commission concludes that variances should be treated as
proceeding-specific issues of compliance that are potentially valid
subjects of a contention in a proceeding for a referencing application.
The revisions to Sec. 52.39 are in agreement with these Commission
conclusions. Section 52.39 is being divided into five paragraphs
addressing different aspects of early site permit finality. Each
paragraph is provided with a subtitle characterizing the subject matter
addressed in that paragraph. Section 52.39(a) focuses on how the NRC
accords finality to an early site permit, with Sec. 52.39(a)(1)
setting forth the circumstances under which the NRC may modify an early
site permit. The rule language is based upon the existing regulation,
but adds additional circumstances. Section 52.39(a)(1)(iii) provides
that the NRC may modify the early site permit if it determines a
modification is necessary based on an update to the emergency
preparedness information under Sec. 52.39(b). Section 52.39(a)(1)(iv)
provides that the NRC may modify the early site permit if a variance is
issued under proposed Sec. 52.39(d) (paragraph (b) in the former
regulations); the NRC considers this a conforming change inasmuch as
the former regulation provided for issuance of variances.
The NRC is clarifying what aspects of the early site permit are
subject to the change restrictions in Sec. 52.39(a)(1) by substituting
the phrase, ``terms and conditions'' of an early site permit for the
former term, ``requirements.'' Under the new language, the NRC may not
change or impose new site characteristics, design parameters, or terms
and conditions on the early site permit, including emergency planning
requirements, unless the special backfitting criteria in Sec.
52.39(a)(1) are satisfied. No substantive change is intended by this
clarification; the language would specify more clearly the broad scope
of matters in an early site permit which the NRC intended to finalize.
The phrase, ``site characteristics, or terms, or conditions, including
emergency planning requirements,'' is used consistently throughout
Sec. 52.39 and corresponding provisions in the revisions to Sec.
52.79.
Section 52.39(a)(2) describes how the NRC treats matters resolved
in the early site permit proceeding in subsequent proceedings on
applications referencing the early site permit, and is drawn from the
former language of Sec. 52.39(a)(2). In the final rule, the NRC has
included a provision extending this finality to enforcement hearings
other than those proceedings initiated by the Commission under
paragraph (a)(1) of this section. This will ensure that finality of an
early site permit extends to NRC-initiated enforcement proceedings and
petitions for enforcement action filed under Sec. 2.206. In addition,
under Sec. Sec. 52.39(a)(2)(i) and (ii), the NRC grants finality to
changes to an early site permit's emergency plan (or major features of
it, under Sec. 52.17(b)(2)) that are made after the issuance of the
early site permit (1) if the early site permit approved an emergency
plan (or major features thereof) that is in use by a licensee of a
nuclear power plant and the changes to the emergency plan (or major
features thereof) are identical to changes made to the licensee's
emergency plans in compliance with Sec. 50.54(q); or (2) if the early
site permit approved an emergency plan (or major features thereof) that
is not in use by a licensee of a nuclear power plant, and the changes
are equivalent to those that could be made under Sec. 50.54(q) without
prior NRC approval had the emergency plan been in use by a licensee.
This change is premised on the view that changes to emergency plans
which are properly implemented under Sec. 50.54(q) do not require NRC
review and approval before implementation. Therefore, by analogy,
similar changes to an early site permit's emergency preparedness plan
made with similar controls, or changes which are equivalent to those
that could be made under Sec. 50.54(q) without prior NRC approval,
should not require NRC review and approval as part of the licensing
process. Any issues related to compliance with Sec. 50.54(q) should be
treated as an enforcement matter. Note that the NRC is making some
adjustments to this position in the final
[[Page 49379]]
rule based on public comments. The proposed rule would not have
excepted changes to early site permit emergency plans not in use by a
current licensee that could be made under Sec. 50.54(q) without prior
NRC approval had the emergency plans been in use by a licensee. The NRC
is making this change in the final rule because the Sec. 50.54(q)
standard ensures adequate protection of safety, and has been accepted
and used by the industry and NRC and it is appropriate to apply this
same standard to changes in all emergency plans approved by the NRC in
the ESP proceeding. The NRC is making similar changes to Sec.
52.79(b)(4) in the final rule to require that all COL applicants
referencing early site permits with complete and integrated emergency
plans or major features of emergency plans identify changes that have
been incorporated into the proposed facility emergency plans and that
constitute or would constitute a decrease in effectiveness under Sec.
50.54(q) of this chapter.
Section 52.39(b) is discussed separately under Section V.C.6.a of
this document, which discusses emergency preparedness requirements for
a combined license applicant referencing an early site permit.
Section 52.39(c) replaces the former criteria in Sec. Sec.
52.39(a)(2)(i) through (iii), governing how the NRC will treat various
issues with respect to the early site permit and its referencing in a
combined license application. Matters regarding compliance with the
early site permit which would be potentially valid subjects of a
contention are listed in Sec. Sec. 52.39(c)(1)(i) through (iii), e.g.,
whether the reactor proposed to be built under the referencing
application fits within the site characteristics and design parameters
specified in the early site permit; whether one or more of the terms
and conditions of the early site permit have been met; and whether a
variance requested by the referencing applicant is unwarranted or
should be modified. The NRC notes that all contentions at the early
site permit stage, including a contention pertaining to a variance,
must meet the requirements for contentions in Sec. 2.309(f). Matters
regarding significant new emergency preparedness or environmental
information material to the combined license proceeding, which would be
potentially valid subjects of contention under the proposed rule, are
listed in Sec. Sec. 52.39(c)(1)(iv) and (v).
Other matters, including changes to the site characteristics,
design parameters, or terms and conditions of the early site permit,
are treated under Sec. 52.39(c)(2) as challenges to the permit and
processed in accordance with Sec. 2.206. The NRC is retaining the
former provision in Sec. 52.39(a)(2)(iii) requiring that the
Commission consider a petition filed under Sec. 2.206, and determine
whether immediate action is required before construction commences, as
well as the former provision indicating that if a petition is granted,
the Commission will issue an appropriate order which does not affect
construction unless the Commission makes its order immediately
effective.
The final rule redesignates the former provision in Sec. 52.39(b)
allowing an applicant for a license referencing an early site permit to
request a variance from one or more ``elements'' of the early site
permit as Sec. 52.39(d). The rule clarifies ``elements'' for which a
variance may be sought by substituting the phrase, ``site
characteristics, design parameters, or terms and conditions of the
early site permit.'' In addition, the NRC is revising this provision
further to include an allowance for applicants to request a variance
from the site safety analysis report (SSAR). The allowance for
requesting variances to the SSAR was inadvertently omitted in the
proposed rule. Because the majority of the early site permit
information that a combined license applicant will be referencing will
be the information in the SSAR, it is logical that the allowance to
request variances be extended to the information in the SSAR given that
the NRC is allowing variances to the permit itself. The NRC notes that
the admission of a contention on a proposed variance, which was
formerly addressed in Sec. 52.39(b), is addressed in Sec.
52.39(c)(iii). The NRC is also adding a provision that precludes the
Commission from issuing a variance once a construction permit or
combined license referencing the early site permit is issued. Any
changes that would otherwise require a variance should instead be
treated as an amendment to the construction permit or combined license.
Finally, the NRC is adding a new paragraph to the ``finality''
section in each subpart of part 52, in this instance Sec. 52.39(f),
entitled ``Information requests,'' which delineates the restrictions on
the NRC for information requests to the holder of the early site
permit. This provision is analogous to the former provision on
information requests in paragraph 8 of appendix O to parts 50 and 52,
and is based upon the language of Sec. 50.54(f). For early site
permits, this provision is contained in Sec. 52.39(d), and requires
the NRC to evaluate each information request on the holder of an early
site permit to determine that the burden imposed by the information
request is justified in light of the potential safety significance of
the issue to be addressed in the information request. The only
exceptions would be for information requests seeking to verify
compliance with the current licensing basis of the early site permit.
If the request is from the NRC staff, the request would first have to
be approved by the Executive Director for Operations (EDO) or his or
her designee.
7. Subpart B, Standard Design Certifications
a. Section 52.41, Scope of Subpart
This section defines the scope of subpart B of part 52. The
requirements on scope and type of nuclear power plants that are
eligible for design certification were moved from former Sec. 52.45(a)
to this section, to ensure a consistent format and presentation among
all the subparts of part 52.
b. Section 52.43, Relationship to Other Subparts
This section defines the relationship of subpart B to other
subparts in 10 CFR part 52. Conforming changes were made to make clear
that an application for a manufacturing license may, but is not
required to, reference a design certification rule (DCR). The
requirements formerly located in Sec. Sec. 52.43(c), 52.45(c), and
52.47(b)(2)(ii) were removed because the Commission decided not to
require a final design approval (FDA) under subpart E as a prerequisite
for certification of a standard plant design. This requirement was
included in part 52, at the time of the original rulemaking, because
the NRC had no experience with design certifications. By requiring an
FDA as a prerequisite to design certification, the NRC indicated that
the licensing processes for design certifications and FDAs were
similar, even though the requirements for and finality of a design
certification differ from that of an FDA. The NRC now has considerable
experience with design certification reviews, and the former
requirement to apply for an FDA as part of an application for design
certification is no longer needed. Future applicants have the option to
apply for either an FDA, a design certification, or both.
c. Section 52.45, Filing of Applications
This section presents the requirements for filing design
certification applications. This section was reformatted for
consistency with the other subparts in part 52 and the references to
specific paragraphs within Sec. Sec. 50.4 and 50.30 were replaced with
references to subpart H of part 2. A new
[[Page 49380]]
Sec. 52.45(c) on design certification review fees, was moved from
Sec. 52.49.
d. Section 52.46, Contents of Applications; General Information
This section was added to set forth general content requirements
from 10 CFR 50.33.
e. Section 52.47, Contents of Applications; Technical Information
This section presents the requirements for contents of a design
certification application and is organized into three sections. The
requirements for the final safety analysis report (FSAR) are set forth
in Sec. Sec. 52.47(a) and 52.47(c), and the technical requirements for
the remainder of the design certification application are in Sec.
52.47(b). The former Sec. 52.47(a)(1)(i) required the submittal of
information required for construction permits and operating licenses by
parts 20, 50 (including the applicable requirements from 10 CFR 50.34),
73, and 100, which were technically relevant to the design and not
site-specific. That general requirement was removed and replaced with
specific requirements that describe what must be included in an FSAR.
In addition, the NRC included technical positions that were developed
after part 52 was originally codified in 1989, e.g., Sec. 52.47(a)(22)
which requires a description of how relevant operating experience was
incorporated into the standard design (see SRM on SECY-90-377, dated
February 15, 1991, ML003707892). Also, the relevant requirements were
revised to clarify their applicability to design certifications and
renumbered. This effort resulted in a comprehensive list of
requirements for a design certification application.
Some commenters recommended that the requirement to demonstrate
technical qualifications [now Sec. 52.47(a)(7)] be deleted because the
AEA only imposes that requirement on applicants for a license. Although
the NRC agrees that the AEA imposes the technical qualification finding
specifically for license applicants, it does not preclude the NRC from
a determination that such a finding is also necessary in other
contexts. The applicant creates information that may become the bases
for a future license and, therefore, must be qualified to perform
design, analyses, and safety determinations. Accordingly, the NRC has
concluded that a technical qualification finding should also be made
for design certification applicants.
Some commenters recommended that the requirement to address the
standard review plan (SRP) be revised to apply to light-water reactors.
The NRC agrees with this comment and has revised this requirement [now
Sec. 52.47(a)(9)] to be applicable to light-water-cooled nuclear power
plants, but notes that much of the SRP review guidance and criteria are
general and would also apply to reviews of gas-cooled reactor designs.
Some commenters recommended that the requirement to provide
information required by Sec. 50.49(d) [now Sec. 52.47(a)(13)] be
deleted because the applicant will not be able to establish
qualification files for all applicable components. The NRC agrees that
applicants may not be able to establish qualification files, but
applicants can provide the electric equipment list required by Sec.
50.49(d). Therefore, the NRC revised the wording in Sec. 52.47(a)(13)
to be consistent with the wording for the same provision in Sec.
52.79(a), which requires that applicants provide the list of electrical
equipment important to safety required by Sec. 50.49(d).
Some commenters recommended that the requirement in Sec.
52.47(a)(22) to demonstrate how operating experience insights have been
incorporated into the plant design be deleted. The NRC disagrees with
this comment. The NRC developed this requirement for future plants (see
SRM on SECY-90-377) and it was implemented in past design certification
applications by addressing NRC's generic letters and bulletins. The NRC
agrees that insights from generic letters and bulletins should be
incorporated into the latest revision of the standard review plan
(SRP). Therefore, for plant designs that are based on or are evolutions
of nuclear plants that have operated in the United States, the
applicant should use NRC's generic letters and bulletins issued after
the most recent revision of the applicable SRP and 6 months before the
docket date of the application. If the application is for a nuclear
plant design that is not based on or is not an evolution of a nuclear
plant that operated in the United States, the applicant should address
how insights from any relevant international operating experience has
been incorporated into that plant design.
Some commenters recommended that the requirement to describe severe
accident design features in the FSAR [now Sec. 52.47(a)(23)] be
deleted. The NRC disagrees with this comment because the Commission has
determined that this requirement is necessary for future light-water
reactor designs (see SRM on SECY-93-087) and was applied to previous
applications. The commenters confused the meaning of design bases
information (see Sec. 50.2) with the requirements for design-basis
accidents (DBAs). Postulated severe accidents are not design-basis
accidents and the severe accident design features do not have to meet
the requirements for DBAs (see SECY-93-087). However, the severe
accident design features are part of a plant's design bases
information.
A new Sec. 52.47(b) was created to set forth the required
technical contents of a design certification application that are not
required to be located in the FSAR. In response to public comments on
the proposed rule, the NRC has deleted proposed Sec. 52.47(b)(1) which
required design certification applicants to submit a design-specific
probabilistic risk assessment (PRA). In its place, the NRC has added
new Sec. 52.47(a)(27) which requires that design certification
applicants submit a description of the design-specific PRA and its
results in the FSAR. The NRC agrees with some commenters that
applicants should not be required to submit their complete design-
specific PRA and that, instead, applicants should only be required to
provide a summary description of the PRA and its results in their FSAR
with the understanding that the complete PRA (e.g., codes) would be
available for NRC inspection at the applicant's offices, if needed. The
NRC expects that, generally, the information that it needs to perform
its review of the design certification application from a PRA
perspective is that information that will be contained in applicants'
FSAR Chapter 19.
The rule language for ITAAC [now Sec. 52.47(b)(1)] was conformed
with the statutory language in the AEA. This clarification of the
language in the former Sec. 52.47(a)(1)(vi), which was a condensed
version of the language in the former Sec. 52.97(b)(1), was intended
to avoid any misunderstandings regarding the statutory requirement.
Some commenters recommended that the rule language in Sec. 52.47(b)(1)
be modified to maintain the language in the former Sec.
52.47(a)(1)(vi) claiming the proposed language could be misconstrued as
expanding the scope of ITAAC needed for design certification. The NRC
disagrees with this comment and notes that it is well understood that
the requirements that are applicable to design certification are
limited to the scope of the certified design.
Some commenters recommended that the requirement in proposed Sec.
52.47(b)(3) (now in 10 CFR 51.55) to evaluate severe accident
mitigation design alternatives (SAMDAs) be deleted and that the NRC
should initiate a rulemaking or policy statement to disposition SAMDA
generically. The NRC disagrees with this comment. The
[[Page 49381]]
NRC has required SAMDA evaluations for previous applications in order
to achieve greater finality for the design features that are resolved
in design certification rulemakings. Further, the initiation of a
rulemaking or policy statement for SAMDAs is outside the scope of the
part 52 update rulemaking. As for the perspective that SAMDA
evaluations need not be performed for current reactor designs because
the severe accident risk for such designs is too remote and
speculative, the NRC has already addressed this issue in other
contexts. The NRC has considered petitions to eliminate the
consideration of SAMDAs previously. The NRC position, both then and now
is that it is not prepared to reach the conclusion that the risks of
all severe accidents are so unlikely as to warrant their elimination
from consideration in our NEPA reviews. As the NRC has stated in
response to other requests to confine or eliminate such issues from
consideration, if new information in the future provides a firm basis
for concluding that severe accidents are remote and speculative, then
the NRC may revisit the issue.
Former Sec. 52.47(b) was reorganized by separating the
requirements on scope of design and modular configuration [now located
in Sec. 52.47(c)] from the testing requirements. This action is part
of the NRC's goal to put the procedural requirements for the licensing
processes in part 52 and maintain the reactor safety requirements in
part 50 (or other parts of 10 CFR Chapter I. As a result, the testing
requirements were relocated to Sec. 50.43(e). Also, see the discussion
on testing for advanced nuclear reactors in Section V.B of this
document.
f. Section 52.54, Issuance of Standard Design Certification
This section was amended to be consistent with the parallel
provisions in Sec. Sec. 50.50 and 50.57 by including requirements
that, after conducting a rulemaking proceeding and receiving the report
submitted by the ACRS, the NRC will determine whether there is
reasonable assurance that the design conforms with the provisions of
the AEA, and the NRC's regulations; that the applicant is technically
qualified; and that issuance of the design certification will not be
inimical to the common defense and security or to the health and safety
of the public. In addition, a new Sec. 52.54(a)(8) was added to state
that the NRC will not issue a design certification unless it finds that
the design certification applicant has implemented the quality
assurance program described in the safety analysis report. This
requirement was added to indicate the NRC's expectation that design
certification applicants will implement the QA program that is required
to be included in their application under Sec. 52.47(a)(19), which is
consistent with the requirement for licensees.
A new Sec. 52.54(b) was added to require that a design
certification specify the site parameters and design characteristics
and any additional requirements and restrictions of the rule, as the
Commission deems necessary and appropriate. Some commenters recommended
that the requirement in Sec. 52.54(b) to list ``design
characteristics'' be removed and noted that the design control document
will contain this information. The NRC disagrees with this comment. The
NRC wants to specifically identify this information to facilitate
future comparisons with ``design parameters'' specified in an early
site permit. The NRC staff will use its experience with current early
site permit reviews to determine what an appropriate list will be for
future design certification reviews.
The NRC also modified Sec. 52.54 to require that applicants for a
design certification agree to withhold access to National Security
Information from individuals until the requirements of 10 CFR parts 25
and/or 95, as applicable, are met. Section 52.54 was amended to include
a new paragraph (c) which requires that every DCR contain a provision
stating that, after the Commission has adopted the final design
certification rule, the applicant for that design certification will
not permit any individual to have access to, or any facility to
possess, Restricted Data or classified National Security Information
until the individual and/or facility has been approved for access under
the provisions of 10 CFR parts 25 and/or 95. The NRC believes that this
amendment, along with the changes to parts 25, 95, and 10 CFR 50.37,
are necessary to ensure that access to classified information is
adequately controlled by all entities applying for NRC certifications.
g. Section 52.63, Finality of Standard Design Certifications
The final rule revises the finality provisions in Sec. 52.63(a) to
provide processes for amending design certification information without
meeting the special backfit requirement in Sec. 52.63(a)(1)(ii). The
special backfit requirement restricted changes to certification
information, thereby ensuring that all plants built under a referenced
certified design would be standardized. Section 52.63(a)(1) was also
revised to replace ``a modification'' with ``the change,'' to clarify
that the criteria for changes apply to modifications, rescissions, or
imposition of new requirements. In addition, Sec. 52.63 was revised to
use the phrase ``certification information'' in order to distinguish
the rule language in the DCRs from the design certification information
(e.g., Tier 1 and Tier 2 information) that is incorporated by reference
in the DCRs.
Section 52.63(a)(1)(iii) was added to provide the NRC with the
ability to make generic changes to the design certification rule
language that reduce unnecessary regulatory burdens. The former Sec.
52.63(a)(1) stated that the Commission may not modify, rescind, or
impose new requirements on the certification unless the change is: (1)
Necessary for compliance with Commission regulations applicable and in
effect at the time the certification was issued; or (2) necessary to
provide adequate protection of the public health and safety or common
defense and security. This requirement did not appear to permit changes
to the rule language which reduce unnecessary regulatory burdens in
circumstances where the change continues to maintain protection to
public health and safety and common defense and security. An example of
a change which could not be made under the former Sec. 52.63(a)(1) was
a change to the rule language in appendices A, B, and C of part 52, to
incorporate into the Tier 2 change process the revised change criteria
in 10 CFR 50.59. Section 50.59 was revised in 1999 to provide new
criteria for, inter alia, making changes to a facility, as described in
the final safety analysis report, without prior NRC approval, to reduce
unnecessary regulatory burden (64 FR 53582, October 4, 1999).
In Section V of the 2006 proposed rule, Question 14, the NRC stated
that it was considering adopting an additional provision in Sec.
52.63(a)(1) that would allow amendments of DCRs to incorporate generic
resolutions of design acceptance criteria (DAC) or other design
information without meeting the special backfit requirement in the
former Sec. 52.63(a)(1). By allowing for an amendment to generically
resolve DAC, the NRC would achieve resolution of additional design
issues, would achieve finality for those issue resolutions, and would
avoid repetitive consideration of those design issues in individual
combined license proceedings. The final rule includes an amendment
process in Sec. 52.63(a)(1)(iv) that allows for generic resolutions of
DAC without meeting the special backfit requirement. These amendments
will
[[Page 49382]]
apply to all plants that have or will reference the DCR under Sec.
52.63(a)(2). The NRC believes that these amendments will enhance
standardization by further completing the certification information.
The NRC will review the amendment application to ensure that the design
acceptance criteria are met and that the new design information
conforms with the applicable regulations.
Some commenters proposed that the amendment process should allow
for generic resolutions of errors in the certification information. The
NRC is aware that design certification applicants have discovered
errors in their design information after the NRC has completed its
review and even after the NRC has certified their design. The final
rule includes a new provision in Sec. 52.63(a)(1)(v) to correct
material errors in the certification information. This provision is
only to be used to correct a material error, which is an error that
significantly and adversely affects a design function or analysis
conclusion described in the design control document (certification
information). The NRC wants to correct material errors by amendment so
that these errors will not have to be addressed in individual licensing
proceedings.
Many commenters encouraged the NRC to adopt an amendment process
that would allow for ``beneficial'' changes to certification
information, would apply the amendment to all plants referencing the
certified design, and would only allow amendments prior to issuance of
the first combined license that referenced the DCR. The NRC agreed with
these comments and included paragraph (a)(1)(vi) to allow for
amendments of certification information that will substantially
increase the overall safety, reliability, or security of facility
design, construction, or operation provided that the direct and
indirect costs of implementation of the amendment are justified in view
of this increased safety, reliability, or security. However, the NRC
does not agree with precluding amendments after issuance of the first
combined license. If licensees who referenced a DCR want to adopt a
proposed amendment in order to achieve enhanced standardization and the
beneficial changes that the amendment would bring, then the NRC may
amend the DCR and apply the amendment to all plants referencing the
DCR.
Also, some commenters requested that the amendment process allow
for changes to the certification information for a wide variety of
other reasons. These commenters claimed that the need for a design
change may be discovered during detailed design work performed after
the original design information was approved by the NRC (so-called
first-of-a-kind-engineering) or that certain components in the original
design may no longer be available for purchase due to the long duration
of a DCR. The NRC's deliberations on this proposal considered the
Commission's goal for design certification, which is to achieve and
maintain the benefits of standardization. The NRC is still determined
to maintain standardization, but has decided to allow amendments for
other design changes [see paragraph (a)(1)(vii)] provided that the
amendment will be applied to all plants that reference the DCR, thereby
increasing standardization. In determining whether to codify a proposed
amendment, the NRC will give special consideration to comments from
applicants or licensees who reference the DCR regarding whether they
want to backfit their plants with these additional design changes.
The final rule includes a new Sec. 52.63(a)(2), which sets forth
procedures for rulemakings conducted under Sec. 52.63(a)(1). Paragraph
(a)(2)(i) requires that for rulemakings under Sec. 52.63(a)(1), except
for rulemakings under Sec. 52.63(a)(1)(ii) necessary to provide
adequate protection, the NRC will give consideration to whether the
benefits justify the costs for plants that are already licensed or for
which an application for a license is under consideration.
The final rule also revised the former Sec. 52.63(a)(2) [now Sec.
52.63(a)(3)] to delete the reference to the former Sec. 52.63(a)(4)
[now Sec. 52.63(a)(5)]. The reference to the former Sec. 52.63(a)(4)
was in error because this paragraph discusses the finality of the
findings required for issuance of a combined license or operating
license, whereas the new Sec. 52.63(a)(3) deals with modifications
that the NRC may impose on a DCR under Sec. Sec. 52.63(a)(4) or
52.63(b)(1). No substantive change is intended by this revision, which
merely clarifies the intent of the rule.
Finally, the NRC restates its previous decision regarding the
ability of any person to request an amendment to a DCR. In Section
II.1.h of the 1989 SOC for part 52 (54 FR 15372), the Commission stated
that Sec. 52.63(a)(1) places a designer on the same footing as the NRC
or any other interested member of the public. Therefore, anyone may
submit a petition for rulemaking to the NRC to correct an error or
otherwise amend the certification information. All amendments to the
certification information must be accomplished through rulemaking, with
an opportunity for public comment under Sec. 52.63(a)(2). Once a
certified design is amended by rulemaking, the new rule would apply to
all applications referencing the DCR as well as all plants referencing
the DCR, unless the change has been rendered ``technically irrelevant''
through other action taken under Sec. Sec. 52.63(a)(4) or (b)(1).
Also, the NRC will decide whether to codify the proposed amendment
based on comments from the referencing applicants and licensees. Thus,
standardization is maintained by ensuring that any generic change to
the certification information is imposed upon all nuclear power plants
referencing the DCR. The duration of the amended DCR will be for the
same period of time as the original DCR and have the same expiration
date.
8. Subpart C, Combined Licenses
a. Emergency Preparedness Requirements for a Combined License Applicant
Referencing an Early Site Permit
The NRC is revising former Sec. Sec. 52.39 and 52.79 to require a
license applicant referencing an early site permit to update and
correct the emergency preparedness information provided under Sec.
52.17(b). The issue of updating an early site permit was first raised
by the Illinois Department of Nuclear Safety, who suggested in a
September 28, 1994, letter that emergency plans and/or offsite
certifications approved as part of an early site permit review be kept
up-to-date throughout the duration of an early site permit and the
construction phase of a combined license.
In SECY-95-090, ``Emergency Planning Under 10 CFR Part 52'' (April
11, 1995), the NRC staff stated that 10 CFR part 52 does not clearly
require an applicant referencing an early site permit to submit updated
information on changes in emergency preparedness information or in any
emergency plans that were approved as part of the early site permit in
accordance with Sec. 52.18. SECY-95-090 indicated (p. 4) that, in view
of the lack of industry interest in pursuing an early site permit,
resolution of this matter could be deferred until a ``lessons learned''
rulemaking, updating 10 CFR part 52, was conducted after the first
design certification rulemakings were issued. Following public release
of a draft SECY paper setting forth the NRC staff's preliminary views
on the licensing process for a combined license, NEI submitted a letter
dated September 8, 1998 (comment 2.d), which expressed opposition to a
requirement for updating emergency preparedness information throughout
[[Page 49383]]
the duration of an early site permit, absent an application referencing
the early site permit. As an alternative to updating throughout the
duration of an early site permit, NEI proposed that emergency planning
information be updated when an application for a license referencing
the early site permit is filed; portions of the emergency plans that
are unchanged would continue to have finality under 10 CFR 52.39. In a
September 3, 1999 letter, the NRC staff identified updating of
emergency preparedness information in early site permits as a possible
subject for the part 52 rulemaking.
The NRC agrees in part with the Illinois Department of Nuclear
Safety. Emergency plans and/or offsite certificates in support of
emergency plans, approved as part of an early site permit review,
should be updated. However, emergency plans do not need to be kept up-
to-date throughout the duration of an early site permit. There is no
need to update the emergency plans approved in an early site permit
until the time the permit is referenced in a combined license
application. At that time, the emergency plans would have to be
reviewed to confirm that they are up-to-date and to provide any new
information that may materially affect the NRC's earlier determination
on emergency preparedness, or correct inaccuracies in the emergency
preparedness information approved in the early site permit in support
of a reasonable assurance determination, in accordance with Sec. 50.47
and appendix E to part 50. In addition, the NRC agrees with NEI that a
``continuous'' early site permit update requirement would impose
burdens upon the early site permit holder without any commensurate
benefit if the early site permit is not subsequently referenced.
Accordingly, the Commission has determined that Sec. Sec. 52.39 and
52.79 should contain an updating requirement to be imposed upon the
applicant referencing an early site permit.
A new Sec. 52.39(b) is added to require an applicant for a
construction permit, operating license, or combined license, whose
application references an early site permit, to update and correct the
emergency preparedness information provided under Sec. 52.17(b). In
addition, the applicant must discuss whether the new information could
materially change the bases for compliance with the applicable NRC
requirements. A parallel requirement is included in Sec. 52.79 to
ensure that applicants for combined licenses referencing an early site
permit will submit the updated emergency preparedness information.
Section 52.39(a)(1)(iii) is also added stating that the Commission may
modify an early site permit if it determines that a modification is
necessary based on updated emergency preparedness information provided
in a referencing license application. New information that materially
changes the bases for compliance includes information that
substantially alters the bases for a previous NRC conclusion with
respect to the acceptability of a material aspect of emergency
preparedness or an emergency preparedness plan, and information that
would constitute a basis for the Commission to modify or impose new
terms and conditions on the early site permit related to emergency
preparedness in accordance with Sec. 52.39(a)(1). New information that
materially changes the NRC's determination of the matters in Sec.
52.17(b), or results in modifications of existing terms and conditions
under Sec. 52.39(a)(1) will be subject to litigation during the
construction permit, operating license, or combined license proceedings
in accordance with Sec. 52.39(c).
Not all new information on emergency preparedness will be subject
to challenge in a hearing under Sec. 52.39(c). For example, an
emergency plan may have to be updated to reflect current telephone
numbers, names of governmental officials whose positions and
responsibilities are defined in the plan (e.g., the name of the current
police chief for a municipality), or current names of hospital
facilities. These corrections do not materially change the NRC's
previously-stated bases for accepting the early site permit emergency
plan, and a hearing contention will not be admitted under Sec.
52.39(c) in a proceeding for a license referencing the early site
permit. In contrast, if an emergency plan submitted as part of an early
site permit relies upon a bridge to provide the primary path of
evacuation, and that bridge no longer exists, the change could
materially affect the NRC's previous determination that the emergency
plan complied with the Commission's emergency preparedness regulations
in effect at the time of the issuance of the early site permit. This
type of information might be the basis for a change in the early site
permit's terms and conditions related to emergency preparedness under
Sec. 52.39(a)(1), as well as the basis for a hearing contention under
Sec. 52.39(c), assuming that the requirements in 10 CFR part 2 for
admission of a contention are met.
b. Resolution of ITAAC
Sections 52.99 and 52.103 are revised to incorporate rule language
from the design certification regulations in 10 CFR part 52 regarding
the completion of ITAAC (see paragraphs IX.A and IX.B.3 of appendix A
to part 52). During the preparation of the design certification rules
for the ABWR and System 80+ designs, the NRC staff and nuclear industry
representatives agreed on certain requirements for the performance and
completion of the inspections, tests, or analyses in ITAAC. In the
design certification rulemakings, the NRC codified these ITAAC
requirements into Section IX of the regulations. The purpose of the
requirement in Sec. 52.99(b) is to clarify that an applicant may
proceed at its own risk with design and procurement activities subject
to ITAAC, and that a licensee may proceed at its own risk with design,
procurement, construction, and preoperational testing activities
subject to an ITAAC, even though the NRC may not have found that any
particular ITAAC has been met.
Section 52.99(c) requires the licensee to notify the NRC that the
prescribed inspections, tests, and analyses in the ITAAC have been or
will be completed and that the acceptance criteria have been met. The
NRC is revising Sec. 52.99(c)(1) in the final rule to more closely
follow the language of Section 185b. of the AEA (in response to a late-
filed comment) and to clarify that the notification must contain
sufficient information to demonstrate that the prescribed inspections,
tests, and analyses have been performed and that the prescribed
acceptance criteria have been met. The NRC is adding this clarification
to ensure that combined license applicants and holders are aware that
(1) it is the licensees' burden to demonstrate compliance with the
ITAAC and (2) the NRC expects the notification of ITAAC completion to
contain more information than just a simple statement that the licensee
believes the ITAAC has been completed and the acceptance criteria met.
The NRC expects the notification to be sufficiently complete and
detailed for a reasonable person to understand the bases for the
licensee's representation that the inspections, tests, and analyses
have been successfully completed and the acceptance criteria have been
met. The term ``sufficient information'' requires, at a minimum, a
summary description of the bases for the licensee's conclusion that the
inspections, tests, or analyses have been performed and that the
prescribed acceptance criteria have been met. The
[[Page 49384]]
NRC plans to prepare regulatory guidance, in consultation with
interested stakeholders, to explain how the functional requirement to
provide ``sufficient information'' with regard to ITAAC submittals
could be met.
The NRC is also revising Sec. 52.99(c) in the final rule by adding
a new paragraph (c)(2) requiring that, if the licensee has not
provided, by the date 225 days before the scheduled date for initial
loading of fuel, the notification required by paragraph (c)(1) of this
section for all ITAAC, then the licensee shall notify the NRC that the
prescribed inspections, tests, or analyses for all uncompleted ITAAC
will be performed and that the prescribed acceptance criteria will be
met prior to operation (consistent with the Section 189.a(1)(B)
requirement governing a request for hearing on acceptance criteria, and
the Section 185.b. requirement that the Commission find that the
acceptance criteria in the combined license are met). The notification
must be provided no later than the date 225 days before the scheduled
date for initial loading of fuel. It is the licensee's burden to
demonstrate that it will comply with the ITAAC and it must provide
sufficient information to demonstrate that the prescribed inspections,
tests, or analyses will be performed and the prescribed acceptance
criteria for the uncompleted ITAAC will be met. The term ``sufficient
information'' requires, at a minimum, a summary description of the
bases for the licensee's conclusion that the inspections, tests, or
analyses will be performed and that the prescribed acceptance criteria
will be met. In addition, ``sufficient information'' includes, but is
not limited to, a description of the specific procedures and analytical
methods to be used for performing the inspections, tests, and analyses
and determining that the acceptance criteria have been met.
Paragraph (e) has been revised to require that the NRC make
available to the public the notifications to be submitted under Sec.
52.99(c)(1) and (c)(2), no later than the Federal Register notice of
intended operation and opportunity for hearing on ITAAC under Sec.
52.103(a). A conforming change is included in Sec. 2.105(b)(3) to
require that the Sec. 52.103(a) notice reference the public
availability of the Sec. 52.99(c)(1) and (2) notifications. The NRC is
requiring that the paragraph (c)(2) notification be made 225 days
before the date scheduled for initial loading of fuel, in order to
ensure that the licensee notifications are publicly available through
the NRC document room and online through the NRC Web site at the same
time that the Sec. 52.103(a) notice is published in the Federal
Register. The NRC's goal is to publish that notice 210 days before the
date scheduled for fuel loading, but in all cases the Sec. 52.103(a)
notice would be published no later than 180 days before the scheduled
fuel load, as required by Section 189.a(1)(B) of the AEA.
In Section V of the Supplementary Information of the proposed rule,
the NRC requested stakeholder feedback on whether a provision on
completion of ITAAC in a set time period prior to fuel load should be
added to the final rule. Commenters did not support addition of a
requirement on completion of ITAAC in a set time period prior to fuel
load and the NRC has not included a provision requiring the completion
of all ITAAC by a certain time prior to the licensee's scheduled fuel
load date. Instead, the NRC has decided to modify the concept slightly
by requiring the licensee to submit, with respect to ITAAC which have
not yet been completed 225 days before the scheduled date for initial
loading of fuel, additional information addressing whether those
inspections, tests, and analyses will be successfully completed and the
acceptance criteria met before initial operation. In the case where the
licensee has not completed all ITAAC by 225 days prior to its scheduled
fuel load date, the NRC expects the information that the licensee
submits related to uncompleted ITAAC to be sufficiently detailed such
that the NRC can determine what activities it will need to undertake to
determine if the acceptance criteria for each of the uncompleted ITAAC
have been met, once the licensee notifies the NRC that those ITAAC have
been successfully completed and their acceptance criteria met. In
addition, the NRC is adopting the requirements in paragraphs (c)(1) and
(c)(2) to ensure that interested persons will be able to meet the
Atomic Energy Act, Section 189.a(1), threshold for requesting a hearing
with respect to both completed and as-yet uncompleted ITAAC. The NRC
therefore expects that the information submitted by licensees in the
Sec. 52.99(c)(2) notification will be sufficiently complete and
detailed. Furthermore, the NRC expects that any contentions submitted
by prospective intervenors regarding uncompleted ITAAC would focus on
the inadequacies of the procedures and analytical methods described by
the licensee for completing those ITAAC in the context of the
reasonable assurance finding under Sec. 52.103(b)(2). Therefore, the
level of detail provided by the licensee should be sufficient to allow
a prospective intervenor to form such judgments by reference to that
information. The NRC plans to prepare regulatory guidance providing
further explanation of what constitutes ``sufficient information'' to
demonstrate that the inspections, tests, or analyses for uncompleted
ITAAC will be successfully completed and the acceptance criteria for
the uncompleted ITAAC will be met.
The NRC notes that, even though it did not include a provision
requiring the completion of all ITAAC by a certain time prior to the
licensee's scheduled fuel load date, the NRC will require some period
of time to perform its review of the last ITAAC once the licensee
submits its notification that the ITAAC has been successfully completed
and the acceptance criteria met. In addition, the Commission will
require some period of time to perform its review of the staff's
conclusions regarding all of the ITAAC and the staff's recommendations
regarding the Commission finding under Sec. 52.103(g). Therefore,
licensees should structure their construction schedules to take into
account these time periods. The NRC intends to develop regulatory
guidance on the licensee's completion and NRC verification of ITAAC and
will provide estimates of the time it expects to take to verify
successful completion of various types of ITAAC. The NRC expects that
such guidance, along with frequent communication with licensees during
construction, will provide licensees with adequate information to plan
initial fuel loading and related activities.
Section 52.99(d) states the options that a licensee will have in
the event that it is determined that any of the acceptance criteria in
the ITAAC have not been met. The NRC is revising Sec. 52.99(d) in the
final rule as a result of comments made on the proposed rule. Proposed
Sec. 52.99(d) stated that, in the event that an activity is subject to
an ITAAC derived from a referenced early site permit or standard design
certification and the licensee has not demonstrated that the ITAAC has
been met, the licensee may take corrective actions to successfully
complete that ITAAC, request a variance from the early site permit
ITAAC, or request an exemption from the standard design certification
ITAAC, as applicable. The language in proposed Sec. 52.99(d) that
referred to requesting variances to ESP ITAAC after the COL is issued
is inconsistent with rule language in other sections of proposed part
52 (e.g., Sec. 52.39(d)). Therefore, the NRC has adopted the
commenters' suggestion to delete references to ESP ITAAC and ESP
variances from Sec. 52.99(d).
[[Page 49385]]
Paragraph (e)(1) requires the NRC to publish, at appropriate
intervals until the last date for submission of requests for hearing
under Sec. 52.103(a), notices in the Federal Register of the NRC
staff's determination of the successful completion of inspections,
tests, and analyses. Paragraph (e)(2) provides that the NRC shall make
publicly available the licensee notifications under paragraphs (c)(1)
and (c)(2). In general, the NRC expects to make the paragraph (c)(1)
notifications availability shortly after the NRC has received the
notifications and concluded that they are complete and detailed.
Furthermore, by the date of the Federal Register notice of intended
operation and opportunity to request a hearing on whether acceptance
criteria have been or will be met (under Sec. 52.103(a)), the NRC will
make available the notifications under paragraph (c)(2), and the
notifications under paragraph (c)(2) for all ITAAC for which paragraph
(c)(1) notifications have not been provided by the licensee.
Finally, Sec. 52.103(h) states that ITAAC do not, by virtue of
their inclusion in the combined license, constitute regulatory
requirements after the licensee has received authorization to load fuel
or for renewal of the license. However, subsequent modifications must
comply with the design descriptions in the design control document
unless the applicable requirements in the Sec. 52.97 (proposed Sec.
52.98) and Section VIII of the design certification rules have been
complied with.
In a letter dated April 3, 2001 (comment 23), NEI requested that
the NRC ``consider incorporating DCR [Design Certification Rule]
general provisions into Subpart C as appropriate.'' The NRC has added
these ITAAC requirements to Sec. 52.99, consistent with NEI's
proposal, because it believes that these provisions embody general
principles that are applicable to all holders of combined licenses.
The NRC revised Sec. 52.99 in the final rule to delete the
requirements in proposed Sec. 52.99(a). Proposed Sec. 52.99(a)
required holders of COLs to comply with the provisions of Sec. Sec.
50.70 and 50.71. Because the language in proposed Sec. Sec. 50.70 and
50.71 requires COL holders to comply with their provisions, and because
of the applicability provisions in Sec. 52.0(b), this duplicate
requirement in Sec. 52.99 is unnecessary.
The NRC has added a new paragraph (a) in Sec. 52.99 that requires
a licensee to submit to the NRC, no later than 1 year after issuance of
the combined license or at the start of construction as defined in 10
CFR 50.10, whichever is later, its schedule for completing the
inspections, tests, or analyses in the ITAAC. Licensees are required to
submit updates to the ITAAC schedule every 6 months thereafter and,
within 1 year of its scheduled date for initial loading of fuel,
licensees must submit updates to the ITAAC schedule every 30 days until
the final notification is provided to the NRC under Sec. 52.99(c). In
Section V of the Supplementary Information of the 2006 proposed rule,
the NRC requested stakeholder feedback on whether such a provision
should be added to the final rule. Although some commenters did not
believe that a regulatory requirement for submission of a schedule was
necessary, the NRC believes it is necessary to ensure the NRC has
sufficient information to plan all of the activities necessary for the
NRC to support the Commission's finding whether all of the ITAAC have
been met prior to the licensee's scheduled date for fuel load.
c. Section 52.73, Relationship to Other Subparts
Section 52.73 clarifies that a design approval issued under subpart
E of part 52 or a manufacturing license under subpart F of part 52 may
also be referenced in an application for a combined license filed under
10 CFR part 52. The former Sec. 52.73 only stated that a combined
license may reference a standard design certification or an early site
permit. The final rule incorporates into new Sec. 52.73(b) the
requirement in the current Sec. 52.63(c) in order to clarify that this
requirement applies to applicants for a combined license. This
provision requires that, before granting a combined license which
references a standard design certification, information normally
contained in certain procurement specifications and construction and
installation specifications be completed and available for audit if the
information is necessary for the NRC to make its safety determinations,
including the determination that the application is consistent with the
certified design. No substantive change is intended by the restatement
of this requirement. In a letter dated April 3, 2001 (comments 3 and
3.a), NEI agreed with the proposed change but recommended that the last
sentence of Sec. 52.63(c) be deleted and the remaining provision be
added to the former Sec. 52.79 rather than the former Sec. 52.73. The
NRC agrees with NEI that 10 CFR part 52 should be modified to clarify
that the requirement in former Sec. 52.63(c) applied to applicants for
a combined license, and that the last sentence be deleted. However, the
Commission is adding the remaining provision to the original Sec.
52.73(b), and not to Sec. 52.79, as recommended by NEI.
d. Section 52.75, Filing of Applications
Section 52.75 provides requirements for the filing of combined
license applications. The NRC has reformatted this section for
consistency with the other subparts in 10 CFR part 52 and to replace
the references to specific paragraphs within Sec. Sec. 50.4 and 50.30
with general references to those sections. The specific references are
no longer needed because the NRC is adopting conforming changes to
Sec. Sec. 50.4 and 50.30 in this final rule which clarify which
provisions are applicable to combined license applications.
e. Section 52.78, Content of Applications; Training and Qualification
of Nuclear Power Plant Personnel
Section 52.78 has been removed, and the requirements applicable to
an applicant for, and holder of, a combined license with respect to the
training program are moved to Sec. 50.120, where the requirements
currently exist for holders of operating licenses.
f. Section 52.79, Contents of Applications; Technical Information in
Final Safety Analysis Report; and Sec. 52.80, Contents of Application;
Additional Technical Information
Section 52.79 is reformatted to divide the requirements for the
technical contents of a combined license application into two separate
provisions. Section 52.79 covers requirements for the contents of the
FSAR, and Sec. 52.80 covers requirements for the remainder of the
technical content of a combined license application.
Former Sec. 52.79 states that a combined license application must
contain the technically relevant information required of applicants for
an operating license by 10 CFR 50.34. The reference to 10 CFR 50.34 is
removed and replaced with Sec. 52.79(a), which contains all of the
relevant requirements from 10 CFR 50.34 that describe what must be
included in the FSAR for a combined license application, including
requirements that are currently applicable to both construction permit
and operating license applications. In addition, requirements from
other sections of 10 CFR part 50 (e.g., Sec. Sec. 50.48 and 50.63) are
included. These requirements were issued after the current fleet of
operating reactors were licensed and, therefore, were not required
contents for these earlier FSARs. In making these modifications,
[[Page 49386]]
the NRC has attempted to capture all relevant requirements regarding
contents of the FSAR for a combined license application.
In addition, Sec. 52.79(a) contains requirements for descriptions
of operational programs that need to be included in the FSAR to allow a
reasonable assurance finding of acceptability. This amendment is in
support of the Commission's direction to the staff in SRM-SECY-02-0067
dated September 11, 2002, ``Inspections, Tests, Analyses, and
Acceptance Criteria for Operational Programs (Programmatic ITAAC),''
that a combined license applicant was not required to have ITAAC for
operational programs if the applicant fully described the operational
program and its implementation in the combined license application. In
this SRM, the Commission stated:
[a]n ITAAC for a program should not be necessary if the program
and its implementation are fully described in the application and
found to be acceptable by the NRC at the COL stage. The burden is on
the applicant to provide the necessary and sufficient programmatic
information for approval of the COL without ITAAC.
The Commission clarified its definition of fully described in SRM-
SECY-04-0032, ``Programmatic Information Needed for Approval of a
Combined License Application Without Inspections, Tests, Analyses, and
Acceptance Criteria,'' dated May 14, 2004, as follows:
In this context, fully described should be understood to mean
that the program is clearly and sufficiently described in terms of
the scope and level of detail to allow a reasonable assurance
finding of acceptability. Required programs should always be
described at a functional level and at an increased level of detail
where implementation choices could materially and negatively affect
the program effectiveness and acceptability.
Accordingly, the NRC is adding requirements for descriptions of
operational programs. In doing so, the NRC has taken into account NEI's
proposal to address SRM-SECY-04-0032 in its letter dated August 31,
2005 (ML052510037). That proposal was reflected in SECY-05-0197
(October 28, 2005, ML052770225), Attachment 1, and approved by the
Commission in SRM-SECY-05-0197 dated February 22, 2006 (ML060530316).
During the preparation of the final rule, the NRC discovered that
several of the operational programs listed in SECY-05-0197 were not
addressed in proposed Sec. 52.79. To ensure the list of requirements
for the contents of applications is complete, the NRC is adding several
new provisions to address operational programs in the final rule.
Specifically, the NRC is adding requirements to Sec. 52.79 for COL
applicants to include a description of: (1) The process and effluent
monitoring and sampling program required by appendix I to 10 CFR part
50 [Sec. 52.79(a)(16)(ii)]; (2) a training and qualification plan in
accordance with the criteria set forth in appendix B to 10 CFR part 73
[Sec. 52.79(a)(36)(ii)]; (3) a description of the radiation protection
program required by Sec. 20.1101 [Sec. 52.79(a)(39)]; (4) a
description of the fire protection program required by Sec. 50.48
[Sec. 52.79(a)(40)]; and (5) a description of the fitness-for-duty
program required by 10 CFR part 26 [Sec. 52.79(a)(44)]. During the
preparation of the final rule, the NRC also noticed that the proposed
rule had not completely implemented the Commission's direction
regarding the treatment of operational programs in a COL application
inasmuch as requirements to address operational program implementation
were not included in proposed Sec. 52.79(a). Therefore, in the final
rule, the NRC has added requirements to address the implementation of
all operational programs required to be described in a COL application.
This is consistent with the Commission's position in SRM-SECY-02-0067
that a combined license applicant is not required to have ITAAC for
operational programs if the applicant ``fully describes the operational
program and its implementation'' in the combined license application
[emphasis added].
In addition, the NRC added a new provision to Sec. 52.79(a) in the
final rule to address the application requirements in current Sec.
20.1406. Section 20.1406 requires applicants for a license to describe
in their application how facility design and procedures for operation
will minimize, to the extent practicable, contamination of the facility
and the environment, facilitate eventual decommissioning, and minimize,
to the extent practicable, the generation of radioactive waste. To
ensure that Sec. 52.79 contains a complete list of the requirements
for the contents of a COL application, the NRC added paragraph (a)(45)
to Sec. 52.79 to require COL applications to include the information
required by Sec. 20.1406. This is not a new requirement but merely a
pointer to an existing requirement to include this information.
Section 52.79(a) requires that emergency plans submitted with a
combined license application be included in the FSAR. This modification
from the former rule is being made for consistency with Sec. 50.34
which requires that emergency plans be included in the FSAR for
operating license applications.
The NRC is adding a new provision in Sec. 52.79(a)(29)(ii) that
the applicant submit plans for coping with emergencies, other than the
plans required by Sec. 52.79(a)(21). Paragraph 52.79(a)(21) requires
the applicant to submit emergency plans complying with the requirements
of Sec. 50.47 and 10 CFR part 50, appendix E. This requirement was
drawn from the existing requirement in Sec. 50.34(b)(6)(v) which
requires applicants to submit ``Plans for coping with emergencies,
which shall include the items specified in appendix E.'' When this
requirement was translated into the associated requirement for combined
license applicants, the NRC inadvertently only included a portion of
the requirements in Sec. 50.34(b)(6)(v), namely, the requirement in
proposed Sec. 52.79(a)(21) to submit emergency plans. The NRC has
corrected this omission in the final rule by including the new
provision in Sec. 52.79(a)(29)(ii) to include other plans for coping
with emergencies. This requirement is meant to capture, for example,
emergency operating procedures as discussed in SRP Section 13.5.2.1,
``Operating and Emergency Operating Procedures.''
The NRC has moved the requirements contained in proposed Sec.
52.79(a)(23) that addressed a request to conduct activities under Sec.
50.10(e) and added them in a new Sec. 52.80(c). The NRC concluded that
it is preferable to include both the list of proposed Sec. 50.10(e)
activities and the redress plan as separate documents in the
application, outside of both the site safety analysis report and the
environmental report. The NRC's conclusion is based on the fact that
the requirements in Sec. 50.10(e) address both safety and
environmental issues. Additional changes were made to Sec. Sec. 51.50
and 52.17 to implement this concept.
Some commenters recommended that the requirement in Sec.
52.79(a)(37) to demonstrate how operating experience insights have been
incorporated into the plant design be deleted. The NRC disagrees with
this comment. The NRC developed this requirement for future plants (see
SRM on SECY-90-377) and it was implemented in past design certification
applications by addressing NRC's generic letters and bulletins. The NRC
agrees that insights from generic letters and bulletins should be
incorporated into the latest revision of the standard review plan
(SRP). Therefore, for plant designs that are
[[Page 49387]]
based on or are evolutions of nuclear plants that have operated in the
United States, the applicant should use NRC's generic letters and
bulletins issued after the most recent revision of the applicable SRP
and 6 months before the docket date of the application. If the
application is for a nuclear plant design that is not based on or is
not an evolution of a nuclear plant that operated in the United States,
the applicant should address how insights from any relevant
international operating experience has been incorporated into that
plant.
Section 52.79(a)(41) requires that the applicant evaluate the
facility against the standard review plan (SRP). For COL applicants
that reference the same design certification rule and adopt a design-
centered approach in preparing their COL applications, the NRC expects
that the ``reference application'' will fully conform with this
requirement and then any follow-on applications will not need to
provide the evaluations for the application information that is
identical to the reference application. The NRC did not require
applicants to evaluate their facility against RG 1.206, ``Combined
License Applications for Nuclear Power Plants.'' However, the NRC
believes that RG 1.206 can provide useful guidance to COL applicants in
preparing their applications and that use of this guidance will
facilitate the NRC's review.
The NRC has moved the requirement that COL applicants submit a
plant-specific PRA that was in proposed Sec. 52.80(a) to a new Sec.
52.79(a)(46) in the final rule based on public comments. In addition,
the NRC has revised the provision to require the applicants submit a
description of their PRA and its results in their COL FSAR. The NRC
agrees with some commenters who believed that applicants should not be
required to submit their complete plant-specific PRA and that, instead,
applicants should only be required to provide a summary description of
the PRA and its results in their FSAR with the understanding that the
complete PRA (e.g., codes) would be available for NRC inspection at the
applicant's offices, if needed. The NRC expects that, generally, the
information that it needs to perform its review of the COL application
from a PRA perspective is that information that will be contained in
applicants' FSAR Chapter 19. The NRC believes that COL application
guidance that the NRC is developing is consistent with the industry
comment in that the staff does not expect the complete PRA to be
included in the COL applicant's FSAR. The guidance focuses on
qualitative description of insights and uses, but also acknowledges
that some quantitative PRA results should be submitted.
Section 52.79(b) describes the variant on the requirements in Sec.
52.79(a) for a combined license application that references an early
site permit. Former Sec. 52.79(a) did not explicitly require the
application to address whether the terms and conditions specified in
the early site permit under Sec. 52.24 have been or will be met by the
combined license holder, although this is implicit by the inclusion of
any terms and conditions in the early site permit. To remove any
ambiguity in this matter, Sec. 52.79(b)(3) requires that the FSAR
demonstrate that all terms and conditions that have been included in
the early site permit will be satisfied by the date of issuance of the
combined license. The NRC is revising Sec. 52.79(b)(3) in the final
rule based on public comments to add an exclusion for terms and
conditions imposed under Sec. 50.36(b) because such environmental
conditions should be addressed in the environmental report and not in
the final safety analysis report. In addition, the Commission is
revising this paragraph to add a provision that any terms or conditions
of the early site permit that could not be met by the time of issuance
of the combined license must be set forth as terms or conditions of the
combined license. This provision is needed to address terms or
conditions of the early site permit that are related to activities that
will not take place until after issuance of the combined license, such
as construction activities. A similar change is being made to
Sec. Sec. 52.79(d)(3) and (e)(3) for referenced design certifications
and manufacturing licenses.
The NRC is making a revision to the language in proposed Sec.
52.79(b)(1) in the final rule. Proposed Sec. 52.79(b)(1) stated that
the FSAR for a combined license application referencing an early site
permit need not contain information or analyses submitted to the NRC in
connection with the early site permit. This rule language led to a
great deal of discussion both within the NRC and in public meetings on
combined license application guidance as to what the NRC expected to
see in a combined license application that referenced an early site
permit. The NRC has concluded that the FSARs in these combined licenses
applications must either include or incorporate by reference the SSAR
for the early site permit. The SSAR must be included or incorporated
into the COL FSAR to ensure that matters addressed in the SSAR legally
become part of the FSAR upon issuance of the COL. This will also ensure
that the information in the SSAR is subject to control under Sec.
50.59 after issuance of the COL. For these reasons, the NRC is
modifying the language in Sec. 52.79(b)(1) to state that the final
safety analysis report need not contain information or analyses
submitted to the NRC in connection with the early site permit. However,
the final safety analysis report must either include or incorporate by
reference the early site permit site safety analysis report. With this
modification, the NRC intends to convey that the combined license
applicant referencing the early site permit does not need to resubmit,
for NRC review, information or analyses that were already reviewed and
resolved in the early site permit proceeding (such as information
provided in responses to NRC requests for additional information). At
the same time, the NRC's goal is to provide COL applicants clear
guidance as to what the combined license application must contain to be
considered complete. For similar reasons, the NRC is also modifying the
language in proposed Sec. Sec. 52.79(c)(1), (d)(1), and (e)(1) to
include the provision that the FSAR in the COL application must either
include or incorporate by reference the FSAR for the design approval,
design certification, or manufacturing license that it is referencing.
Note that each of the existing design certification rules covered in
appendices A through D of part 52 prohibit the use of incorporation by
reference in COL FSARs that reference them. At the time those rules
were issued, the NRC was concerned that the staff would not have easy
access to the final version of the design certification FSAR (i.e.,
DCD) if it were not included in the COL application. The NRC will
continue to put restrictions in individual design certification rules
(and possibly in early site permits, design approvals, or manufacturing
licenses) if it does not have confidence that the safety analysis
reports can be easily accessed by the staff if they are incorporated by
reference in COL applications.
Section 52.79(c) describes the requirements for combined license
applications that reference a standard design approval. Previously, no
guidance was provided regarding a combined license application that
referenced a standard design approval. The requirements in Sec.
52.79(c) are essentially the same as those for a combined license
application that references a standard design certification in Sec.
52.79(d).
Section 52.79(d) describes the requirements for combined license
applications that reference a standard
[[Page 49388]]
design certification. Section 52.79(d) states that the FSAR for a
combined license application referencing a standard design
certification need not contain information or analyses submitted to the
NRC in connection with the design certification. However, the final
safety analysis report must either include or incorporate by reference
the standard design certification final safety analysis report (see
discussion above) and must contain, in addition to the information and
analyses otherwise required, information sufficient to demonstrate that
the characteristics of the site fall within the site parameters
specified in the design certification. In addition, paragraph (d)
requires that the plant-specific PRA information must use the PRA
information for the design certification and must be updated to account
for site-specific design information and any design changes or
departures. In the case where a COL application is referencing a design
certification, the NRC only expects the design changes and differences
in the modeling (or its uses) pertinent to the PRA information to be
addressed to meet the submittal requirement of Sec. 52.79(d)(1).
Section 52.79(d) also requires that the FSAR demonstrate that the
interface requirements established for the design under Sec. 52.47
have been met and that all requirements and restrictions that may have
been set forth in the referenced design certification rule be satisfied
by the date of issuance of the combined license.
Section 52.79(e) describes the requirements for a combined license
application that references a manufactured reactor. Previously, no
guidance was provided regarding a combined license application that
referenced a manufactured reactor. These requirements are similar to
those for the content of an FSAR for a combined license referencing a
design certification. Specifically, Sec. 52.79(e) states that the FSAR
need not contain information or analyses submitted to the NRC in
connection with the manufacturing license. However, the final safety
analysis report must either include or incorporate by reference the
manufacturing license final safety analysis report and must contain, in
addition to the information and analyses otherwise required,
information sufficient to demonstrate that the site characteristics
fall within the site parameters specified in the manufacturing license.
This language was slightly different in the proposed rule and has been
corrected in the final rule to be consistent with Sec. 52.79(d). In
addition, Sec. 52.79(e) requires that the plant-specific PRA
information must use the PRA information for the manufactured reactor
and must be updated to account for site-specific design information and
any design changes or departures. Section 52.79(e) also requires that
the FSAR demonstrate that the interface requirements established for
the design have been met and that all terms and conditions that have
been included in the manufacturing license be satisfied by the date of
issuance of the combined license.
Section 52.80 is added to cover the required technical contents of
a combined license application that are not contained in the FSAR.
These application contents include the ITAAC, the environmental report,
and the request to perform activities under Sec. 50.10(e) with the
associated redress plan. This last item was moved to Sec. 52.80(c) in
the final rule from its location in Sec. 52.79(a)(23) in the proposed
rule. The NRC concluded that it is preferable to include both the list
of proposed activities and the redress plan as separate documents in
the application, outside of both the site safety analysis report and
the environmental report. The NRC's conclusion is based on the fact
that the requirements in Sec. 50.10(e) address both safety and
environmental issues. Additional changes were made to Sec. Sec. 51.50
and 52.17 to implement this concept.
g. Section 52.81, Standards for Review of Applications
10 CFR parts 54 and 140 are added to the list of standards that the
NRC will use to review combined license applications. Part 54 addresses
applications for renewal of combined licenses and part 140 includes the
requirements applicable to nuclear reactor licensees with respect to
financial protection and Indemnity Agreements to implement Section 170
of the AEA, commonly referred to as the Price-Anderson Act.
h. Section 52.83, Finality of Referenced NRC Approvals; Partial Initial
Decision of Site Suitability
The former Sec. 52.83, Applicability of part 50 provisions, is
removed and replaced by a new section addressing the finality of NRC
approvals which are referenced in a combined license application.
Former Sec. 52.83 provides that, unless otherwise specifically
provided for in subpart C to part 52, all provisions of 10 CFR part 50
and its appendices applicable to holders of construction permits for
nuclear power reactors also apply to holders of combined licenses.
Similarly, Sec. 52.83 provides that all provisions of 10 CFR part 50
and its appendices applicable to holders of operating licenses also
apply to holders of combined licenses issued under this subpart, once
the Commission has made the findings required under Sec. 52.99. The
NRC believes that the former Sec. 52.83 is not necessary because this
proposed rulemaking will provide conforming changes throughout 10 CFR
part 50 (as well as all other parts in Title 10 Chapter I) to identify
which requirements are applicable to combined license applicants and
holders. Former Sec. 52.83 also provides provisions that address the
duration of a combined license and these provisions would be moved to
proposed Sec. 52.104, Duration of combined license.
The new Sec. 52.83 states that, if an application for a combined
license references an early site permit, design certification rule,
standard design approval, or manufacturing license, the scope and
nature of matters resolved for the application and any combined license
issued are governed by the relevant provisions addressing finality,
including Sec. Sec. 52.39, 52.63, 52.98, 52.145, and 52.171. This
provision clarifies the relationship between a combined license
application and any other license or regulatory approval that an
applicant may reference in the combined license application as far as
issue resolution is concerned.
i. Section 52.89, Environmental Review
Section 52.89 is removed and reserved for future use. Former Sec.
52.89 required that, if a combined license application references an
early site permit or a certified standard design, the environmental
review must focus on whether the design of the facility falls within
the parameters specified in the early site permit and any other
significant environmental issue not considered in any previous
proceeding on the site or the design. Former Sec. 52.89 further stated
that, if the application does not reference an early site permit or a
certified standard design, the environmental review procedures set out
in 10 CFR part 51 must be followed, including the issuance of a final
environmental impact statement, but excluding the issuance of a
supplement under Sec. 51.95(a). This provision is removed because the
requirements for compliance with NEPA are now captured in Sec.
52.79(a) and in the revisions to part 51.
[[Page 49389]]
j. Section 52.91, Authorization To Conduct Site Activities
Section 52.91(a)(2) formerly provided requirements for a combined
license application that does not reference an early site permit, but
that contains a site redress plan and states that the applicant may not
perform the site preparation activities allowed by 10 CFR 50.10(e)(1)
without first submitting a site redress plan in accordance with Sec.
52.79(a)(3), and obtaining the separate authorization required by 10
CFR 50.10(e)(1). This provision further states that authorization must
be granted only after the presiding officer in the proceeding on the
application has made the findings and determination required by 10 CFR
50.10(e)(2), and has determined that the site redress plan meets the
criteria in Sec. 52.17(c). This provision is amended to state that
authorization may [emphasis added] be granted only after the presiding
officer in the proceeding on the application has made the findings and
determination required by 10 CFR 50.10(e)(2), and has determined that
the site redress plan meets the criteria in Sec. 52.17(c). This
amendment is consistent with Sec. 52.91(a)(3), which states that
authorization to conduct the activities described in 10 CFR
50.10(e)(3)(i) may be granted only after the presiding officer in the
combined license proceeding makes the additional finding required by 10
CFR 50.10(e)(3)(ii). The NRC believes that may is the proper term to
use in both of these provisions, to reflect the NRC's residual
authority to decline to authorize the ESP holder to conduct Sec.
50.10(e)(3)(i) activities, even if the NRC's regulations are met.
k. Section 52.93, Exemptions and Variances
Paragraph (a) of Sec. 52.93, which includes a discussion of the
requirements regarding requests for an exemption from any part of a
referenced design certification, is revised to state that the
Commission may grant the request if it determines that the exemption
complies with any exemption provisions of the referenced design
certification rule, or with Sec. 52.63 if there are no applicable
exemption provisions in the referenced design certification rule. This
provision formerly referred to compliance with Sec. 50.12(a). The NRC
is revising paragraph (b) of this section in the final rule to include
an allowance for applicants to request a variance from the early site
permit SSAR. The allowance for requesting variances to the SSAR was
inadvertently omitted in the proposed rule. Because the majority of the
early site permit information that a combined license applicant will be
referencing will be the information in the SSAR, it is logical that the
allowance to request variances be extended to the information in the
SSAR given that the NRC is allowing variances to the permit itself. In
the final rule, the NRC is also adding a provision to paragraph (b) of
this section that precludes the NRC from issuing a variance once a
construction permit, operating license, or combined license referencing
the early site permit is issued; any changes that would otherwise
require a variance should instead be treated as an amendment to the
construction permit or combined license.
Section 52.93 is also revised in the final rule to add a discussion
of requests for departures from a referenced nuclear power reactor
manufactured under a manufacturing license in new paragraph (c) of this
section. This provision was inadvertently omitted in the proposed rule,
although similar provisions were addressed in the proposed rule in
Sec. Sec. 52.98 and 52.171. However, the proposed rule incorrectly
used the term ``variance'' to describe an application-specific change
to a reactor manufactured under a manufacturing license. The NRC has
corrected these provisions in the final rule to use the term
``departure'' for such changes, consistent with the terminology used
for changes to a referenced design certification. New paragraph (c) of
this section is consistent with these other sections and states that an
applicant for a combined license who has filed an application
referencing a nuclear power reactor manufactured under a manufacturing
license may include in the application a request for a departure from
one or more design characteristics, site parameters, terms and
conditions, or approved design of the manufactured reactor. The NRC may
grant a request only if it determines that the departure will comply
with the requirements of 10 CFR 52.7, and that the special
circumstances outweigh any decrease in safety that may result from the
reduction in standardization caused by the departure. The criteria for
granting the departure is the exemption criterion in Sec. 52.7;
however, the departure itself is not considered an exemption (unless,
of course, the departure also involves a non-compliance with an
underlying Commission regulatory requirement in 10 CFR Chapter I).
Thus, the Commission will not approve a departure unless the Commission
finds, in addition to the routine exemption criteria in Sec. 52.7,
that special circumstances outweigh any decrease in safety that may
result from the reduction in standardization caused by the departure.
These limitations are intended to maintain the standardization of
manufactured reactors in operation to the extent practicable. The
licensee may not depart from the design characteristics, site
parameters, terms and conditions, or approved design of the
manufactured reactor through the provisions of Sec. 50.59.
Finally, the provision contained in paragraph (c) of this section
in the 2006 proposed rule (and in paragraph (b) in the former rule) has
been moved to paragraph (d) of this section in the final rule. This
provision states that issuance of a variance under paragraph (b) or a
departure under paragraph (c) is subject to litigation during the
combined license proceeding in the same manner as other issues material
to that proceeding.
l. Section 52.97, Issuance of Combined Licenses
The NRC has modified Sec. 52.97 to be more consistent with the
parallel provision in Sec. 50.50, Issuance of licenses and
construction permits, by including requirements that, after conducting
a hearing and receiving the report submitted by the ACRS, the NRC finds
that there is reasonable assurance that the applicant is technically
and financially qualified to engage in activities authorized; and that
issuance of the license will not be inimical to the common defense and
security or to the health and safety of the public. Section 52.97(c) is
added, consistent with Sec. 50.50, which states that a combined
license shall contain conditions and limitations, including technical
specifications, as the NRC deems necessary and appropriate. Former
Sec. 52.97(b)(2) is moved to new Sec. 52.98 because the issues
addressed in this section are issues associated with finality of
combined license provisions.
m. Section 52.98, Finality of Combined Licenses; Information Requests
Section 52.98, which addresses the finality associated with the
issuance of combined licenses, is added to subpart C of part 52,
consistent with the other subparts in 10 CFR part 52. Section 52.98(a)
states that, after issuance of a combined license, the Commission may
not modify, add, or delete any term or condition of the combined
license, the design of the facility, the inspections, tests, analyses,
and acceptance criteria contained in the license which are not derived
from a referenced standard design certification or manufacturing
[[Page 49390]]
license, except in accordance with the provisions of Sec. Sec. 52.103
or 50.109, as applicable.
Section 52.98 includes provisions to clarify the applicability of
the change processes in 10 CFR part 50 and Section VIII of the design
certification rules in 10 CFR part 52 to a combined license. Section
52.98(b) states that the change processes in 10 CFR part 50 apply to a
combined license that does not reference a design certification rule or
a reactor manufactured under a manufacturing license. Section 52.98(c)
states that the change processes in Section VIII of the design
certification rules apply to changes within the scope of the referenced
certified design. However, if the proposed change affects the design
information that is outside of the scope of the design certification
rule, the part 50 change processes apply unless the change also affects
the design certification information. For that situation, both change
processes may apply.
Section 52.98(d) is added to address changes to a combined license
that references a reactor manufactured under a manufacturing license.
Section 52.98(d)(1) states that, if the combined license references a
reactor manufactured under a subpart F manufacturing license, then
changes to or departures from information within the scope of the
manufactured reactor's design are subject to the change processes in
Sec. 52.171. Note that the proposed rule incorrectly used the term
``variance'' to describe an application-specific change to a reactor
manufactured under a manufacturing license. The NRC has corrected this
provision in the final rule to use the term ``departure'' for such
changes, consistent with the terminology used for changes to a
referenced design certification. Section 52.98(d)(2) states that
changes that are not within the scope of the manufactured reactor's
design are subject to the applicable change processes in 10 CFR part 50
(e.g., Sec. Sec. 50.54, 50.59, and 50.90). The NRC made all of these
requirements to clarify, in one location, the finality provisions
applicable to all portions of a combined license.
Finally, the NRC has added a new paragraph (g) to the ``finality''
section in each subpart of part 52, including Sec. 52.98, entitled
``Information requests,'' which delineates the restrictions on the NRC
for information requests to the holder of the combined license. This
provision is analogous to the former provision on information requests
in paragraph 8 of appendix O to parts 50 and 52, and is based upon the
language of Sec. 50.54(f). For combined licenses, this proposed
provision is in Sec. 52.98(g), and requires the NRC to evaluate each
information request of the holder of a combined license to determine
that the burden imposed by the information request is justified in
light of the potential safety significance of the issue to be addressed
in the information request. The only exception is for information
requests seeking to verify compliance with the current licensing basis
of the facility. If the request is from the NRC staff, the request will
first have to be approved by the EDO or his or her designee.
n. Section 52.103, Operation Under a Combined License
Section 52.103(g) formerly required the NRC to find that the
acceptance criteria in the combined license are met before operation of
the facility, but did not refer to loading of fuel. However, Sec.
52.103(f) stated that fuel loading and operation under the combined
license will not be affected by the granting of a petition to modify
the terms and conditions of the combined license unless a Commission
order is made immediately effective. In the proposed rule, this section
was amended to require the NRC to find that the acceptance criteria in
the combined license are met before fuel load and operation of the
facility. The NRC has decided not to adopt the proposed rule language
which would have precluded loading of fuel into the reactor until
acceptance criteria have been met. The NRC believes that the rule
should reflect, as closely as possible, the statutory requirement in
Section 185.b of the AEA. The NRC has historically viewed ``operation''
as including loading of fuel into the reactor, however it is not
necessary to change the language of Sec. 52.103(g) to continue the
historical practice. The NRC believes that this is the common
interpretation of Sec. 52.103(g).
o. Section 52.104, Duration of Combined License; Sec. 52.105, Transfer
of Combined License; Sec. 52.107, Application for Renewal; Sec.
52.109, Continuation of Combined License; and Sec. 52.110, Termination
of License
Five new provisions are added to subpart C of part 52 for
consistency with the other subparts in 10 CFR part 52 and to parallel
requirements in 10 CFR part 50 for operating licenses. Section 52.104,
addresses the duration of a combined license and contains requirements
that formerly existed in Sec. 52.83. In addition, the Commission has
amended these requirements to indicate that, where the Commission has
allowed operation under a combined license during an interim period
under Sec. 52.103(c), the period of operation is not to exceed 40
years from the date allowing operation during the interim period.
Section 52.105 provides requirements for the transfer of a combined
license that refer the applicant to Sec. 50.80. Section 52.107
provides a reference to 10 CFR part 54 for the renewal of a combined
license.
Section 52.109 provides provisions for the continuation of a
combined license and Sec. 52.110 would provide requirements for the
termination of a combined license. Formerly, part 52 did not address
decommissioning of combined licenses (reactors that are manufactured
under a part 52 manufacturing license do not raise decommissioning
concerns until they are emplaced at a site, inasmuch as a manufacturing
license does not permit loading of fuel or operation) and the
termination of the combined license. By contrast, Sec. Sec. 50.51 and
50.82 address the permanent shutdown of a nuclear power plant, its
decommissioning, and the termination of the part 50 operating license.
There are two possible ways of addressing this omission: Sec. Sec.
50.51 and 50.82 could be modified to reference combined licenses under
part 52, or the provisions analogous to these sections could be added
to part 52. The NRC believes that the second alternative is the best
approach. The combined license holder's responsibilities upon
expiration of its license is more a matter of regulatory authority and
therefore is best placed in part 52. While the question is closer with
respect to decommissioning, the NRC believes that most users would
likely turn to part 52 rather than part 50 to determine the
requirements for decommissioning, inasmuch as decommissioning involves
questions of both procedure and technical requirements.
9. Subpart D, Reserved
10. Subpart E, Standard Design Approvals (Sec. Sec. 52.131 Through
52.147)
The former appendix O to part 52 set forth the requirements for NRC
staff approval of a standard design for a nuclear plant or a major
portion of a nuclear plant. This licensing process was first adopted by
the NRC in 1975 and has been used many times, including issuance of
four final design approvals (FDAs) under appendix O to part 52 from
1994 through 2004. These FDAs were issued during previous design
certification reviews when FDAs were a prerequisite to certification of
a standard plant design (see SOC
[[Page 49391]]
discussion on 10 CFR 52.43 in this document).
When the NRC adopted part 52 in 1989, the Commission did not re-
examine the regulatory scheme for standard design approvals to
determine if the bases for adopting part 52 and the licensing processes
codified in part 52 would also be an impetus for reorganizing the
design approval process. However, the Commission did undertake a re-
examination of appendix O to part 52 in the 2003 proposed rule and
proposed certain changes. In view of the substantial reorganization and
rewriting of part 52 in this rulemaking, the Commission gave further
consideration to the licensing process in appendix O to part 52 and has
made additional changes to enhance the regulatory effectiveness and
efficiency of that licensing process.
The Commission continues to believe that the best approach for
obtaining early resolution of design issues is through the design
certification process in subpart B of part 52. Design certification
will provide greater finality and standardization than the design
approval process. Consequently, the Commission favors use of the design
certification process, which suggests that the design approval process
could be eliminated. However, given the frequent use of appendix O to
part 52 in the past, the Commission has decided to retain this process
and to reorganize and reformat the design approval process to be
consistent with other subparts.
The design approval process, formerly located in appendix O to part
52, has been moved to subpart E of part 52 and reformatted to be
consistent with other subparts. A new Sec. 52.133 was created to
describe the relationship of the design approval process with other
subparts. An FDA may be referenced in an application for a construction
permit or operating license under part 50 or a design certification,
combined license, or manufacturing license under part 52.
The filing requirements for design approvals are consistent with
other subparts of part 52. The applicants may still request approval of
either the entire facility or major portions thereof, but the
applications are limited to final design information. There are several
reasons for this change. First, the Commission's recent experience with
FDAs and design certifications demonstrates that nuclear plant
designers are technically capable of developing essentially complete
and final design information for NRC review and approval. Furthermore,
the economic incentives with respect to design certification also apply
to final design approvals. In addition, approval of final design
information removes the unpredictability of issuing a construction
permit that references only preliminary design information and
initiating construction while the final design information is being
developed. Approval of a final design ensures early consideration and
resolution of technical matters before there is any substantial
commitment of resources associated with the construction of the plant,
which will greatly enhance regulatory stability and predictability.
The Commission has decided that the contents of applications for
design approvals should contain essentially the same technical
information that is required of design certification applications
(e.g., demonstration of compliance with technically relevant Three Mile
Island requirements, proposed technical resolutions of unresolved
safety issues and medium- and high-priority generic safety issues, and
design-specific probabilistic risk assessment information).
Regarding applications for a major portion of the standard plant
design, such as the nuclear steam supply system, the application only
needs to contain the information required for the contents of
applications that are applicable to the major portion of the plant for
which NRC staff approval is requested.
The requirements for contents of applications for design approvals
(Sec. 52.137) were renumbered to be consistent with the numbering of
requirements in Sec. 52.47. Also, many of the public comments on
contents of applications for design certification apply to the
requirements for design approvals (see the SOC of this document for the
discussion for Sec. 52.47). Some commenters recommended that the
requirement for coping with emergencies [Sec. 52.137(a)(11)] be
deleted because applicants for design approvals will not be responsible
for certain emergency planning design features. The Commission
disagrees with this comment. This requirement was taken from the
original appendix O of part 52, paragraph 3, and it applies to design
features for coping with emergencies in the operation of the reactor,
not for emergency planning.
A new Sec. 52.139, which specifies the standards that will be used
to review applications for design approvals and new Sec. Sec. 52.145
and 52.147, which specify the finality and duration of design approvals
was added to be consistent with other subparts. In a letter dated
November 13, 2001, NEI commented that ``Industry recommends FDAs be
valid for 15 years.'' The Commission agrees with NEI's recommendation
and has decided that the duration of standard design approvals should
correspond to the duration of design certifications, inasmuch as both
design approvals and design certifications constitute approvals of
nuclear power plant designs, and the period of effectiveness of the
approval from a technical standpoint is not a function of whether the
approval is granted by the NRC staff or the Commission. Some commenters
recommended that Sec. 52.147 be rewritten to provide for renewals of
standard design approvals. The Commission disagrees with this comment.
The original appendix O to part 52 did not contain a process for
renewing design approvals and most of the design approvals issued under
appendix O to part 52 were for a 5-year duration. In this rulemaking,
the Commission has tripled the duration for a design approval and
believes that renewals will not be necessary. Also, as stated before,
the Commission favors the use of the design certification process,
which includes a process for renewals.
11. Subpart F, Manufacturing Licenses
The following discussion explains the requirements in subpart F of
part 52 generically, and covers Sec. Sec. 52.151, 52.153, 52.155,
52.156, 52.157, 52.159, 52.161, 52.163, 52.165, 52.167, 52.169, 52.171,
52.173, 52.175, 52.177, 52.179, and 52.181.
Former appendix M of parts 50 and 52 set forth the NRC's
requirements governing manufacturing licenses. Appendix M, which was
first adopted by the NRC in 1973 as an appendix to part 50, provided
for issuance of a license authorizing the manufacture of a nuclear
power reactor to be incorporated into a nuclear power plant under a
construction permit and operated under an operating license at a
different location from the place of manufacture. Under the licensing
regime in former appendix M, the NRC did not approve a final reactor
design to be manufactured as part of the issuance of the manufacturing
license. Rather, analogous to the two-step construction permit/
operating license process, the NRC would issue a manufacturing license
based upon the review and approval of a preliminary design equivalent
to that provided in a construction permit application. Upon issuance of
the manufacturing license, manufacturing of the reactor can commence,
although the NRC must approve the final design of the manufactured
reactor by license amendment before the manufactured reactor may be
transported from the
[[Page 49392]]
place of manufacture to the site where it is to be operated.
When the NRC adopted part 52 in 1989, it added appendix M to part
52. However, the NRC did not re-examine the regulatory scheme for
manufacturing licenses in order to determine if the bases for adopting
part 52 would also be an impetus for changing the regulatory scheme for
manufacturing licenses. Nor did the NRC undertake such a re-examination
as part of the process leading to the 2003 proposed rule. However, the
NRC has reconsidered the efficacy of the manufacturing license process
in former appendix M to part 52, and has decided to adopt substantial
changes to those requirements in order to enhance regulatory
effectiveness and efficiency. These new requirements are contained in a
new subpart F to part 52.
The most important shift in the manufacturing license concept in
subpart F is that a final reactor design, equivalent to that required
for a standard design certification under part 52 or an operating
license under part 50, must be submitted and approved before issuance
of a manufacturing license. There are several reasons for this shift.
First, the Commission's experience with standard design certifications
demonstrates that nuclear power plant designers are technically capable
of developing a complete reactor design for Commission review.
Furthermore, the economic incentives and limitations with respect to
approval of a standard reactor design certification also apply to the
approval of a design of a manufactured reactor. Indeed, one could argue
that the holder of a manufacturing license may structure the commercial
transaction to reduce the economic risk associated with the application
for a manufacturing license for a final reactor design, as compared to
the economic risk associated with a standard design certification.
Second, approval of a final reactor design removes the former awkward
regulatory process of issuing a manufacturing license, and subsequently
amending the license when a final design is submitted. Approval of a
final design ensures early consideration and resolution of technical
matters before there is any substantial commitment of resources
associated with the actual manufacture of the reactor, which will
greatly enhance regulatory stability and predictability. Finally,
Commission approval of standardized manufacturing processes, coupled
together with the potential for a stable workforce and the application
of manufacturing process feedback, has great opportunities for
maintaining and even improving the quality and consistency of
manufacture, as compared to the traditional method of constructing
reactors onsite by a variety of contractors and subcontractors.
The technical information required to be included in an application
for a manufacturing license, as set forth in Sec. Sec. 52.157 and
52.158, reflects both the expansion of the scope of approval to include
the final design of the reactor to be manufactured, as well as lessons
learned with respect to the NRC's review of early site permits. Section
52.157, which sets forth the technical information to be submitted in
support of the design of a reactor, is derived from the existing
requirements in current part 52, subparts B and C, governing the
technical information to be submitted in support of an application for
a standard design certification and combined license. In addition,
Sec. 52.157 requires that the application address the provisions with
respect to the demonstration by test, analysis, experience, or a
combination thereof, of simplified, inherent, passive, or other
innovative means to accomplish safety functions, or the results of
testing of a prototype plant, as set forth in revisions to Sec. 50.43.
As discussed separately with respect to Sec. 50.43, these testing and
prototype requirements incorporated into Sec. 50.43 were derived from
the former requirements in Sec. 52.47(b).
Information which must be submitted as part of an application, but
is not typically considered part of a final safety analysis report, is
identified in Sec. 52.158. This includes proposed ITAAC to be used by
the licensee who will construct and operate a nuclear power plant at
its site using the manufactured reactor and an environmental report for
the manufactured reactor. Note that, in the final rule, the NRC has
moved proposed Sec. 52.158(a) to a new Sec. 52.157(f)(31) which
requires that manufacturing license applicants submit a description of
the design-specific PRA and its results in the FSAR. The NRC agrees
with some commenters that applicants should not be required to submit
their complete design-specific PRA and that, instead, applicants should
only be required to provide a summary description of the PRA and its
results in their FSAR with the understanding that the complete PRA
(e.g., codes) would be available for NRC inspection at the applicant's
offices, if needed. The NRC expects that, generally, the information
that it needs to perform its review of the manufacturing license
application from a PRA perspective is that information that will be
contained in applicants' FSAR Chapter 19.
The environmental report must address SAMDAs, similar to standard
design certifications, because the design approval stage is usually the
most cost-effective opportunity for incorporating design features for
addressing severe accidents. The NRC notes that the environmental
report need not address environmental impacts associated with the
actual manufacture of the reactor at any manufacturing location,
inasmuch as a manufacturing license does not represent NRC approval of
any specific location, facility, or appurtenance for manufacturing.
Rather, the NRC is approving a reactor design for manufacture and the
ITAAC for verifying that it has been acceptably manufactured and
integrated into a nuclear power facility so that it can be safely
operated in accordance with the approved manufactured reactor design,
the NRC's regulations, and the requirements of the AEA. These
determinations were reflected in proposed Sec. Sec. 52.158(c)(1),
51.54, and 51.75(c)(3). However, in the final rule, the Commission has
removed from proposed Sec. Sec. 52.158(c)(1) and (2) (final Sec. Sec.
52.158(b)(1) and (2)) the rule language addressing the content of the
environmental report, and integrated that language into Sec. Sec.
51.54 and 51.75(c)(3). Proposed Sec. 52.158(c)(2) (final Sec.
52.158(b)(2)) has been revised in the final rule to address the scope
of the environmental report if the manufacturing license application
has referenced a standard design certification.
Section 52.163 of the March 2006 proposed rule would have required
that the NRC conduct a ``mandatory'' hearing in connection with the
initial issuance of a manufacturing license, even though the AEA does
not require a mandatory hearing for issuance of manufacturing licenses.
For the reasons set forth in the NRC's response to Commission Question
2, and the discussion on Sec. Sec. 2.104 and 2.105, the NRC has
decided not to require a ``mandatory'' hearing for initial issuance of
a manufacturing license, and Sec. 52.163 is revised in the final rule
to refer to a publication of a notice of proposed action under Sec.
2.105, rather than a notice of hearing under Sec. 2.104.
In light of the NRC's review and approval of a final design as part
of issuance of a manufacturing license, the final rule provides a
greater degree of finality to a manufacturing license as compared with
a standard design certification. Under Sec. 52.171(a)(1), the same
degree of issue finality accorded to the ``certified design'' applies
throughout the term of the manufacturing license. Under this
[[Page 49393]]
provision, the NRC may not impose any change or modification to the
approved design (including site parameters, or design characteristics)
for the manufacturing license unless the NRC determines that the change
or modification is necessary either for adequate protection or for
compliance with requirements applicable and in effect at the time the
manufacturing license was issued. Similarly, the manufacturing license
holder may not make changes to the design under the provisions of 10
CFR 50.59. Any change to the design will require a license amendment.
The Commission regards this as similar to the level of change control
imposed on designs which are the subject of a standard design
certification. The Commission is imposing this stringent level of
change control because one of the key reasons for licensing
manufactured reactors is to enhance standardization--one of the
original objectives of the 1989 part 52 rulemaking. Unlike design
certification, which is an approval of a ``paper design,'' the NRC's
proposed concept of a manufacturing license is pre-approval of the
procurement, manufacturing, and quality assurance processes that
translates the approved reactor design into a manufactured assembly in
a controlled environment, with the capability to optimize techniques
and procedures based upon feedback. Some of these advantages may be
lost if each ``manufactured'' reactor were treated as a ``one-off''
custom product. Imposing the discipline of a license amendment process
should ensure that a profusion of changes are not made to the approved
design at random intervals. The Commission disagrees with commenters on
the proposed rule that the design of a manufactured reactor should be
subject to less-stringent change provisions than a standard design
certification. The commenters have not demonstrated that there are
special or unique aspects of manufacturing, as compared with the
construction of a nuclear power plant based upon a referenced standard
design certification, that would weigh against maintaining the high
degree of design standardization achieved by design certification. One
commenter correctly noted that changes in such manufacturing matters as
procurement, manufacturing processes, or quality assurance are not
subject to the proposed Sec. 52.171(b)(1) change restriction, because
these matters do not constitute changes to the approved design of the
reactor to be manufactured. These changes would be governed by the
applicable change process and restrictions already established in the
Commission's regulations such as Sec. 50.59, and Sec. 50.54(a), and
may not require license amendments.
The only relevant rationale provided by the commenters is that
obsolescence of components and component manufacturers' changes would
necessitate minor changes to the reactor design over a 15-year period.
Although the Commission acknowledges the likelihood of these factors,
the NRC staff does not see any reason why these factors are more likely
to affect the design of a manufactured reactor as compared with the
design approved in a design certification. It is not clear why a change
in component sourcing would necessarily result in a ``design change''
requiring an amendment to the manufacturing license. Finally, the
Commission notes that the proposed rule does not mandate ``zero changes
in a reactor design.'' As specifically stated in the SOC of the March
13, 2006 (71 FR 12801), proposed rule (second column), proposed Sec.
52.171(b)(1) would allow the manufacturer to make changes to the
approved design to be manufactured, albeit by license amendment.
The final rule provides that the term of a manufacturing license to
be for no less than 5, or more than 15 years from the date of issuance.
The Commission established the 15-year maximum term to be consistent
with the maximum term for a standard design certification. The 5-year
minimum term was established by the Commission to encourage the use of
a manufacturing license for the manufacture of more than one nuclear
power reactor. The language of Sec. 52.171 has been corrected in the
final rule by replacing the reference in paragraph (b)(1) to Sec.
50.12 with a reference to Sec. 52.7, and replacing the term,
``exemption,'' in paragraph (b)(2) with ``departure.''
In proposed Sec. 52.167(b)(3), the Commission included a provision
which would have required the manufacturing license to specify the
number of reactors authorized to be manufactured under the
manufacturing license. Upon further consideration in response to a
comment on the proposed rule, the Commission has decided that there is
no valid regulatory basis for including this provision, and it may in
fact serve as a disincentive for the manufacturer to improve the
efficiency and productivity of the manufacturing process. Accordingly,
this provision is not included in the final rule.
Under Sec. 52.177(c), the holder of a manufacturing license may
not commence manufacturing of a reactor less than 3 years before the
expiration date, but may continue the manufacturing of a reactor whose
manufacture commenced before the 3-year deadline up to license
expiration. If, however, an application for renewal is timely-filed
with the NRC, manufacturing of a reactor whose manufacture commenced
before the 3-year deadline may continue until the time that the NRC
completes action on the renewal application in accordance with the
Timely Renewal Doctrine of the Administrative Procedure Act (APA). The
Commission believes that the timely renewal period should be based upon
the time reasonably needed by the agency to complete action on a
renewal application, so that an applicant's reliance upon timely
renewal is the rare exception rather than the rule. The NRC selected
the 3-year deadline as a reasonable period for completing the
manufacture of a nuclear power reactor, based in large part upon public
statements by various reactor vendors that they have set goals for
constructing complete nuclear power plants onsite within 3 years. It
seems reasonable, therefore, that a manufactured reactor, built in a
controlled environment using industrial manufacturing processes, would
be able to be manufactured in the same 3-year period as the
construction of an entire facility onsite. Paragraph (b) is corrected
in the final rule by removing the phrase, ``that the Commission may
impose,'' in order to avoid the possible misinterpretation that the
Commission could choose not to impose new adequate protection
requirements identified by the Commission. In addition, paragraph
(b)(2) is corrected by removing the reference to ``site permit'' and
substituting the term, ``manufacturing license.''
The final rule does not require that the manufacturing license
specify an earliest and latest date for completion of manufacture of
any individual reactor. Section 185 of the AEA directs that ``[t]he
construction permit shall state the earliest and latest date for
completion of the construction or modification.'' Inasmuch as a
manufacturing license is not a construction permit, there does not
appear to be any legal need for the manufacturing license to specify
the earliest and latest date of completion of manufacture. The language
of this section has been corrected in the final rule to make clear that
the duration of the renewed manufacturing license consists of the
renewed term plus any period remaining on the superseded license
(analogous to the determination
[[Page 49394]]
of the duration of a renewed operating license under part 54).
12. Subpart G of Part 52 [Reserved]
13. Subpart H of Part 52--Enforcement
This subpart contains two provisions, Sec. 52.301 and Sec.
52.303, which are comparable to former Sec. 52.111 and Sec. 52.113,
and are analogous to provisions contained in other parts of 10 CFR
Chapter I imposing requirements on regulated entities. Section 52.301
reiterates, and provides notice to licensees and applicants under part
52 of the Commission's authority to obtain injunctions or other court
orders for the enumerated violations. Section 52.113 provides notice to
all persons and entities subject to part 52 that they are subject to
criminal sanctions for willful violations, attempted violations, or
conspiracy to violate certain regulations under part 52. The
regulations listed in paragraph (b), for which criminal sanctions do
not apply, have been updated to reflect the final part 52 rulemaking.
Section 52.99 was erroneously listed in paragraph (b) in the proposed
rule. Because that regulation contains substantive requirements which
are promulgated under Section 161.b., i, and o of the AEA, it has been
removed from the list of regulations in paragraph (b).
14. Appendices A, B, C, and D to Part 52--Design Certifications for
ABWR, System 80+, AP600, and AP1000
The NRC amended paragraphs VI.B.4, 5, and 6 of the design
certification rules (DCRs) in appendices A, B, and C to part 52 for the
U.S. ABWR, System 80+, and AP600 designs, respectively, by substituting
the phrase ``but only for that plant'' for the erroneous phrase ``but
only for that proceeding'' (emphasis added). The new phrase correctly
characterizes the scope of issue resolution in three situations.
Paragraph VI.B.4 describes how issues associated with a DCR are
resolved when an exemption has been granted for a plant referencing the
DCR. Paragraph VI.B.5 describes how issues are resolved when a plant
referencing the DCR obtains a license amendment for a departure from
Tier 2 information. Paragraph VI.B.6 describes how issues are resolved
when the applicant or licensee departs from the Tier 2 information on
the basis of paragraph VIII.B.5, which waives the requirement to obtain
NRC approval for such departures. Thus, once a matter (e.g., an
exemption in the case of paragraph VI.B.4) is addressed for a specific
plant referencing a DCR, the adequacy of that matter for that plant
would not ordinarily be subject to challenge in any subsequent
proceeding or action (such as an enforcement action) listed in the
introductory portion of paragraph IV.B, but there would not be any
issue resolution on that subject matter for any other plant.
Each of the DCRs includes a Section VIII on processes for changes
and departures. These processes apply to changes and departures
depending upon the category of certification information affected. For
plant-specific Tier 2 information, the departure process established in
the rule mirrors, in large part, that in the former 10 CFR 50.59. The
final rule amends paragraph VIII.B.5 of the DCRs in appendices A, B,
and C to conform the terminology in the Sec. 50.59-like process to
that used in the current Sec. 50.59. This amendment deleted references
to unreviewed safety questions and safety evaluations, and conformed
the evaluation criteria concerning when prior NRC approval is needed.
Also, a definition was added to the DCRs (paragraph II.G) for
``departure from a method of evaluation'' to support the evaluation
criterion in paragraph VIII.B.5.b(8) of appendices A, B, and C to part
52.
In an earlier rulemaking (see 64 FR 53582; October 4, 1999), the
NRC revised Sec. 50.59 to incorporate new thresholds for permitting
departures from a plant design as described in the FSAR without NRC
approval. For consistency and clarity, similar changes were adopted for
part 52 applicants or licensees. Because of some differences in how the
requirements are structured in the DCRs, certain criteria contained in
Sec. 50.59 are not necessary for or applicable to part 52 and are not
being included in this rule. One criterion definition that the NRC did
include was from Sec. 50.59 for a ``Departure from a method of
evaluation,'' which is appropriate to include in this rulemaking so
that the eighth criterion in paragraph VIII.B.5.b of appendices A, B,
and C to part 52 will be implemented as intended.
Each of the DCRs includes a special process in Section VIII for
departures from selected severe accident issues. The Commission
believes that the resolution of severe accident issues should be
preserved and maintained in the same fashion as all other safety issues
that were resolved during the design certification review (refer to SRM
on SECY-90-377). However, because of the increased uncertainty in
severe accident issue resolutions, the Commission codified separate
criteria in paragraph B.5.c of Section VIII for determining if a
departure from design information that resolves these severe accident
issues would require a license amendment. The final rule amends
paragraph B.5.c to clarify that the special process applies to ex-
vessel severe accident design features that are described in the plant-
specific design control document (DCD).
For purposes of applying the special criteria in paragraph B.5.c of
Section VIII, severe accident resolutions are limited to those design
features where the intended function of the design feature is relied
upon to resolve postulated accidents when the reactor core has melted
and exited the reactor vessel (ex-vessel severe accidents) and the
containment is challenged. The location of the ex-vessel severe
accident design information in the DCD is not important to the
application of this special departure process in paragraph B.5.c. Some
design features may have intended functions to meet both ``design
basis'' requirements and to resolve ex-vessel severe accidents. If
these design features are reviewed under paragraph VIII.B.5, then the
appropriate criteria from either paragraph B.5.b or B.5.c are selected
depending upon which function the departure is being taken from.
Each of the DCRs in appendices A, B, and C to part 52 includes a
section on records and reporting. The NRC revised paragraph X.B.3.b in
appendices A, B, and C to part 52 to change the reporting frequency
from quarterly to semi-annually, and to extend the period of increased
reporting frequency, relative to the frequency of 10 CFR 50.59(d) and
50.71(e)(4), from the date of a license application that references a
DCR to the date that the Commission makes the finding under 10 CFR
52.103(g). The requirement to report plant-specific departures from,
and updates to, the design control document during the interval from
the application for a combined license until the Commission makes the
finding under Sec. 52.103(g) is to facilitate NRC's monitoring of
changes to the nuclear power plant, to achieve a common understanding
of how the as-built facility conforms to the design information, and to
adjust the inspection program to reflect the design changes.
The amendment to paragraph X.B.3.b of appendices A, B, and C to
part 52 reduced the frequency of reporting during the period of
construction and increased the frequency of reporting during the
application review period. The NRC believes that these changes in the
reporting burden balance each other and provide the information needed
by the NRC to fulfill its responsibilities in the licensing of future
nuclear power plants. In order to make the finding
[[Page 49395]]
under Sec. 52.103(g), the NRC must monitor the design changes made
under Section VIII of the DCRs. Frequent reporting of design changes
will be particularly important in times when the number of design
changes could be significant, such as during the procurement of
components and equipment, the detailed design of the plant before and
during construction, and during pre-operational testing. After the
facility begins operation, the frequency of reporting would revert to
the requirement in paragraph X.B.3.c, which is consistent with
operating plant requirements.
Additional editorial changes to the design certification rule
language in appendices A, B, C, and D to part 52 are discussed in the
NRC's responses to public comments on Question 11 (see Section IV of
this document).
15. Appendix N to Part 52--Combined Licenses for Nuclear Power Reactors
of Identical Design
Prior to this final rulemaking, appendix N in parts 50 and 52
contained the NRC's procedures governing the review and issuance of
licenses for nuclear power plants of ``duplicate design.'' Hearings for
applications filed under appendix N in both parts 50 and 52 are
governed by subpart D of part 2. In the March 2006 proposed rule, the
NRC proposed deleting appendix N in part 52, and retaining these
provisions only in part 50. Although no comment was received on this
proposal, the NRC has decided to withdraw its proposal to delete
appendix N in part 52. Since the preparation of the March 2006 proposed
rule, several industry groups have announced their intention to seek
combined licenses utilizing the same design. In view of this industry
development, the NRC believes that there is potential utility to
keeping the option of appendix N in part 52 open to potential combined
license applicants. Accordingly, the NRC is retaining in part 52 the
procedural alternative provided in appendix N to part 52, and to revise
its language to make its provisions applicable to combined licenses
using identical designs. As part of this revision, the NRC set forth
more explicit direction on the information to be submitted, the NRC
docketing review, notice, and the content of the EIS under appendix N
of part 52. However, the NRC decided against a wholesale revision of
appendix N to part 52, together with conforming changes in part 51,
inasmuch as these changes were not the subject of public comment, and
because such a course of action would have delayed the overall part 52
rulemaking. Inasmuch as the changes to appendix N of part 52
constitute, in essence, revisions to the NRC's rules of procedure and
practice (albeit located within part 52), the NRC may adopt them in
final form without further notice and comment, under the rulemaking
provisions of the APA, 5 U.S.C. 553(b)(A).
The overall concept of the revised appendix N to part 52 is that
each application is to be treated as a separate application, with the
exception of the common design. Hence, appendix N to part 52 requires
separate applications, separate determinations of sufficiency for
docketing, separate notices of docketing, and so forth. Sections
requiring further explanation are discussed below.
Paragraph 2 of appendix N to part 52 requires that each application
state that the applicant wishes to have the application considered
under appendix N to part 52, and to list all of the applications that
are to be treated together. This requirement ensures that the NRC is
clearly informed of the intentions of all applicants, and to ensure
that any individual reviewing the application can easily determine all
of the applications using the identical (``common'') design.
Paragraph 3 of appendix N to part 52 requires that each application
identify the common design, and that the FSAR either incorporate by
reference or include the common design. This ensures that there will be
a single physical FSAR document that may be utilized by the NRC, and
viewed by members of the public.
Paragraph 5 of appendix N to part 52 provides that, upon an NRC
determination that each application is acceptable for docketing under
10 CFR 2.101, each application will be separately docketed (i.e., each
application will be given a separate docket number, but that docket
number may include a special designator signifying that it is part of a
group of applications filed under appendix N to part 52). Ordinarily,
the NRC will publish in the Federal Register a separate notice of
docketing for each application, so that delays in the docketing of one
application will not delay the docketing and subsequent technical
review of other applications filed in accordance with appendix N to
part 52. However, if circumstances allow (e.g., sufficiency review for
multiple applications are completed simultaneously), the NRC may
publish a single notice of docketing for multiple applications. The
notice of docketing must state that the application will be processed
under the provisions of 10 CFR part 52, appendix N and subpart D of
part 2. As discussed under subpart D of part 2, the NRC also has
discretion to either publish a notice of hearing for each application
(possibly with the period for the filing of petitions to intervene
running from the notice of hearing for the last application of the
group), or to publish a joint notice of hearing for multiple
applications.
Paragraph 6 of appendix N to part 52 sets forth the procedures by
which the NRC will fulfill its obligations under NEPA. The NRC staff
will prepare a separate draft EIS for each application, but the NRC may
conduct joint scoping on environmental issues related to the common
design. If the applications reference a standard design certification
or the use of a manufactured reactor, then the EIS must incorporate by
reference the EA prepared for either the design certification or the
manufacturing license, as applicable. The NRC has decided that the EA
need not be included in the EIS. The Commission has required other
documents to be incorporated into the FSAR in order to maximize the
utility and ease of use of the FSAR, which is used repeatedly by the
NRC staff over the lifetime of the licensed reactor. By contrast, the
EIS is not typically utilized by the staff in such a manner; hence, the
NRC deemed it unnecessary to require physical incorporation of the
referenced design certification or manufacturing license EA into the
referencing combined license EIS.
Paragraph 7 of appendix N to part 52 requires the ACRS to report on
each of the combined license applications, as required by Sec. 52.87.
Each ACRS report is to be limited to the safety matters which are not
relevant to the common design. In addition, the ACRS must issue a
report on the safety of the common design--except for those matters
relevant to the safety of a referenced design certification or
manufactured reactor. Issuance of separate reports for each application
will facilitate NRC staff internal review, consideration, and response
to the ACRS report. It will also ensure that issues relevant to one
application (e.g., siting) are not addressed in the proceeding and
hearing for another application. Issuance of a single report on the
common design will also facilitate the issuance of the presiding
officer's partial initial decision on the common design, as required by
paragraph 8 of appendix N to part 52, and 10 CFR 2.405 of subpart D of
part 2. The NRC notes that there may be circumstances where the common
design extends beyond the design matters covered in a referenced design
[[Page 49396]]
certification or manufactured reactor. For example, a common design
could reference the use of a specific design certification and a common
ultimate heat sink. In such circumstances, the ACRS would issue a
common report limited to the safety matters for the ultimate heat
sink.\6\
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\6\ The site-specific environmental impacts of the heat sink
would ordinarily be addressed in each of the separate EISs prepared
for each application, inasmuch as the environmental impacts would
differ depending upon factors and characteristics at each site.
Section 7 does not govern the scope of EISs prepared for common
design elements.
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Paragraph 8 of appendix N to part 52 provides that the NRC will
designate a presiding officer to conduct the portion of the hearing on
matters related to the common design, and that the presiding officer
must issue a partial initial decision on the common design. As
discussed previously, hearing procedures for appendix N to part 52
proceedings are set forth in subpart D to part 2. To avoid duplication
and possible (future) conflicts with subpart D to part 2, the NRC did
not include in appendix N to part 52 further provisions addressing the
conduct of hearings.
D. Changes to 10 CFR Part 50
1. General Provisions, Sec. 50.2, Definitions
New definitions are added as conforming changes to Sec. 50.2. A
definition of an applicant is added to clarify that a person or entity
applying for Commission ``permission or approval'' is an applicant.
This will ensure that part 50 requirements for applicants apply to a
person or entity seeking an NRC approval not constituting a license,
such as a standard design approval under part 52.
Definitions for license and licensee are added to clarify that
early site permits and combined licenses under part 52 are licenses,
and that holders of these types of licenses are licensees for purposes
of part 50.
A definition for prototype plant is added to describe the type of
nuclear reactor that is the subject of Sec. 50.43(e). A prototype
plant is a licensed nuclear reactor test facility that is similar to
and representative of the first-of-a-kind nuclear plant in all features
and size, but may have additional safety features. The purpose of the
prototype plant is to perform testing of new or innovative design
features for the first-of-a-kind nuclear plant design, as well as being
used as a commercial nuclear power facility.
2. Requirement of License, Exceptions, Sec. 50.10, License Required
Section 50.10 addresses the circumstances under which a license for
a production or utilization facility is required, and describes
activities which do not constitute ``construction'' for purposes of
obtaining a license for a nuclear power plant. Section 50.10(b)
formerly prohibited a person from beginning construction of a
production or utilization facility unless a construction permit has
been issued. Inasmuch as activities constituting construction (as
defined in Sec. 50.10(b)) are authorized under a combined license,
Sec. 50.10(b) is revised to refer to combined licenses.
Formerly Sec. 52.17(c) authorized an early site permit applicant
to request authority to perform the activities allowed under Sec.
50.10(e)(1). The NRC notes that the regulation did not provide for the
holder of an early site permit to request authority to conduct Sec.
50.10(e)(1) activities after the early site permit has been issued, and
the NRC does not plan to change the current restriction. It will
conserve the NRC's resources to consider the safety and environmental
issues associated with Sec. 50.10(e)(1) activities during the agency's
consideration of the early site permit application. Late consideration
of these requests after completion of the NRC's consideration of the
application could entail substantial diversion of resources from other
application reviews. For these reasons, the NRC does not allow an early
site permit holder to request authority to perform activities allowed
under Sec. 50.10(e)(1) after issuance of the early site permit (the
Commission notes that under former part 52, early site permit holders
may not seek authority to perform activities allowed under Sec.
50.10(e)(3) after issuance of the early site permit).
3. Classification and Description of Licenses
a. Section 50.23, Construction Permits
Section 50.23 formerly provided that a construction permit for the
construction of a production or utilization facility must be issued
before issuance of a license for the facility, and then only upon ``due
completion'' of the facility. Section 50.23 is revised to clarify that
if the NRC issues a combined license for a nuclear power plant under
part 52, the construction permit and operating license are issued
simultaneously (i.e., are merged into a ``combined license'' under
subpart C of part 52). This is consistent with Section 185.b of the
AEA, which provides the NRC with explicit statutory authority to
combine a construction permit and an operating license for a nuclear
power plant into a single combined license. The Commission notes that
Sec. 50.23 is not limited to nuclear power plants; it also allows the
NRC to combine, under Section 161.h of the AEA, a construction permit
and operating license for production facilities or utilization
facilities other than nuclear power plants.
4. Applications for Licenses, Certifications, and Regulatory Approvals;
Form; Contents; Ineligibility of Certain Applicants
a. Section 50.30, Filing of Application; Oath or Affirmation
Section 50.30 establishes the NRC's general procedural requirements
on filing of applications for licenses (including construction permits)
for production and utilization facilities. The NRC is making conforming
changes throughout Sec. 50.30 to include necessary references to part
52 processes other than design certification (subpart H of part 2
governs the filing of standard design certification applications),
viz., early site permits, combined licenses, standard design approvals,
and manufacturing licenses. In addition, Sec. 50.30(a) is revised to
ensure that the submission requirements governing applications (and
amendments to these applications) in Sec. 52.3 apply to part 52
processes other than design certification.
b. Section 50.33, Contents of Applications; General Information
Section 50.33 identifies the general information that must be
included in applications for licenses (including construction permits)
for production and utilization facilities. Section 50.33(f) requires
certain applicants for nuclear power plant licenses to submit
information sufficient to determine whether the applicant has the
financial qualifications to carry out, in accordance with the NRC's
regulations, the activities for which a license or permit is sought.
Section 50.33 is revised to require applicants for combined licenses to
submit financial qualifications information. Financial qualifications
information need not be submitted by applicants for early site permits,
standard design certifications, standard design approvals, and
manufacturing licenses. An NRC review to determine whether an applicant
has adequate financial qualifications to conduct the activities
authorized by an early site permit would contribute little, if
anything, to providing reasonable assurance of adequate protection with
respect to early site permit activities. Ordinarily, an early site
permit authorizes no activities, unless the early site permit
application requested
[[Page 49397]]
authority to conduct the activities permitted under Sec. 50.10(e)(1).
The NRC has determined that no safety finding per se is necessary to
authorize the licensee to conduct these activities. The NRC's review of
a Sec. 50.10(e)(1) application is focused on siting and environmental
matters.
With respect to a standard design approval, the argument applies
with even more force, inasmuch as a design approval authorizes no
activities of any kind, and the finality associated with a design
approval is significantly less than for an early site permit. The NRC
concludes that no regulatory purpose appears to be served by a
financial qualifications review for early site permits and standard
design approvals. The NRC believes that there is little additional
regulatory value in requiring a financial qualifications review for a
manufacturing license. While it is true that a lack of sufficient
financial resources could result in inadequate manufacture of a
reactor, under the NRC's proposed concept of a manufacturing license
under subpart F of part 52, each manufactured reactor cannot be
operated until ITAAC specified in the manufacturing license are
successfully completed by the licensee authorized to construct the
nuclear power facility using the manufactured reactor. Successful
completion of the manufactured reactor's ITAAC should ensure that any
problems with manufacture attributable to lack of financial resources
of the manufacturing license holder can be identified before operation.
Moreover, the licensee authorized to construct the facility (either
under a construction permit or a combined license) using a manufactured
reactor would have been subject to a financial qualifications review.
This review should be sufficient to determine if the applicant has
sufficient financial resources to carry out facility construction and
the completion of the manufactured reactor's inspections, tests, and
acceptance criteria. Finally, the NRC notes that it does not require
the fabricators of safety-related and important to safety structures,
systems, and components (SSCs) to be licensed and subject to a
financial qualifications review. The NRC believes that a holder of a
manufacturing license conducts activities which appear to be, in large
part, analogous to these current non-licensed fabricators. Accordingly,
the NRC concludes that a financial qualifications review of the
applicant for a manufacturing license will not add significant
regulatory value to justify the cost of such a review.
Section 50.33(g) addresses radiological emergency response plans
for State and local government entities that must be submitted in
applications for operating licenses. The final rule makes a conforming
change to ensure that applicants for combined licenses must also submit
this information, as well as applicants for early site permits who
decide under Sec. 52.17(b)(2)(ii) to seek NRC review and approval of
complete emergency plans. In addition, Sec. 50.33(g) provides
requirements for the plume exposure pathway emergency planning zone
(EPZ) and the ingestion pathway EPZ. The NRC has made a conforming
change to Sec. 50.33(g) in the final rule to address early site permit
applications that propose major features of emergency plans describing
the EPZs under 10 CFR 52.17(b)(2)(i). Such provisions were
inadvertently left out of the proposed rule. For an application for an
early site permit that proposes major features of the emergency plans
describing the EPZs, the change requires the descriptions of the EPZs,
to meet the requirements of Sec. 50.33(g). This is necessary for the
NRC to be able to find that major features describing the EPZs are
acceptable under Sec. 52.18.
Section 50.33(h) formerly required applicants that propose to
construct or alter a production or utilization facility to state in
their application the earliest and latest dates for completion of the
construction or alteration. This section is being revised in the final
rule, based on public comments, to exclude combined license applicants.
The NRC believes that combined license applications need not specify
the earliest and latest date for completion of construction, in light
of the amendment to Section 185 of the AEA that was made by the Energy
Policy Act of 1992. By adding a new Section 185.b. of the AEA, the
Commission believes that Congress intended that Section 185.b supersede
Section 185.a of the AEA, so that the Section 185.a requirements for
``stand-alone'' construction permits, such as the need to specify the
earliest and latest date for completion of construction, do not apply
to the construction permit portion of a combined license under Section
185.b of the AEA. Accordingly, the final rule removes the requirements
from Sec. Sec. 50.33(h), 52.77, and 52.79(a)(39) that the combined
license application specify the earliest and latest date for completion
of construction.
Section 50.33(k) currently requires applicants for operating
licenses to provide a report, as described in Sec. 50.75, indicating
how reasonable assurance that funds will be available for the
decommissioning process is provided. The final rule makes a conforming
change to add a reference to combined licenses. The content of this
report, reflecting the unique considerations of a combined license, is
addressed separately in the revision to Sec. 50.75.
c. Section 50.34, Contents of Construction Permit and Operating License
Applications; Technical Information
The NRC is changing the heading of Sec. 50.34 from Contents of
applications; technical information to read, Contents of construction
permit and operating license applications; technical information.
Section 50.34(a) currently provides the requirements for the technical
contents of an application for a stationary power reactor construction
permit, design certification or combined license, and Sec. 50.34(b)
provides the requirements for the technical contents of an application
for a stationary power reactor operating license application. However,
the former version of 10 CFR part 52 provides requirements for design
certification and combined license applications that are not consistent
with the current version of Sec. 50.34. For example, former Sec.
52.47 stated that an application for design certification must contain
the technical information which is required of applicants for
construction permits and operating licenses by part 50 which is
technically relevant to the design and not site-specific. This would
encompass requirements in both Sec. Sec. 50.34(a) and (b). Also,
former Sec. 52.79 stated that applications for combined licenses must
contain the technically relevant information required of applicants for
an operating license by 10 CFR 50.34, which are found in Sec.
50.34(b). In addition to the requirements for technical information in
Sec. Sec. 50.34(a) and (b), Sec. Sec. 50.34(c) through (h) provide
requirements for the contents of licensing applications related to
security plans, compliance with Three Mile Island (TMI) related
requirements, combustible gas control, and conformance with the
standard review plan. Finally, the NRC notes that the subject of
contents of an application is an administrative matter, rather than a
strictly technical matter. Therefore, these administrative requirements
for part 52 processes are more properly located in part 52, rather than
in Sec. 50.34. To provide maximum clarity in the requirements for the
content of each of the different types of licensing applications, the
NRC is revising Sec. 50.34 to make it applicable to construction
permit and operating license applications only and to provide separate
sections for the technical
[[Page 49398]]
contents of applications for the other types of licenses or regulatory
approvals in 10 CFR part 52 (early site permits in Sec. 52.17, design
certifications in Sec. 52.47, combined licenses in Sec. 52.79, design
approvals in Sec. 52.137, and manufacturing licenses in Sec. 52.157).
In its revisions to 10 CFR part 52, the NRC has brought forward the
requirements from Sec. 50.34 that are applicable to each of the
licensing and approval processes in 10 CFR part 52. One exception to
this structure is the provisions in Sec. 50.34(f) related to
compliance with TMI related requirements. Due to the length and
complexity of the requirements in this paragraph, Sec. 50.34(f) is
being amended to indicate that each applicant for a design
certification, design approval, combined license, or manufacturing
license under part 52 of this chapter must demonstrate compliance with
any technically relevant portions of the requirements in Sec.
50.34(f)(1) through (3), except for paragraphs (f)(1)(xii), (f)(2)(ix),
and (f)(3)(v). The NRC chose this approach rather than repeat the
requirements in each of the relevant sections in part 52. The NRC is
adding the phrase ``except for paragraphs (f)(1)(xii), (f)(2)(ix), and
(f)(3)(v)'' in the last sentence of Sec. 50.34(f) based on public
comments. The commenters pointed out that proposed Sec. 50.34(f) was
inconsistent with proposed Sec. Sec. 52.47(a)(17), 52.79(a)(17),
52.137(a)(17), and 52.157(e)(12), which included the exceptions that
are being added to Sec. 50.34(f) in the final rule.
d. Section 50.34a, Design Objectives for Equipment To Control Releases
of Radioactive Material in Effluents--Nuclear Power Reactors; and Sec.
50.36a, Technical Specifications on Effluents From Nuclear Power
Reactors
Section 50.34a requires that construction permit and operating
license applications include a description of the equipment and
procedures for the control of gaseous and liquid effluents and for the
maintenance and use of equipment installed in radioactive waste
systems. Section 50.34a also requires these applications to include an
estimate of (1) the quantity of each of the principal radionuclides
expected to be released annually to unrestricted areas in liquid
effluents produced during normal reactor operations; and (2) the
quantity of each of the principal radionuclides of the gases, halides,
and particulates expected to be released annually to unrestricted areas
in gaseous effluents produced during normal reactor operations. In
addition, Sec. 50.34a requires a general description of the provisions
for packaging, storage, and shipment offsite of solid waste containing
radioactive materials resulting from treatment of gaseous and liquid
effluents and from other sources. Section 50.34a is revised to clarify
its applicability to the 10 CFR part 52 licensing and approval
processes. Section 50.34a applies to combined licenses by virtue of the
provision in former Sec. 52.83, Applicability of Part 50 Provisions,
which states that all provisions of 10 CFR part 50 and its appendices
applicable to holders of construction permits and operating licenses
also apply to holders of combined licenses. Applicants for design
certification are also required to include the information required by
Sec. 50.34a in their applications by virtue of the provision in former
Sec. 52.47(a)(1)(i), which states that an application for design
certification must contain the technical information which is required
of applicants for construction permits and operating licenses by 10 CFR
part 50 which is technically relevant to the design and not site-
specific. Former appendix O to 10 CFR part 52, Section O.3, explicitly
required applicants for design approvals to include the applicable
technical information required by Sec. 50.34a. Finally, former
appendix M to 10 CFR part 52, Section M.1, states that the provisions
in part 50 applicable to construction permits apply in context, with
respect to matters of radiological health and safety, environmental
protection, and the common defense and security, to manufacturing
licenses. Therefore, new provisions in Sec. 50.34a(d) are adopted to
address the applicable requirements for combined license applications
that parallel the requirements for an operating license application.
New provisions in Sec. 50.34a(e) are adopted to address the applicable
requirements for applications for design approvals, design
certifications, and manufacturing licenses to include: (1) A
description of the equipment for the control of gaseous and liquid
effluents and for the maintenance and use of equipment installed in
radioactive waste systems; and (2) an estimate of the quantity of each
of the principal radionuclides expected to be released annually to
unrestricted areas in liquid effluents produced during normal reactor
operations, and the quantity of each of the principal radionuclides of
the gases, halides, and particulates expected to be released annually
to unrestricted areas in gaseous effluents produced during normal
reactor operations.
e. Section 50.36, Technical Specifications
Section 50.36(a) currently requires that each applicant for a
license authorizing operation of a production or utilization facility
include in its application proposed technical specifications in
accordance with the requirements of Sec. 50.36. The existing language
in Sec. 50.36(a) encompasses combined license applicants. However,
applicants for design certification are also required to include
proposed technical specifications in their applications by virtue of
the provision in former Sec. 52.47(a)(1)(i) stating that an
application for design certification must contain the technical
information required of applicants for construction permits and
operating licenses by 10 CFR part 50 that is technically relevant to
the design and not site-specific. Similarly, applicants for design
approvals are also required to include proposed technical
specifications in their applications by virtue of the provision in
former appendix O to part 52, Section O.3, which states that the
submittal for review of a standard design shall include the applicable
technical information under Sec. 50.34 (a) and (b), as appropriate.
Section 50.36 is revised to clarify that design certification and
manufacturing license applications must also include proposed technical
specifications. The new provisions in Sec. 50.36(c) require each
applicant for a design certification or a manufacturing license to
include proposed generic technical specifications in its application
for the portion of the plant that is within the scope of the design
certification or manufacturing license application.
f. Section 50.36a, Technical Specifications on Effluents From Nuclear
Power Reactors
Section 50.36a(a) requires each licensee of a nuclear power reactor
to include technical specifications to keep releases of radioactive
materials to unrestricted areas during normal conditions, including
expected occurrences, as low as is reasonably achievable. The former
language in Sec. 50.36a(a) encompassed combined license holders.
However, applicants for design certification are also required to
include proposed technical specifications on effluents in their
applications by virtue of the provision in current Sec. 52.47(a)(1)(i)
which states that an application for design certification must contain
the technical information which is required of applicants for
construction permits and operating licenses by 10 CFR part 50
[[Page 49399]]
which is technically relevant to the design and not site-specific. In
addition, former appendix M to 10 CFR part 50, Section M.1, states that
the provisions in part 50 applicable to construction permits apply in
context, with respect to matters of radiological health and safety to
manufacturing licenses. Therefore, Section 50.36a(a) is revised to
state that each licensee of a nuclear power reactor and each applicant
for a design certification or a manufacturing license will include
technical specifications to keep releases of radioactive materials to
unrestricted areas during normal conditions, including expected
occurrences, as low as is reasonably achievable. The proposed rule did
not include the provisions for manufacturing licenses. However,
proposed Sec. 52.157(e)(18) did require manufacturing license
applicants to include proposed technical specifications in accordance
with Sec. 50.36a. Therefore, it was clearly the NRC's intent that the
provisions of Sec. 50.36a be applicable to manufacturing license
applications and the NRC has corrected this omission in the final rule.
Some commenters on the 2006 proposed rule identified an additional
conforming change needed in Sec. 50.36a that the NRC did not make in
the proposed rule. Section 50.36(a)(2) currently requires that each
licensee submit a report to the Commission annually that specifies the
quantity of each of the principal radionuclides released to
unrestricted areas in liquid and in gaseous effluents during the
previous 12 months, including any other information as may be required
by the Commission to estimate maximum potential annual radiation doses
to the public resulting from effluent releases. The NRC has modified
this provision to state that each holder of a combined license is only
required to begin submitting reports after the Commission has made the
finding under Sec. 52.103(g) that allows fuel load and operation. This
would apply the requirements in Sec. 50.36a consistently for part 50
and part 52 licensees, because for a part 50 licensee, the annual
reporting requirement is effective only after an operating license is
issued.
The NRC is also making conforming changes to appendix I to 10 CFR
part 50. These changes parallel the changes to Sec. Sec. 50.34a and
50.36a.
g. Section 50.36b, Environmental Conditions
Section 50.36b authorizes the Commission to include conditions to
protect the environment in each license authorizing operation of a
production or utilization facility and each license for a nuclear power
reactor facility for which the certification of permanent cessation of
operations required under Sec. 50.82(a)(1) has been submitted. These
conditions are to be derived from information contained in the
environmental report and the supplement to the environmental report as
analyzed and evaluated in the NRC record of decision. The conditions
must identify the obligations of the licensee in the environmental
area, including, as appropriate, requirements for reporting and keeping
records of environmental data, and any conditions and monitoring
requirement for the protection of the nonaquatic environment.
The NRC has made conforming changes to Sec. 50.36b in the final
rule to address all applicable part 52 licenses. The changes were made
in response to public comments that highlighted the need for
clarification in Sec. 50.36b. The NRC provided proposed requirements
for identifying environmental conditions on early site permits and
combined licenses in the proposed rule in Sec. Sec. 51.50(b) and (c).
Requirements for identifying environmental conditions for construction
permits were contained in former Sec. 51.50 and proposed Sec.
51.50(a). The proposed rule stated that, in an application for a
construction permit, an early site permit, or a combined license, the
applicant shall identify ``any conditions and monitoring requirements
for protecting the non-aquatic environment, proposed for possible
inclusion in the license as environmental conditions in accordance with
Sec. 50.36b of this chapter.'' However, the NRC neglected to make the
additional conforming changes to Sec. 50.36b in the proposed rule. To
correct this oversight, the NRC has modified Sec. 50.36b in the final
rule to make the requirements in this section consistent with the
requirements in Sec. 51.50. In doing so, the NRC has provided separate
paragraphs for imposing conditions during construction and for imposing
conditions during operation and decommissioning. Paragraph 50.36b(a)
addresses requirements for imposing conditions on construction permits,
early site permits, and combined licenses to protect the environment
during construction. Paragraph 50.36b(b) addresses requirements for
imposing conditions on licenses authorizing operation and licenses for
a facility in decommissioning to protect the environment during
operation and decommissioning. These changes provide consistency in
requirements for environmental conditions across parts 50 and 51.
h. Section 50.37, Agreement Limiting Access to Classified Information
Section 50.37 requires that a license or construction permit
applicant agree in writing that it will not permit any individual to
have access to or any facility to possess Restricted Data or classified
National Security Information until the individual and/or facility has
been approved for access under the provisions of 10 CFR parts 25 and/or
95. Section 50.37 also requires that this agreement be part of the
application for a license or construction permit and that the agreement
of the applicant shall be deemed part of the license or construction
permit, whether stated or not. The former language of Sec. 50.37
encompassed early site permit, combined license, and manufacturing
license applicants under 10 CFR part 52 because these products are all
licenses. However, the NRC is revising Sec. 50.37 to encompass
applicants for design certification and for standard design approvals
under 10 CFR part 52 for consistency with the changes to 10 CFR part
25. Part 25 sets forth the NRC's requirements governing the granting of
access authorization to classified information to certain individuals,
and the Commission is making modifications to part 25 to reflect the
licensing and regulatory approval processes in part 52. Accordingly,
the Commission is revising Sec. 50.37. Section 50.37 is revised to
require that an applicant for a license, construction permit, design
certification, or design approval under part 52 agree in writing that
it will not permit any individual to have access to or any facility to
possess Restricted Data or classified National Security Information
until the individual and/or facility has been approved for access under
the provisions of 10 CFR parts 25 and/or 95. Section 50.37 also
requires that this agreement be part of the application and be deemed
part of the license, or construction permit, or NRC standard design
approval whether stated or not. Section 52.54 is revised to include a
new provision which requires that every standard design certification
rule issued contain a provision that states that, after the Commission
has adopted the final standard design certification rule, the applicant
will not permit any individual to have access to or any facility to
possess Restricted Data or classified National Security Information
until the individual and/or facility has been approved for access under
the provisions of 10 CFR parts 25 and/or 95. The NRC believes that
these revisions, along with the complementary changes to parts 25 and
95, are necessary to
[[Page 49400]]
ensure that access to classified information is adequately controlled
by all entities applying for NRC licenses, design certifications, or
design approvals.
5. Standards for Licenses, Certifications, and Approvals
a. Section 50.40, Common Standards
This section sets forth standards for issuance of a license.
Sections 50.40(a), (b), and (c) are revised to add conforming
references to the additional licensing processes issued under 10 CFR
part 52 that are applicable to these standards.
b. Section 50.43, Additional Standards and Provisions Affecting Class
103 Licenses and Certifications for Commercial Power
The text and heading of this section are revised to clarify that
certain additional standards and provisions for class 103 licenses
apply to applications for combined licenses, design certifications, and
manufacturing licenses issued under part 52, in addition to
applications for construction permits and operating licenses issued
under part 50. Section 50.43(e) is added to clarify that the
requirements to demonstrate new safety features by testing, which were
previously set forth in part 52, apply to applicants for operating
licenses issued under part 50 and applicants for combined licenses,
design certifications, and manufacturing licenses issued under part 52.
This amendment conforms to the goal of having reactor safety
requirements in part 50 and procedural requirements in part 52. Only
the requirements in Sec. 50.43(e) apply to applications for design
certification. Refer to the generic discussion on testing requirements
for advanced reactors in Section V.B of this document.
c. Section 50.45, Standards for Construction Permits, Operating
Licenses, and Combined Licenses
This section is revised to include the standards for review of an
application to alter a facility that was constructed under a combined
license, after the findings under Sec. 52.103(g) of this chapter are
made by the Commission. Some commenters recommended that the proposed
rule be revised to reference the applicable requirements in part 52
rather than the requirements in 10 CFR 50.31 through 50.43 and claimed
that most of those requirements were moved to part 52 in the proposed
rule. The Commission does not agree with that claim but does
acknowledge that most of Sec. 50.34 was moved to the contents of
application section for each of the licensing processes in part 52.
Therefore, Sec. 50.45 was revised to set forth the standards for
review of an application to alter a facility after the Commission makes
the finding under Sec. 52.103(g) of this chapter. The standards for
issuance of a combined license are set forth in Sec. 52.97.
d. Section 50.46, Acceptance Criteria for Emergency Core Cooling
Systems for Light-Water Nuclear Power Reactors
Section 50.46(a)(3) contains reporting requirements for changes to
or errors in emergency core cooling system (ECCS) evaluation models.
Conforming references to design approvals, design certifications, and
licenses issued under part 52 were made to Sec. 50.46, so that the NRC
will be notified of changes to or errors in acceptable evaluation
models, or the application of such models, that were used in licenses,
certifications, and approvals issued under part 52.
e. Section 50.47, Emergency Plans, Sec. 50.54(gg), and Appendix E to
Part 50, Emergency Planning and Preparedness for Production and
Utilization Facilities
Section 50.47 and appendix E to 10 CFR part 50 contain emergency
planning requirements for nuclear power plants. Prior to this
rulemaking, these regulations did not clearly address early site permit
or combined license applicants or holders. Accordingly, the NRC is
making a number of changes in these regulations. Section 50.47(a)(1)
states that no initial operating license for a nuclear power reactor
will be issued unless a finding is made by the NRC that there is
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency, and that no finding
under Sec. 50.47 is necessary for issuance of a renewed nuclear power
reactor operating license. The NRC is revising Sec. 50.47(a)(1) to
include provisions to address combined licenses and early site permits
which include either complete and integrated plans or major features of
the emergency plans. The NRC inadvertently left out provisions to
address early site permits that include major features of the emergency
plans in the proposed rule and a new provision has been added to
address applicants in the final rule.
The NRC is making some additional changes to Sec. 50.47(a)(1) in
the final rule. Proposed Sec. 50.47(a)(1)(ii) stated that ``Except as
provided in paragraph (e) of this section, no initial combined license
under part 52 of this chapter will be issued unless a finding is made
by the NRC that there is reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological
emergency.'' In the final rule, the NRC is removing the phrase ``except
as provided in paragraph (e)'' because paragraph (e) does not address
issuance of the combined license, but, rather, addresses the Commission
finding under Sec. 52.103(g). Likewise, the NRC is making a change to
paragraph (e) of this section in the final rule to remove the reference
to paragraph (a) of this section.
Finally, the NRC is removing the statement in proposed Sec.
50.47(a)(1)(iii) that ``No finding under this section is necessary for
issuance of a renewed early site permit.'' The NRC included this
provision in the proposed rule to be consistent with the existing
requirement for operating licenses. However, upon further
consideration, the NRC concludes that the basis for this exclusion for
an operating license and for a combined license does not apply to an
early site permit. The original license renewal rule, which limited the
scope of matters to be addressed in the renewal proceeding, was based
upon a determination that the regulatory process maintains and updates
the licensing basis for operating licenses, that matters like the state
of the emergency preparedness plans need not be addressed in license
renewal. The bases for the license renewal rule described the process,
in each substantive regulatory area, for maintaining and updating the
current licensing basis. This logic does not directly apply to
emergency preparedness information submitted in an early site permit
application, because there is no maintenance or update requirement for
the early site permit. Therefore, the NRC cannot exclude the need to
address emergency preparedness in an early site permit renewal
proceeding.
Section 50.47(c)(1) provides a process for operating license
applicants that fail to meet the applicable standards of Sec.
50.47(b). The NRC is revising Sec. 50.47(c)(1) to clarify that this
process is applicable to combined license applicants as well.
Section 50.47(d) formerly provided that no NRC or Department of
Homeland Security (DHS) review, findings, or determinations concerning
the state of offsite emergency preparedness or the adequacy of and
capability to implement State and local or utility offsite emergency
plans are required before issuance of an operating license authorizing
only fuel loading or low-power testing and training (up to 5 percent of
the rated power). Section 50.47(d) further stated that a license
authorizing fuel loading and/or low-power testing and training may be
[[Page 49401]]
issued after a finding is made by the NRC that the state of onsite
emergency preparedness provides reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency and provides the standards by which the NRC will
base such a finding. The NRC is adding a new Sec. 50.47(e) to provide
essentially parallel provisions for a combined license holder by
stating that a combined license holder may not load fuel or operate
except as provided in accordance with appendix E to part 50 and,
because of the nature of the combined license process, the NRC is
adding new Sec. 50.54(gg) that would add a condition to all combined
licenses. This is necessary to account for the fact that the combined
license will already be issued at the time of the first full or partial
participation exercise.
The NRC's findings regarding the state of emergency preparedness
for a combined license holder will be taken into account in the NRC's
review under Sec. 52.103(g). The NRC will make its determination by
judging whether the licensee has met the acceptance criteria in the
combined license for the inspections, tests, and analyses related to
the conduct of the first full or partial participation exercise under
paragraph IV.F.2.a of appendix E to part 50. Paragraph 50.54(gg) states
that if, following the conduct of the exercise required by paragraph
IV.F.2.a of appendix E to part 50, DHS identifies one or more
deficiencies in the state of offsite emergency preparedness, the holder
of a combined license may operate at up to 5 percent of rated thermal
power only if the Commission finds that the state of onsite emergency
preparedness provides reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological
emergency. Paragraph 50.54(gg) also provides the standards by which the
NRC will base such a finding.
The NRC is revising appendix E to part 50 to conform to the changes
proposed for Sec. Sec. 50.47 and 50.54. The introduction to appendix E
to part 50 states that each applicant for an operating license is
required by Sec. 50.34(b) to include in the final safety analysis
report plans for coping with emergencies. The NRC is adding a parallel
statement for combined license applicants, and a statement that an
early site permit applicant may submit emergency plans. The final rule
also makes additional conforming changes to the second paragraph of the
introduction that were inadvertently overlooked in the proposed rule.
Similar modifications are proposed in Section III of appendix E to part
50 regarding the content of final safety analysis reports and site
safety analysis reports for an early site permit. The NRC is making a
correction to Section III in the final rule to replace references to
the early site permit application with references to the site safety
analysis report. The NRC is also adding a statement that the site
safety analysis report for an early site permit which proposes major
features must address the relevant provisions of 10 CFR 50.47 and 10
CFR part 50, appendix E, within the scope of emergency preparedness
matters addressed in the major features. This is consistent with the
requirements in Sec. 52.17(b).
In Section IV of appendix E to part 50, the NRC is modifying
paragraph F.2.a, to address combined licenses in addition to operating
licenses. Paragraph F.2.a currently provides requirements regarding the
conduct of full participation exercises and states that a full
participation exercise shall be conducted within 2 years before the
issuance of the first operating license for full power of the first
reactor. Paragraph F.2.a also requires that, if the full participation
exercise is conducted more than 1 year before issuance of an operating
licensee for full power, an exercise which tests the licensee's onsite
emergency plans shall be conducted within 1 year before issuance of an
operating license for full power. The NRC is designating the
requirements for operating licenses as paragraph F.2.a.i, and adding a
new paragraph F.2.a.ii that contains the requirements for combined
licenses. Paragraph F.2.a.ii states that, for a combined license, the
first full participation exercise must be conducted within 2 years of
the scheduled date for initial loading of fuel and operation under
Sec. 52.103. Paragraph F.2.a.ii also requires that, if the first full
participation exercise is conducted more than 1 year before the
scheduled date for initial loading of fuel and operation under Sec.
52.103, an exercise which tests the licensee's onsite emergency plans
must be conducted within 1 year before the scheduled date for initial
loading of fuel and operation under Sec. 52.103. The modifications
further state that, if DHS identifies one or more deficiencies in the
state of offsite emergency preparedness as the result of the first full
participation exercise, or if the NRC finds that the state of emergency
preparedness does not provide reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency, the provisions of Sec. 50.54(gg) will apply,
as previously discussed.
The NRC is adding a new paragraph IV.F.2.a.iii to appendix E to
part 50 to require that, if the applicant has an operating reactor at
the site, an exercise, either full or partial participation, be
conducted for each subsequent reactor constructed on the site. This
exercise may be incorporated in the exercise requirements of paragraphs
(2)(b) and (2)(c) of Section IV.F. If DHS identifies one or more
deficiencies in the state of offsite emergency preparedness as the
result of this exercise for the new reactor, or if the NRC finds that
the state of emergency preparedness does not provide reasonable
assurance that adequate protective measures can and will be taken in
the event of a radiological emergency, the provisions of Sec.
50.54(gg) apply just as they do for the first reactor at a site. This
new provision is desirable because of the nature of ITAAC for emergency
preparedness requirements. The emergency preparedness ITAAC,
specifically ITAAC that will be demonstrated through an exercise,
provide the necessary reasonable assurance for programs and facilities
associated with the yet-unbuilt reactor. Recent agreements between the
NRC and external stakeholders on emergency preparedness ITAAC are based
on the understanding that ITAAC on the emergency preparedness exercise
would serve to demonstrate various aspects of emergency preparedness
(e.g., programs and facilities) that did not warrant their own
specific/detailed ITAAC. For example, there is no ITAAC for determining
whether an adequate staffing roster exists for the technical support
center or emergency offsite facility, but its existence and adequacy
could be demonstrated during an exercise. Therefore, appendix E to part
50 requirements for emergency preparedness exercises must be included
for the current concepts regarding emergency preparedness ITAAC to be
viable. With regard to subsequent reactors, those aspects of an
exercise which address currently untested (i.e., unexercised) aspects
of emergency preparedness for the proposed new reactor must be
addressed in new emergency preparedness ITAAC for the subsequent
reactor. If various generic exercise-related aspects of emergency
preparedness for the site have been previously addressed and satisfied,
then there would be no ITAAC for those emergency preparedness aspects
for subsequent reactors.
The NRC is also modifying Section V of appendix E to part 50, which
states
[[Page 49402]]
that no less than 180 days before the scheduled issuance of an
operating license for a nuclear power reactor or a license to possess
nuclear material, the applicant's detailed implementing procedures for
its emergency plan shall be submitted to the Commission. Paragraph V
also requires that licensees submit any changes to the emergency plan
or procedures to the NRC within 30 days of these changes. The NRC is
clarifying that paragraph V is also applicable to COL holders by
stating that they must submit their detailed implementing procedures
for their emergency plans to the NRC no less than 180 days before the
scheduled date for initial loading of fuel. The wording of this
requirement has been changed slightly in the final rule. In the
proposed rule, this provision required that COL holders submit their
detailed implementing procedures for their emergency plans to the NRC
no less than 180 days before the date that the Commission authorizes
fuel load and operation under Sec. 52.103. The NRC has modified the
provision to make the target date 180 days before scheduled date for
initial loading of fuel because this will be a known date whereas the
licensee would not know the date that the Commission will make the
Sec. 52.103(g) finding. This change is also consistent with other
requirements in appendix E that are tied to the scheduled date for
initial fuel load.
f. Section 50.48, Fire Protection
Section 50.48(a)(1) is revised to clarify that holders of an
operating license issued under part 50 and a combined license issued
under part 52 must have a fire protection plan. Section 50.48(a)(4) is
added to clarify that applications for design approvals, design
certifications, and manufacturing licenses issued under part 52 must
meet the fire protection design requirements set forth in general
design criterion 3 of appendix A to part 50.
g. Section 50.49, Environmental Qualification of Electric Equipment
Important to Safety for Nuclear Power Plants
Section 50.49(a) is revised to clarify that these programmatic
requirements apply to applicants for and holders of operating licenses
issued under part 50 and combined licenses and manufacturing licenses
under part 52.
h. Section 50.54, Conditions of Licenses; and Sec. 50.55, Conditions
of Construction Permits, Early Site Permits, Combined Licenses, and
Manufacturing Licenses
Section 50.54 sets forth various provisions that are deemed to be
conditions ``in every license issued,'' while Sec. 50.55 sets forth
the provisions deemed to be conditions of every construction permit. In
making the conforming changes to these regulations to reflect part 52,
the NRC has decided to maintain this dichotomy. Conditions applicable
to part 52 processes which are either licenses or prerequisites to
licenses, and do not address activities analogous to construction for
which a construction permit license is required under the AEA, are
addressed in Sec. 50.54. By contrast, conditions applicable to part 52
processes which address construction activities, or activities
analogous to construction for which a construction permit license is
required under the AEA, are covered in Sec. 50.55. Combined licenses
represent a special case, inasmuch as they address both construction
and operation. The NRC addresses combined licenses by placing the
conditions applicable only to construction in Sec. 50.55, which
indicates that these conditions are applicable until the date that the
Commission makes the finding under Sec. 52.103(g). Conditions which
are applicable during construction and operation or only during
operation are set forth in Sec. 50.54. The NRC is revising the
introductory paragraph of Sec. 50.54 to refer to combined licenses,
and to exclude manufacturing licenses from its provisions. The NRC is
making revisions to Sec. 50.54 in the final rule based on public
comments. In the proposed rule, the NRC did not distinguish which
provisions in Sec. 50.54 are applicable only during operation from
those that are applicable during both construction and operation. In
the final rule, the NRC has revised the introductory paragraph to
indicate which provisions are applicable only after the Commission
makes the finding under Sec. 52.103(g). In making these revisions, the
NRC determined that the provisions that need to be applied during both
construction and operation are paragraphs (a) through (h), (o), (p),
(q), (t), (v), and (aa) through (ee). All of these provisions have some
requirements that will be implemented prior to the Commission finding
under Sec. 52.103(g).
In addition, the NRC is adding paragraphs (r) and (u) to the list
of provisions in the introduction that are not applicable to combined
licenses. This is because paragraph (r) only applies to research and
test reactor facilities and paragraph (u) was only applicable for 60
days after the amendment to Sec. 50.54 that added paragraph (u).
Finally, the NRC is also revising the first sentence of the
introduction to indicate that paragraphs (r) and (gg) do not apply to
nuclear power reactor operating licenses. In the proposed rule, the
introduction stated that they did not apply to operating licenses,
which would have included research and test reactor operating licenses.
The NRC is revising Sec. 50.54(a)(1) to indicate that the quality
assurance (QA) requirements applicable to operation, as described in a
combined license holder's SAR, become effective 30 days before the
scheduled date for the initial loading of fuel.
The NRC is revising Sec. 50.54(i-1) to indicate its applicability
to combined licenses. Specifically, Sec. 50.54(i-1) requires that
within 3 months after the date that the Commission makes the finding
under Sec. 52.103(g) for a combined license, the licensee shall have
in effect an operator requalification program that must, as a minimum,
meet the requirements of Sec. 55.59(c) of this chapter.
The NRC has added changes to Sec. 50.54(p) and (q) in the final
rule. The changes to paragraph (p) are being made to include references
to appropriate part 52 sections in addition to the existing references
to part 50 sections. The change to paragraph (q) is being added to
include a statement that, for combined licenses, the requirement to
follow and maintain in effect emergency plans which meet the standards
in Sec. 50.47(b) and the requirements in appendix E of part 50 is only
applicable after the Commission makes the finding under Sec.
52.103(g). However, the remainder of the requirements in paragraph (p)
apply from the time the combined license is issued (e.g., requirements
to retain records of emergency plan changes). This is consistent with
the change made to the introductory paragraph of Sec. 50.54 discussed
earlier in this section.
The NRC is adding a new Sec. 50.54(gg). These revisions are
discussed with related requirements in Section IV.D.4.f of this
document, ``Section 50.47, Emergency plans, Sec. 50.54(gg), and
appendix E to part 50.''
Although the NRC generally views Sec. 50.55 as the appropriate
section in part 50 for specifying the conditions applicable to
construction permits and part 52 processes analogous to construction
permits, the NRC does not believe that all of the conditions in Sec.
50.55 should apply equally to all of the part 52 processes.
Accordingly, the introductory text to Sec. 50.55 is revised to specify
which paragraphs apply to a construction permit, early site permit,
combined license, and manufacturing license.
[[Page 49403]]
Sections 50.55(a) and (b) of the March 2006 proposed rule would
have required a combined license to state the earliest and latest dates
for completion of construction or modification, and to provide for
forfeiture of the combined license if the construction or modification
is not completed by the stated date. The Commission has reconsidered
this position and has decided to remove this requirement from the final
rule. The statutory requirement for a construction permit to state the
earliest and latest date for completion of construction is now
contained in Section 185.a of the AEA. The combined license, by
contrast, is address in Section 185.b. The Commission believes that in
the absence of specific language regarding the restriction in paragraph
a. applicable to combined licenses in paragraph b., the combined
license is not subject to any of the statutory restrictions in
paragraph a. The NRC believes that the provisions of Section 185 of the
AEA do not apply to a manufacturing license, inasmuch as a
manufacturing license is not, per se, a construction permit.
Accordingly, no earliest and latest date for completion of manufacture
would be required to be stated in a manufacturing license.
Section 50.55(c) makes the license conditions in Sec. 50.54 also
apply to construction permits, unless otherwise modified. In the
proposed rule, the NRC revised this paragraph to add a reference to
combined licenses. However, upon further consideration, the NRC has
determined that no change to Sec. 50.55(c) is necessary because the
introduction to Sec. 50.54 outlines which provision in that section
apply to combined licenses.
Section 50.55(e) addresses the obligation of holders of
construction permits and their contractors and subcontractors, to
report defects constituting a substantial safety hazard. These
requirements, which implement Section 206 of the ERA, as amended, are
comparable to the requirements in 10 CFR part 21. As discussed with
respect to the NRC's changes to part 21, the NRC is retaining the
current regulatory structure, whereby persons and entities engaged in
activities constituting construction (and their contractors and
subcontractors) are subject to Sec. 50.55(e), and persons and
licensees who are authorized to operate a nuclear power plant (and
their contractors and subcontractors) are subject to part 21. Inasmuch
as a combined license under part 52 authorizes both construction and
operation, a combined license holder would be subject to the reporting
requirements in Sec. 50.55(e) from the date of issuance of the
combined license until the Commission makes the finding under Sec.
52.103. Thereafter, the combined license holder would be governed by
the reporting requirements in part 21. The manufacture of a nuclear
power reactor under a manufacturing license is the functional
equivalent of construction. Accordingly, the NRC's view is that the
holder of a manufacturing license should be subject to reporting under
Sec. 50.55(e). Standard design approvals under subpart E to part 50
(former appendix M to part 52) and design certifications under subpart
B of part 52 are not directly associated with construction, and the NRC
believes that their reporting should be addressed under part 21.
Accordingly, the NRC is revising Sec. 50.55(e)(1) to provide that the
reporting requirements in Sec. 50.55(e) apply to a holder for a
combined license (until the NRC makes the finding under Sec.
52.103(g)), and a manufacturing license under part 52. As discussed
further in Section J on part 21 of this document, early site permits do
not authorize ``construction'' or its functional equivalent. Therefore,
early site permits are subject to the requirements of part 21 rather
than Sec. 50.55(e) under the final rule.
Section 50.55(f) sets forth the NRC's requirements with respect to
compliance with the QA requirements in 10 CFR part 50, appendix B, and
implementation of the construction permit holder's QA program as
described in its SAR. Comparable provisions applicable to holders of
operating licenses are contained in Sec. 50.54(a); requirements
governing the SAR's description of the QA program are contained in
Sec. 50.34. A detailed discussion of all changes related to QA
requirements can be found in Section IV.D.13.b of this document.
i. Section 50.55a, Codes and Standards
Section 50.55a provides requirements relating to codes and
standards for construction permits and operating licenses for boiling
or pressurized water-cooled nuclear power facilities. The NRC is
revising Sec. 50.55a to clarify how the regulations in Sec. 50.55a
apply to approvals, certifications, and licenses issued under 10 CFR
part 52. Section 50.55a formerly applied to combined licenses by virtue
of the provision in current Sec. 52.83, which stated that all
provisions of 10 CFR part 50 and its appendices applicable to holders
of construction permits and operating licenses also apply to holders of
combined licenses. Also, Sec. 50.55a formerly applied to design
certifications by virtue of the provision in former Sec. 52.48, which
states that design certification applications will be reviewed for
compliance with the standards set out in 10 CFR part 50 as it applies
to applications for construction permits and operating licenses for
nuclear power plants, and as those standards are technically relevant
to the design proposed for the facility. Although former appendix O to
part 52 does not explicitly require applicants for design approvals to
comply with the requirements of Sec. 50.55a, the NRC is requiring
design approval holders to comply with Sec. 50.55a because the NRC
believes that the requirements for a design approval should be the same
as the requirements for design certification, given that the reviews
performed by the NRC staff for the two products are essentially
identical. Finally, appendix M to part 52, Section M.1, states that the
provisions in part 50 applicable to construction permits apply in
context, with respect to matters of radiological health and safety,
environmental protection, and the common defense and security, to
manufacturing licenses. Therefore, the NRC is modifying Sec. 50.55a to
state that each combined license for a utilization facility is subject
to the conditions in Sec. 50.55a, but is only subject to the
conditions in Sec. Sec. 50.55a(f) and (g) after the NRC makes the
finding under Sec. 52.103. The modifications to Sec. 50.55a also
state that each manufacturing license, design approval, and design
certification application is subject to the conditions in Sec. Sec.
50.55a(a), (b)(1), (b)(4), (c), (d), (e), (f)(3), and (g)(3), which are
the provisions related to nuclear power facility design.
j. Section 50.59, Changes, Tests, and Experiments
This section presents a change process for information contained in
the FSAR. Section 50.59(b) is revised to clarify that this change
process is applicable to holders of operating licenses issued under
part 50 and combined licenses issued under part 52. If the combined
license references a design certification rule, then the information in
the design control document is controlled by the change process in the
applicable design certification rule. Section 50.59(d)(2) is revised to
conform the frequency that summary reports are submitted for holders of
combined licenses with the frequency set forth in the design
certification rules. Section 50.59(d)(3) is revised to clarify that the
requirement for maintaining records applies to holders of operating
licenses issued under part 50 and combined licenses issued under part
52.
[[Page 49404]]
k. Section 50.61, Fracture Toughness Requirements for Protection
Against Pressurized Thermal Shock Events
This section is revised to clarify that the fracture toughness
requirements apply to an operating license for a pressurized water
reactor issued under part 50 or a combined license for a pressurized
water reactor issued under 10 CFR part 52.
l. Section 50.62, Requirements for Reduction of Risk From Anticipated
Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear
Power Plants
Paragraph (d) of Sec. 50.62 provides implementation requirements
for the requirements of the section. This paragraph is revised to
indicate that these implementation requirements only apply to light-
water-cooled nuclear power plant operating licenses issued before the
effective date of this final rule. Section 50.62 is revised to require
each light-water-cooled nuclear power plant operating license
application submitted after the effective date of this final rule to
submit information in its final safety analysis report demonstrating
how it will comply with paragraphs (c)(1) through (c)(5) of Sec.
50.62. Similarly, the NRC is adding provisions to Sec. Sec. 52.47,
52.79, 52.137, and 52.157 requiring that applicants for standard design
certifications, combined licenses, standard design approvals, and
manufacturing licenses include the information required by this section
in their final safety analysis reports.
m. Section 50.63, Loss of All Alternating Current Power
Conforming changes are made to this section to clarify that the
requirements for station blackout apply to applications for
construction permits, combined licenses, design approvals, design
certifications, manufacturing licenses, and operating licenses.
n. Section 50.65, Requirements for Monitoring the Effectiveness of
Maintenance at Nuclear Power Plants
This section presents the requirements for monitoring the
effectiveness of maintenance at nuclear power plants. Paragraph
50.65(a) is revised to clarify that holders of operating licenses
issued under part 50 and combined licenses issued under part 52 must
comply with the requirements in this section. In the proposed rule,
Sec. 50.65(c) was revised to specify that, for new licenses issued
after the effective date of this regulation, the requirements of this
section must be implemented 30 days before the initial fuel loading of
the reactor. Commenters recommended that NRC should not require
implementation prior to fuel load when not all systems will have been
placed in service. The NRC agrees with this comment and has deleted the
proposed revision to Sec. 50.65(c). Under the final rule, licensees
are required to implement the requirements of this section by the time
that initial fuel loading has been authorized.
6. Inspections, Records, Reports, Notifications
a. Section 50.70, Inspections
Section 50.70(a) requires that each licensee and each holder of a
construction permit allow inspection, by duly authorized
representatives of the Commission, of its records, premises,
activities, and of licensed materials in possession or use, related to
the license or construction permit as may be necessary to effectuate
the purposes of the AEA. The language in Sec. 50.70(a) encompasses
combined license holders and manufacturing license holders because they
are licensees. In addition, the provision in former Sec. 52.83, states
that all provisions of 10 CFR part 50 and its appendices applicable to
holders of construction permits and operating licenses also apply to
holders of combined licenses. Also, former Section M.1 of appendix M to
part 52, states that the provisions in part 50 applicable to
construction permits apply in context, with respect to matters of
radiological health and safety, environmental protection, and the
common defense and security, to manufacturing licenses. Section
50.70(a) is revised to clarify that these inspection requirements also
apply to holders of early site permits under 10 CFR part 52. An early
site permit is a partial construction permit and therefore should be
subject to the same inspection requirements as a construction permit.
In addition, the NRC is clarifying that the inspection requirements
also apply to applicants for licenses, construction permits, and early
site permits. It is common for applicants to perform activities related
to NRC regulations before issuance of the license or permit for which
they are applying and it has been the NRC's practice to inspect these
activities whenever they are performed. Therefore, the modification to
require that the inspection requirements in Sec. 50.70(a) apply to
applicants is simply a codification of the NRC's current practices.
Section 50.70(b)(1) requires that each licensee and each holder of
a construction permit provide rent-free office space for the exclusive
use of NRC inspection personnel. The existing language in this
provision encompasses combined license holders and manufacturing
license holders. Section 50.70(b)(2) provides requirements regarding
the space to be provided for a site with a single power reactor
facility licensed under 10 CFR part 50 and for sites containing
multiple power reactor units. The NRC is revising Sec. 50.70(b)(2) to
clarify that these requirements also apply to sites for combined
license holders under 10 CFR part 52 and to facilities issued
manufacturing licenses under 10 CFR part 52.
b. Section 50.71, Maintenance of Records, Making of Reports
Section 50.71 establishes the NRC's requirements for maintenance
and retention of records and reports, and updating of FSARs. Section
50.71(a) requires each licensee and each holder of a construction
permit to maintain all records and make all reports as may be required
by license, or by the NRC's regulations. The former language does not
apply to non-licensees, such as holders of standard design approvals
and applicants for standard design certifications, even though it would
appear that these requirements should Accordingly, the NRC is revising
Sec. 50.71(a) to make its provisions applicable to holders of standard
design approvals and all applicants for design certification during the
period of NRC consideration of the application for design
certification, and those applicants for design certification whose
designs are certified via rulemaking in accordance with subpart B of 10
CFR part 52.
Section 50.71(c) specifies that the default record retention period
(i.e., the period that applies if a record retention period is not
specified by the regulation requiring the record) ends when the NRC
``terminates the facility license.'' A manufacturing license is not a
``facility'' license, inasmuch as subpart F of part 52 is limited to
the manufacture of reactors, not a ``facility.'' Finally, some licenses
(e.g., early site permits and manufacturing licenses) may either be
terminated by the NRC, or ``expire'' as a matter of law at the end of
their term. Accordingly, the NRC is revising Sec. 50.71(c) to
establish the records retention period and to properly refer to
manufacturing licenses, early site permits, and construction permits.
Section 50.71(e) establishes the updating requirements for the
FSAR, including the information that must be included in each update.
The former regulation, however was deficient in two respects. First, it
did not address the updating requirements for combined license
applicants and holders. Second,
[[Page 49405]]
the regulation, if applied to manufacturing licenses under subpart F of
part 52, imposed unnecessary regulatory burden with respect to periodic
updating.
Accordingly, the NRC is revising Sec. 50.71(e) to specify the FSAR
updating requirements for combined license applicants and holders. In
addition, current Sec. 50.71(f) is redesignated as Sec. 50.71(g), and
a new Sec. 50.71(f) is added.
Section 50.71(e)(3)(iii) is added to contain the provisions
applicable to combined license holders during the period of time from
docketing of the application to the Commission finding under Sec.
52.103(g). The update frequency during this period is established as
annually, which is consistent with requirements in Section X.B.3.b of
the design certification rules in appendices A through D of part 52 for
combined license holders that reference those rules. After the
Commission finding under Sec. 52.103(g), the frequency would be
governed by Sec. 50.71(e)(4), as for other operating reactors.
Section 50.71(f) is revised to require the holder of the
manufacturing license to update the FSAR to reflect any modifications
to the design of the reactor authorized to be manufactured which have
been approved by the NRC under Sec. 52.171, or any new analyses
requested to be performed by the NRC. Periodic updating of an FSAR for
a manufacturing license is not required by Sec. 50.71(f), inasmuch as
the NRC's concept for a manufacturing license is for the design of the
reactor authorized to be manufactured to be stable with no changes
except as specifically approved by the NRC as necessary for adequate
protection to public health and safety or common defense and security,
or to ensure compliance with the NRC's requirements in effect at the
time of issuance of the manufacturing license. The provision in Sec.
50.71(f) requiring the FSAR for a manufacturing license to be updated
to reflect new safety analyses required by the NRC is analogous to the
existing updating requirement in Sec. 50.71(e). This assures that new
analyses performed to demonstrate the continuing adequacy of the
unchanged manufactured reactor design are appropriately reflected in
the FSAR.
Paragraph (g), formerly (f), is being revised to add reference to
Sec. 52.110(a)(1) for permanent cessation of operation for plants
licensed under part 52.
Finally, paragraph (h) is being added to 50.71. This paragraph
contains requirements for licensees to maintain and upgrade the PRA
periodically throughout the plant life. These provisions apply only to
COLs under part 52, but are included in part 50 in this section
covering maintenance of records and making of reports, consistent with
the Commission's practice elsewhere in development of the requirements
for the part 52 processes.
These new requirements are a culmination of the Commission's
interest in use of risk-informed processes as articulated in its 1995
Policy Statement (``Use of Probabilistic Risk Assessment Methods in
Nuclear Activities: Final Policy Statement,'' (60 FR 42622; August 16,
1995)).In the original part 52 rule, each design certification holder
was required to include as part of the application a design-specific
PRA. The Commission has been engaged in an effort to improve PRA
quality through support and endorsement of consensus standards on PRA
methods.
In the proposed rule published in March 2006, the Commission
included a specific request for comment (Question 10, ``New
Requirements for Periodic Updates to the PRA''--see section IV of this
document) about part 52 licensees periodically updating the PRA
throughout the life of the facility, on a schedule similar to that for
FSAR updates. Several commenters noted that the proposed rule did not
include a frequency for updating the PRA. These commenters stated that
they believed that PRA update frequency should be addressed in guidance
rather than regulations. These commenters indicated a frequency of once
every two operating cycles would be reasonable and consistent with
existing requirements in 10 CFR 50.69(e). After considering the
comments received, the Commission has decided to require combined
license holders to maintain and upgrade a PRA to meets endorsed
standards over the lifetime of the facility. To implement this
decision, new requirements are being placed in Sec. 50.71(h).
Paragraph (h)(1) requires each holder of a combined license, by the
time of the scheduled fuel load date for the facility, to develop a
plant-specific PRA. The PRA is to be both level 1 and level 2 and must
cover those modes of operation and initiating events for which NRC-
endorsed consensus standards are in effect one year prior to that date.
Level 1 refers to the identification and quantification of sequences
leading to the onset of core damage. Level 2 refers to identification
and quantification of severe accident progression and containment
response. Additional information about scope and quality of PRA to meet
these provisions will be addressed in the NRC documents endorsing the
standards, or in the standards themselves.
The one year time period was chosen to allow time for the licensee
to develop and upgrade its PRA and conduct peer review prior to the
date when the PRA must be completed (i.e., by the scheduled date for
initial fuel load). The scheduled fuel load date was selected because
the COL holder chooses this date, and thus is in a position to
determine when the ``one-year prior'' requirement comes into effect.
Note that this provision does not require that this PRA be submitted to
the NRC for review and approval. The need for any such submittal or
review would be determined by any risk-informed application for which
the licensee might wish to use this PRA, such as in support of
licensing actions.
Paragraph (h)(2) requires the COL holder to maintain the PRA until
permanent cessation of operations under Sec. 52.110(a). The Commission
intends PRA maintenance to be consistent with how it is defined in the
American Society of Mechanical Engineers (ASME) ``Standard for
Probabilistic Risk Assessment for Nuclear Power Plant Applications''
(ASME-RA-Sb-2005), that is ``the update of the PRA models to reflect
plant changes, such as modifications, procedure changes or plant
performance.'' No specific frequency is defined in the rule for such
maintenance; the Commission expects licensees to follow the ASME (or
other consensus body) guidance on this aspect.
The paragraph further provides that the PRA must be upgraded every
four years, to cover initiating events and operational modes contained
in NRC-endorsed consensus standards in effect one year prior to each
required upgrade. The Commission intends PRA upgrade to be consistent
with how it is defined in consensus standards, such as ASME-RA-Sb-2005,
that is, ``the incorporation into a PRA model of a new methodology or
significant changes in scope or capability.'' If no new standards are
issued during a four-year upgrade cycle, licensees would not be
required to upgrade their PRAs; however, the requirement to maintain
the PRA would still be in effect. It should also be noted that there
may be situations where a PRA upgrade is needed more frequently than
the four year cycle, as for instance to support a new risk-informed
application.
Finally, paragraph (h)(3) specifies that each holder of a combined
license shall, no later than the date on which the licensee submits an
application for a renewed license, upgrade the PRA to
[[Page 49406]]
cover all modes and all initiating events. This requirement is not
premised on the existence of NRC-approved consensus standards, and an
all-mode, all-initiator PRA must be developed even if standards do not
yet exist. The requirement to develop and maintain such a PRA by the
time of license renewal application is intended only to establish a
timing requirement for completing the upgrade of the PRA, and does not
have any implications on the current requirements for license renewal.
The upgraded PRA is not an element of any (i.e., past, present, or
future) review or approval of a license renewal application.
In implementing these new requirements, it is the NRC's expectation
that industry stakeholders will work with the NRC and appropriate codes
and standard setting bodies to continually upgrade the relevant codes
and standards, identify potential issues, resolve problems, and create
relevant guidance to assist in periodically improving the quality and
comprehensiveness of the PRA.
c. Section 50.72, Immediate Notification Requirements for Operating
Nuclear Power Reactors
Section 50.72 currently requires holders of operating licenses
under part 50 for nuclear power plants to notify the NRC Operations
Center via the Emergency Notification System of the declaration of any
of the emergency classes specified in the licensee's approved emergency
plan and of certain non-emergency events. The NRC's regulatory interest
in these events also extends to nuclear power plants operating under a
combined license under subpart C of part 52, but the former language
did not impose the notification requirements on combined license
holders. Accordingly, in a conforming change in the final rule, the NRC
is extending the notification requirements to holders of combined
licenses under part 52 after the Commission has made the finding under
Sec. 52.103(g). The NRC did not include a conforming change to this
section in the proposed rule. However, based on public comments, the
NRC is including the change in the final rule to make it clear that the
requirements of Sec. 50.72 only apply to a combined license holder
after the Commission makes the finding under Sec. 52.103(g). The NRC
is not extending the notification requirements to other part 52
processes because the events to be reported under the existing rule
concern events which can only occur upon fuel load and operation, and
the remaining part 52 licensing and regulatory approval processes do
not authorize fuel load or operation.
d. Section 50.73, Licensee Event Report System
Section 50.73 requires holders of operating licenses under part 50
for nuclear power plants to submit licensee event reports (LERs) on the
occurrence of certain operating events to the NRC. LERs facilitate the
NRC's oversight of operating nuclear power plants, by alerting the NRC
to the occurrence and underlying causes of events having potential
safety implications. The NRC's regulatory interest in these events also
extends to nuclear power plants operating under a combined license
under subpart C of part 52, but the former language did not impose the
LER requirement on combined license holders. Accordingly, in a
conforming change, the NRC is extending the LER reporting requirements
to holders of combined licenses under part 52 after the Commission has
made the finding under Sec. 52.103(g). The final rule does not extend
the LER requirement to other part 52 processes, because the events to
be reported under the existing rule concern events which can only occur
upon fuel load and operation, and the remaining part 52 licensing and
regulatory approval processes do not authorize fuel load or operation.
e. Section 50.75, Reporting and Recordkeeping for Decommissioning
Planning
The requirements in Sec. 50.75 are intended to ensure that
entities who construct and ultimately operate a nuclear power plant
will have sufficient funds at the end of the operational life of the
plant to complete the decommissioning of the plant. Section 50.75
requires a nuclear power plant operating license application to address
the predicted costs of decommissioning, provide financial assurance by
one of the means specified in the regulation, and submit evidence that
one or more of these means has been established. Section 50.75 also
requires the operating license holder to update the cost estimates for
decommissioning on an annual basis, and to submit reports to the NRC
every 2 years describing, inter alia, any adjustments to the amount of
funds collected annually to reflect any changes in projected
decommissioning cost. When a plant is within 5 years of its projected
end of its operation, the reports must be submitted annually, and a
site-specific decommissioning cost estimate must be submitted. Some of
these requirements are directed at the two phase licensing process in
10 CFR part 50, in which the NRC issues a construction permit followed
by an operating license. These requirements are not well-suited to the
combined license process under part 52. For example, requiring the
combined license applicant to comply with the current requirement in
Sec. 50.75(b)(4) that the operating license applicant submit a copy of
the financial instrument obtained to satisfy the requirements of Sec.
50.75(e), would place a more stringent requirement on the combined
license applicant, inasmuch as that applicant would be required to fund
decommissioning assurance at an earlier date as compared with the
operating license applicant.
To address these discrepancies, the NRC is revising Sec. 50.75 to
address decommissioning funding assurance for combined licenses. Under
the final rule, the combined license applicant must submit a
decommissioning report as required by Sec. 50.33(k), but it need not
obtain a financial instrument to fund decommissioning or to submit a
copy to the NRC. Instead, under Sec. 50.75(b)(1) and (4), the combined
license application must contain a certification that the financial
assurance will be provided no later than 30 days after the NRC
publishes notice in the Federal Register under Sec. 52.103(a). See
Sec. 50.75(b)(1).
The proposed rule would have required the combined license holder
to submit, by March 31 of each year until the date that the NRC
authorizes fuel load under Sec. 52.103(g), an updated certification of
the information required by paragraph (b)(1). The proposed rule also
would have required the combined license holder to submit, no later
than 30 days after the Commission publishes notice in the Federal
Register under Sec. 52.103(a), a certification that financial
assurance is being provided in the relevant amount together with a copy
of the financial instrument obtained to satisfy the requirements of
Sec. 50.75(e). Once the Commission has made the finding under Sec.
52.103, the proposed rule would have required the combined license
holder to be subject to the reporting and updating requirements as an
operating license holder under part 50, including the requirements
applicable when the plant is within 5 years of the projected end of
operation. A commenter objected to the annual reporting requirement,
arguing that an annual update during the construction period would
serve no purpose and is unnecessary and unduly burdensome. The
commenter proposed that the holder be allowed to adjust or update the
original certification at the time construction is complete and the
plant is ready to begin operation. Upon
[[Page 49407]]
further consideration, the Commission has decided to modify the final
rule by eliminating the requirement for annual reports, and instead
requiring the updating reports 2 years and 1 year before the date
scheduled for initial loading of fuel load (consistent with the
schedule required by Sec. 52.99(a)). The Commission's objective is to
have sufficient time to evaluate the projected costs of
decommissioning, and any licensee-proposed changes in the financial
assurance mechanism for funding before fuel is loaded into the reactor
and operation commences. This will allow the Commission to take any
necessary regulatory action before fuel loading and commencement of
operation.
The final rule requires that no later than 30 days after the
Commission publishes notice in the Federal Register under Sec.
52.103(a), the combined license holder must submit a report to the NRC.
The report must contain a certification that financial assurance is
being provided in an amount specified in the licensee's most recent
updated certification (i.e., the certification provided 1 year before
the scheduled date for initial loading of fuel, in accordance with the
first sentence of Sec. 50.75(e)(3)). The certification must include a
copy of the financial instrument obtained to provide decommissioning
funding assurance. The requirements in paragraph (f)(1) of Sec.
52.103(a), which are applicable to the combined license holder after
the Commission has made the finding under Sec. 52.103, are adopted in
the final rule without change from the proposed rule.
The Sec. 50.75 decommissioning funding requirements do not apply
to an applicant for, and holder of, a manufacturing license under part
52. The NRC did not intend, when it first adopted Sec. 50.75, to
subject holders of manufacturing licenses to the requirements of that
section. It is clear from the words of former Sec. 50.33(k)(1) that
the rule applies only to applications for operating licenses for
production and utilization facilities. A manufacturing license by
itself does not authorize either fuel load or operation, which are the
activities necessitating the expenditure of funds for decommissioning.
Therefore, there is no need for a holder of a manufacturing license,
who does not intend to operate the reactor being manufactured to
provide funding.
7. US/IAEA Safeguards Agreement
a. Section 50.78, Installation Information and Verification
Since 1980, the U.S./International Atomic Energy Agency (IAEA)
Safeguards Agreement has allowed IAEA inspection and verification
activities at U.S. facilities that the IAEA selects from the U.S.
Eligible Facilities List. The safeguards agreement is implemented under
the Nuclear Non-Proliferation Treaty, which provides assurance that all
nuclear materials declared to be in peaceful use are not diverted to
potential use in nuclear explosives. Although 10 CFR part 75 contains
most of the NRC requirements intended to implement the installation,
inspection, and verification provisions of the Safeguards Agreement
with IAEA, Sec. 50.78 requires each holder of a construction permit to
submit certain information on Form N-71, permit verification by
representatives of the IAEA, and take any other action necessary to
implement the Safeguards Agreement. Inasmuch as combined licenses
authorize construction of a nuclear power plant at a fixed site, the
provisions of Sec. 50.78 should also apply to a holder of a combined
license under part 52. Accordingly, Sec. 50.78 is revised to specify
that holders of combined licenses must, if requested by the NRC, submit
installation information on Form N-71, permit verification of that
information by the IAEA, and take other action as may be necessary to
implement the Safeguards Agreement, in the manner set forth in Sec.
75.6, and Sec. Sec. 75.11 through 75.14.
8. Transfers of Licenses--Creditors' Rights--Surrender of Licenses
a. Section 50.80, Transfer of Licenses
Section 50.80 implements Sections 101 and 184 of the AEA, which
require Commission approval for the transfer of a license for a
production or utilization facility, including a nuclear power reactor.
Section 50.80(a) explicitly refers to transfers of a ``license for a
production or utilization facility * * *,'' which would include
construction permits under part 50, as well as all licenses and permits
issued under part 52. However, to explicitly recognize the
applicability of Sec. 50.80(a) to both permits under parts 50 and 52
and all licenses under part 52, Sec. 50.80(a) is revised to explicitly
refer to permits under parts 50 and 52, and licenses under part 52. The
proposed rule would have only made these clarifying revisions. A
commenter on the proposed rule stated that some of the requirements in
Sec. 50.80 are not relevant to transfers of an ESP. The NRC agrees,
and has revised the final rule to specify which criteria are applicable
to transfer of an ESP. Specifically, paragraph (b)(1)(ii) requires an
application for transfer of an ESP to include as much of the
information described in Sec. Sec. 52.16 and 52.17 with respect to the
identity and technical qualifications of the proposed transferee as
would be required by those sections if the application were for an
initial license. This change removes the requirement for the applicant
for transfer of an ESP to address financial qualifications since this
is not required of an initial ESP applicant. In addition, this change
removes the provision that the NRC may require additional information
as part of an ESP transfer with respect to data on proposed safeguards
against hazards from radioactive materials and the applicant's
qualifications to protect against such hazards. Information on these
subject matters is not relevant to an ESP transfer, inasmuch as an ESP
does not authorize the holder to possess radioactive material.
The NRC declines to adopt the suggestion of a commenter who
suggested that the statement of considerations clarify when a transfer
of an ESP is necessary. The NRC's revision to Sec. 50.80 is a
conforming change to a procedural regulation, the process by which the
NRC processes and determines a transfer of a license. Section 50.80
does not, by itself, specify the circumstances for which a license
transfer is necessary; it simply addresses what procedures must be
followed if a license transfer request is received. Therefore, the NRC
does not believe that it is necessary or desirable to provide such
guidance in the context of this rulemaking.
b. Section 50.81, Creditor Regulations
Section 50.81 implements Section 184 of the AEA, which requires the
consent of the Commission for the creation of any mortgage, pledge or
other lien upon any Commission-licensed facility or special nuclear
material. To ensure that the reach of Sec. 50.81 is as broad as the
statutory requirement, the NRC is revising the definition of license
and facility. The definition of license in this section is revised to
explicitly refer to all licenses under 10 CFR, and early site permits
under part 52. The definition of facility is revised to add a new
paragraph which explicitly refers to an early site permit under part
52, and a reactor manufactured under a manufacturing license under part
52.
[[Page 49408]]
9. Amendment of License or Construction Permit at Request of Holder
a. Section 50.90, Application for Amendment of License or Construction
Permit; section 50.91, Notice for Public Comment; State Consultation;
and section 50.92, Issuance of Amendment
Sections 50.90, 50.91, and 50.92 govern the procedures and criteria
for NRC consideration and issuance of amendments to licenses and
construction permits. The regulations do not clearly address early site
permits, combined licenses, or manufacturing licenses. Accordingly, the
NRC is making a number of changes in these regulations.
Section 50.90 provides that applicants for amendment of a license
or construction permit must file their application with the NRC as
described in Sec. 50.4, following the form prescribed for the original
application. Although the term, license, as amended in Sec. 50.2
includes combined licenses, manufacturing licenses, and early site
permits under part 52, Sec. 50.92 is revised to explicitly refer to
these part 52 licenses to eliminate any confusion with respect to the
applicability of this section to part 52 licenses. A similar change is
made in the introductory paragraph of Sec. 50.91.
Sections 50.92 and 50.91(a)(4) implement the Commission's authority
under Section 189 of the AEA to dispense with the advance publication
of a Federal Register document requesting a hearing with respect to
license amendments, and to make operating license and combined license
amendments immediately effective upon issuance, if the NRC finds that
the amendment involves no significant hazards consideration. The NRC is
revising Sec. 50.92(c) to clarify that, consistent with Section 189 of
the AEA, the NRC may make a no significant hazards consideration
determination for amendments of combined licenses under part 52.
Combined licenses are explicitly mentioned in Section 189.a.(2)(A) of
the AEA with respect to immediate effectiveness following a Commission
determination of a no significant hazards consideration. In addition, a
combined license merges into a single license the authority otherwise
contained in a construction permit and an operating license, and the
language of Section 189.a.(1)(A) of the AEA which refers to both
amendments of construction permits and operating licenses, also applies
to amendments of combined licenses.
Finally, Sec. 50.92(a) is revised to provide that a separate
application for a construction permit is not required even where a
holder of a combined license or a manufacturing license must seek a
license amendment because of a material alteration. There is no safety
or regulatory benefit in requiring the licensee to concurrently submit
an application for a new construction permit in addition to a license
amendment, inasmuch as NRC review of the alteration is assured.
10. Revocation, Suspension, Modification, Amendment of Licenses and
Construction Permits, Emergency Operations by the Commission
a. Section 50.100, Revocation, Suspension, Modification of Licenses,
Permits, and Approvals for Cause
Section 50.100 is revised to explicitly address the Commission's
authority to suspend, modify, or revoke any standard design approval
under subpart E of parts 50 or 52 for any material false statement in
the application, or because of any statement in any report, record,
inspection, or condition revealed by the application, or by other
means, which would warrant the NRC to refuse to grant the design
approval on an original application. The former language of Sec.
50.100, which is retained as paragraph (a) in the final rule, applied
to any license or any license or construction permit issued under part
50 for any material false statement in the application for the license
or permit, or because of any statement in any report, record,
inspection, or condition revealed by the application, or by other
means, which would warrant the NRC to refuse to grant a license on an
original application, or for failure to construct or operate a facility
in accordance with the applicable license or permit. While this
language applies to early site permits, combined licenses and
manufacturing licenses, by virtue of their status as licenses under the
AEA, it does not clearly apply to standard design approvals as these
are not licenses. Nonetheless, the Commission possesses authority to
modify, suspend or revoke the regulatory approvals. Accordingly, the
NRC is revising this section to add a reference to a standard design
approval.
The final rule is different than the proposed rule in several ways.
A reference to part 50 is added in the clause governing revocations,
suspensions, and modifications of licenses. The word, ``provided * *
*,'' is revised to read ``provided, however,* * *.'' Finally, a
reference to a combined license is added to the clause stating that a
failure to meet the timely completion of proposed construction or
alteration is subject to Sec. 50.55(b) (which is also revised in this
final rulemaking to make its provisions applicable to combined
licenses).
11. Backfitting
a. Section 50.109, Backfitting
The backfit rule, 10 CFR 50.109, provides certain protection to
nuclear power plant licensees against changes in the NRC requirements
and NRC staff positions on those requirements. Prior to the final rule,
the backfitting provisions in Sec. 50.109 applied to standard design
approvals, construction permits, and operating licenses, but did not
address combined licenses or manufacturing licenses. Part 52 contains
special backfitting requirements on early site permits, design
certification rules, but prior to this rulemaking, neither Sec. 50.109
or part 52 addressed backfitting of a combined license, although the
NRC recognizes that backfitting restraints for an early site permit and
a design certification rule would apply to a combined license
referencing either or both. To address these gaps in backfitting, and
to clarify the application of special backfitting provisions, Sec.
50.109(a)(1) is revised by establishing the date that backfitting
protection begins for a manufacturing license, a construction permit
for a duplicate design license, and a combined license. Moreover, with
respect to a part 50 construction permit, a part 50 operating license,
and a part 52 combined license, Sec. 50.109 is revised by listing the
specific backfitting restrictions that apply if an early site permit,
standard design approval, or standard design certification rule is
referenced, or if a nuclear power reactor manufactured under a part 52
manufacturing license is used.
In the statement of considerations for the 2006 proposed rule, the
Commission asked whether, instead of conforming the language of Sec.
50.109 to reflect the licensing and regulatory approval processes in
part 52, the Commission should adopt a general backfitting provision,
analogous to Sec. 50.109, in part 52. Commenters either expressed no
opinion on the matter, or otherwise indicated that they did not have a
preference. Accordingly, the Commission has decided to revise Sec.
50.109 to include the conforming changes, rather than adopting a
backfitting provision in part 52.
[[Page 49409]]
12. Enforcement
a. Section 50.120, Training and Qualification of Nuclear Power Plant
Personnel
This section sets forth the requirements for training and
qualifying nuclear power plant personnel. In a conforming change, the
NRC is revising Sec. 50.120 to add applicants for and holders of
combined licenses as being subject to this provision.
13. Appendices
a. Appendix A to Part 50--General Design Criteria for Nuclear Power
Plants
The first paragraph of the Introduction to appendix A to part 50 is
revised to clarify that the general design criteria in appendix A to
part 50 apply to applications for combined licenses, design approvals,
design certification, and manufacturing licenses, as well as for
construction permits. Also, General Design Criterion (GDC) 19 of
appendix A to part 50, which sets forth requirements for a main control
room in a nuclear power plant, is revised to clarify that the radiation
protection requirements in GDC 19 for applications filed after January
10, 1997, apply to design approvals and manufacturing licenses issued
under part 52, in addition to design certifications and combined
licenses.
b. Appendix B to Part 50--Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Appendix B to part 50 states that every applicant for a
construction permit is required to include in its preliminary safety
analysis report a description of the quality assurance program to be
applied to the design, fabrication, construction, and testing of the
SSCs of the facility and every applicant for an operating license is
required to include, in its FSAR, information pertaining to the
managerial and administrative controls to be used to assure safe
operation. The NRC is revising appendix B to part 50 to clarify that
these requirements also apply to early site permits, design approvals,
design certifications, combined licenses, and manufacturing licenses
under 10 CFR part 52. Specifically, the introduction to appendix B to
part 50 is revised to state that every applicant for a combined license
is required by the provisions of Sec. 52.79 to include in its FSAR a
description of the quality assurance program applied to the design, and
to be applied to the fabrication, construction, and testing of the SSCs
of the facility and to the managerial and administrative controls to be
used to assure safe operation. The introduction also states that, for
applications submitted after the effective date of the final rule,
every applicant for an early site permit is required by the provisions
of Sec. 52.17 to include in its site safety analysis report a
description of the quality assurance program applied to site activities
related to the design, fabrication, construction, and testing of the
SSCs of a facility or facilities that may be constructed on the site.
The introduction states that every applicant for a design approval or
design certification is required by the provisions of Sec. Sec. 52.137
and 52.47, respectively, to include in its FSAR a description of the
quality assurance program applied to the design of the SSCs of the
facility. Finally, the introduction states that every applicant for a
manufacturing license is required by the provisions of 10 CFR 52.157 to
include in its FSAR a description of the quality assurance program
applied to the design, and to be applied to the manufacture of, the
SSCs of the reactor. The wording in appendix B of part 50 and in the
related provisions in the contents of application sections in 10 CFR
part 52 is modified slightly in the final rule to reflect that some
activities have already occurred when the application is submitted
(e.g., design of SSCs for design certification applicants). Therefore,
instead of requiring that the application describe the QA program ``to
be applied'' to these activities, the final rule requires that the
application describe the QA program ``applied'' to these activities,
since they have already occurred.
The NRC is maintaining the current regulatory structure for
requirements that implement appendix B to part 50 whereby QA for
construction activities is governed by Sec. 50.55(f), and QA for
operation is governed by Sec. 50.54(a). Because a combined license
under part 52 authorizes both construction and operation, a combined
license holder should be subject to the QA requirements in Sec.
50.55(f) from the date of issuance of the combined license until the
Commission makes the finding under Sec. 52.103(g) that allows the
licensee to load fuel and operate. Thereafter, the combined license
holder should be governed by the QA requirements in Sec. 50.54(a). The
manufacture of a nuclear power reactor under a manufacturing license is
the functional equivalent of construction. Accordingly, the NRC is
revising Sec. 50.55(f) to refer to holders of manufacturing licenses
under part 52. Early site permits under subpart A precede construction
and are considered partial construction permits. Hence the NRC believes
that they should be subject to QA under Sec. 50.55(f), and Sec.
50.55(f) is revised accordingly.
Appendix B to part 50 was formerly applicable to combined licenses
under the provisions of Sec. 52.83, which states that all provisions
of 10 CFR part 50 and its appendices applicable to holders of operating
licenses also apply to holders of combined licenses. Appendix B to part
50 formerly applied to design certifications by virtue of the provision
in former Sec. 52.48, which stated that design certification
applications will be reviewed for compliance with the standards set out
in 10 CFR part 50 as they apply to applications for construction
permits and operating licenses for nuclear power plants, and as those
standards are technically relevant to the design proposed for the
facility. Former appendix O to part 52, Section O.3, required
applicants for design approvals to include the information required by
Sec. Sec. 50.34(a) and (b), as appropriate, and stated that the
information required by Sec. 50.34(a)(7) (a description of the quality
assurance program and a discussion of how the applicable requirements
of appendix B to part 50 will be satisfied), shall be limited to the QA
program to be applied to the design, procurement and fabrication of the
SSCs for which design review has been requested. Appendix B to part 50
formerly applied to manufacturing licenses by virtue of the provision
in former appendix M to part 52, Section M.1, which stated that the
provisions in part 50 applicable to construction permits apply in
context, with respect to matters of radiological health and safety,
environmental protection, and the common defense and security, to
manufacturing licenses.
Early site permits are considered partial construction permits,
therefore, the NRC believes that they should be subject to the QA
requirements of appendix B to part 50. Section 52.39, with certain
specific exceptions, requires the Commission to treat matters resolved
in an early site permit proceeding as resolved in making findings for
issuance of a construction permit, operating license, or combined
license. Because of this finality, conclusions made during the early
site permit phase will be relied upon for use in subsequent design,
construction, fabrication, and operation of a reactor that might be
constructed on the site for which an early site permit is issued.
Therefore, the NRC believes that the level of quality used to control
activities related to safety-related SSCs should be equivalent in the
early site permit and combined license phases. For these reasons,
applicants must apply quality
[[Page 49410]]
controls to each early site permit activity associated with the
generation of design information for safety-related SSCs that meet the
criteria in appendix B to part 50. Therefore, the NRC is revising
appendix B to part 50 to make it applicable to early site permits.
c. Appendix C to Part 50--A Guide for the Financial Data and Related
Information Required To Establish Financial Qualifications for
Construction Permits and Combined Licenses
Section 182.a of the AEA requires an applicant for a license for a
production or utilization facility to submit information in its
application * * * ``as the Commission, regulation, may determine to be
necessary to decide such of the technical and financial qualifications
of the applicant * * * as the Commission may deem appropriate for the
license.'' The NRC has long determined the need for non-utility
applicants for nuclear power plant construction permits and operating
licenses to establish their financial qualifications (see 10 CFR
50.33(f)), and has set forth the specific information on financial
qualifications to be provided by applicants for construction permits in
appendix C to part 50. Inasmuch as holders of combined licenses under
part 52 are authorized to perform the same construction activities with
respect to a nuclear power plant as a holder of a construction permit
under part 50, the NRC believes that applicants for combined licenses
should be subject to the requirements of appendix C to part 50.
Accordingly, the title of appendix C is revised to make clear the
applicability of this appendix to applicants for combined licenses.
This change constitutes a conforming change to the revision of Sec.
50.33.
With the exception of manufacturing licenses, none of the other
regulatory processes under part 52, e.g., early site permits, standard
design certifications, and standard design approvals, authorize any
activities constituting ``construction'' under the AEA and the
Commission's regulations.\7\ Therefore, the final rule does not refer
to early site permits, design certifications, or design approvals under
part 52. With respect to a reactor manufacturing license, the NRC does
not believe that a financial qualifications review is necessary for
several reasons. A financial qualifications review at the manufacturing
license stage would appear to be redundant to the financial
qualifications review that is already necessary at the construction
permit and operating license stages, or combined license stage.
Sufficient safety and quality assurance reviews, including the use of
ITAAC in the case of a combined license, should be sufficient to
address any adverse impacts on safety as the result of inadequate
financial resources to properly manufacture the reactor. Furthermore,
the NRC notes that manufacture of a reactor is, in many respects, no
different than fabrication of components and systems by third party
vendors, who are not required to obtain an NRC license and demonstrate
financial qualifications. There seems to be no regulatory value to
mandate a financial qualifications review of manufacturing license
applicants, when this type of review is not conducted by the NRC for
fabricators of nuclear power plant systems and components.
---------------------------------------------------------------------------
\7\ Although early site permit applicants may seek the authority
to conduct activities allowed under 10 CFR 50.10(e)(1) (but not
activities allowed under Sec. 50.10(e)(3), see Sec. 52.17(c)),
these activities are not considered ``construction.''
---------------------------------------------------------------------------
d. Appendix E to Part 50--Emergency Planning and Preparedness for
Production and Utilization Facilities
See discussion in Section V.D.4.f of this document.
e. Appendix I to Part 50--Numerical Guides for Design Objectives and
Limiting Conditions for Operation To Meet the Criterion ``as Low as is
Reasonably Achievable'' for Radioactive Material in Light-Water-Cooled
Nuclear Power Reactor Effluents
The Commission is revising appendix I to part 50 to conform to the
changes in Sec. Sec. 50.34a and 50.36a which are being made as part of
this final rule. Specifically, a statement is added in Section I of
appendix I to part 50, stating that Sec. Sec. 52.47, 52.79, 52.137,
and 52.157 provide that applications for design certification, combined
license, design approval, or manufacturing license, respectively, shall
include a description of the equipment and procedures for the control
of gaseous and liquid effluents and for the maintenance and use of
equipment installed in radioactive waste systems. In addition, Section
II of appendix I to part 50 is revised to state that the guides on
design objectives set forth in appendix I to part 50 may be used by an
applicant for a combined license as guidance in meeting the
requirements of Sec. 50.34a(d) or by an applicant for a design
approval, a design certification, or a manufacturing license as
guidance in meeting the requirements of Sec. 50.34a(e). Section IV of
appendix I to part 50 is revised to state that the guides on limiting
conditions for operation for light-water-cooled nuclear power reactors
in appendix I to part 50 may be used by an applicant for an operating
license or a design certification or combined license, or a licensee
who has submitted a certification of permanent cessation of operations
under Sec. 50.82(a)(1) or Sec. 52.110 as guidance in developing
technical specifications under Sec. 50.36a(a) to keep levels of
radioactive materials in effluents to unrestricted areas as low as is
reasonably achievable. Finally, Section V of appendix I to part 50 is
revised to state that the guides for limiting conditions for operation
set forth in appendix I are applicable to any application filed on or
after January 2, 1971, for a construction permit for a light-water-
cooled nuclear power reactor, or a design certification, a combined
license, or a manufacturing license for a light-water-cooled nuclear
power reactor under part 52. Note that the NRC added the phrase ``for a
light-water-cooled nuclear power reactor'' to Section V in the final
rule. This phrase was inadvertently left out of the introduction to
Section V in the proposed rule. The NRC did not intend to change the
applicability of appendix I in this rulemaking and is, therefore,
correcting this omission in the final rule. The NRC has also removed
the conforming change it had proposed to paragraph A.3 of the
Concluding Statement of Position of the Regulatory Staff (Docket-RM-50-
2) Guides on Design Objectives for Light-Water-Cooled Nuclear Power
Reactors in appendix I. The design objectives in this staff position
are only applicable to those light-water-cooled nuclear power reactors
that applied for a construction permit before January 2, 1971 (per
Appendix I, Section V, B.2.). Because part 52 did not exist before
1971, the proposed change is unnecessary.
f. Appendix J to Part 50--Primary Reactor Containment Leakage Testing
for Water-Cooled Power Reactors
Section 50.54(o) provides a condition for all operating licenses
for water-cooled power reactors that primary reactor containments must
meet the containment leakage test requirements set forth in appendix J
to part 50. These test requirements provide for preoperational and
periodic verification by test of the leak-tight integrity of the
primary reactor containment, and systems and components which penetrate
containment of water-cooled power reactors, and establish the
acceptance criteria for these tests. The purpose of the tests are to
assure that leakage through the primary reactor containment systems and
components penetrating primary containment shall not exceed allowable
leakage rate values
[[Page 49411]]
as specified in the technical specifications or associated bases, and
periodic surveillance of reactor containment penetrations and isolation
valves is performed so that proper maintenance and repairs are made
during the service life of the containment, and systems and components
penetrating primary containment. The Commission is revising appendix J
to clarify that these requirements also apply to combined licenses
under 10 CFR part 52. This is consistent with former Sec. 52.83, which
stated that all provisions of 10 CFR part 50 and its appendices
applicable to holders of operating licenses also apply to holders of
combined licenses.
g. Appendices M and O to Part 50 [Removed]
The NRC has removed appendices M and O from 10 CFR part 50.
Appendix M provided for issuance of a license authorizing the
manufacture of a nuclear power reactor to be incorporated into a
nuclear power plant under a construction permit and operated under an
operating license at a different location from the place of
manufacture. Appendix O addressed the approval of standard designs for
nuclear power reactors. These appendices were transferred to 10 CFR
part 52 when it was first issued (54 FR 15372; April 18, 1989).
However, the NRC failed to remove those appendices from 10 CFR part 50,
though the NRC intended to do so (see 54 FR 15385; April 18, 1989).
h. Appendix S to Part 50--Earthquake Engineering Criteria for Nuclear
Power Plants
Appendix S to part 50 provides earthquake engineering criteria for
nuclear power plants and applies to applicants for a design
certification or combined license under part 52 or a construction
permit or operating license under part 50. The final rule revises
appendix S to clarify that the requirements in appendix S also apply to
applicants for design approvals and manufacturing licenses issued under
10 CFR part 52. Although former appendix O to part 52 did not
explicitly require applicants for design approvals to comply with the
requirements of appendix S, the NRC is requiring design approval
holders to comply with appendix S to part 50 because the NRC believes
that the requirements for a design approval should be the same as the
requirements for a design certification, given that the reviews
performed by the NRC staff for the two products are essentially
identical. Finally, appendix S formerly applied to manufacturing
licenses by virtue of former appendix M to part 52, Section M.1, which
stated that the provisions in part 50 applicable to construction
permits apply in context, with respect to matters of radiological
health and safety, environmental protection, and the common defense and
security, to manufacturing licenses. Therefore, the Commission is
revising the General Information section of appendix S to part 50 to
state that the appendix applies to applicants for a design
certification, design approval, combined license, or manufacturing
license under 10 CFR part 52 or a construction permit or operating
license under 10 CFR part 50. The NRC also made conforming changes to
the Introduction, paragraph (a) to appendix S to part 50, and added
definitions for design approval and manufacturing license to Section
III of appendix S to part 50, to be consistent with the definitions in
proposed part 52.
E. Change to 10 CFR Part 1
1. Section 1.43, Office of Nuclear Reactor Regulation
Section 1.43 describes the responsibilities of the Office of
Nuclear Reactor Regulation (NRR), which includes the development and
implementation of regulations, policies, programs and procedures for
the receipt, possession or ownership of source, byproduct and special
nuclear material that is used or produced at nuclear power plants.
Inasmuch as power plants may be licensed under part 52 as well as part
50, Sec. 1.43(a)(2) is revised to clarify that NRR has authority over
the development and implementation of regulations, policies, programs
and procedures for the receipt, possession or ownership of source,
byproduct and special nuclear material that is used or produced at
nuclear power plants licensed under part 52. In addition, a correction
has been made to reference part 54, to clarify that NRR has the same
authority with respect to renewed operating licenses for nuclear power
plants.
F. Changes to 10 CFR Part 2
1. Section 2.1, Scope
The statement of scope for part 2 is revised by adding a reference
to rulemaking and standard design approvals. Previously, the scope
statement did not mention rulemakings, even though subpart H of part 2
applied to rulemakings, nor did it mention standard design approvals
even though the NRC processed applications for design approvals in
accordance with the procedures in part 2. Accordingly, the change in
the statement of scope for part 2 correctly reflects the applicability
of its procedures to both rulemaking and the processing of standard
design approvals.
2. Section 2.4, Definitions
The definitions of contested proceeding, license, and licensee, are
revised in part 2 by adding conforming references, as appropriate, to
the licensing processes in part 52. The revised definition of contested
proceeding clarifies that contested proceedings include those involving
permits, such as early site permits and construction permits. The
revised definition of license, ensures that early site permits and
construction permits, as well as part 52 combined licenses and
manufacturing licenses, are considered to be licenses for purposes of
part 2. Similarly, the revised definition of licensee ensures that
holders of early site permits and construction permits, as well as
combined licenses and manufacturing licenses, are considered to be
licensees for purposes of part 2.
3. Section 2.100, Scope of Subpart
This section is revised by adding conforming references to issuance
of a standard design approval under subpart E of part 52.
4. Section 2.101, Filing of Application
This section, which governs the procedures for, and the timing and
content of applications, has been revised in several respects.
Paragraphs (a)(1), (a)(2), the introductory paragraph of (a)(3),
paragraph (a)(3)(iii), and paragraph (a)(4) are revised by adding
conforming references to combined licenses, early site permits, and
standard design approvals. The Commission notes that the former
language of Sec. 2.101 already applied to combined licenses, as well
as early site permits, inasmuch as they are both licenses. Nonetheless,
consistent with the revisions to the definitions of license and
licensee, Sec. 2.101 has been revised to explicitly refer to early
site permits, as applicable.
In response to public comment on the proposed rule, paragraph
(a)(5) of Sec. 2.101 and paragraph (a-1) are revised to allow
applicants for combined licenses--as well as applicants for
construction permits as provided under this section--to submit
applications in parts. Paragraph (a)(5) of the final rule allow
applicants for combined licenses and construction permits to submit an
application in two parts, with one part containing the environmental
report required under Sec. 50.30(f) if the application is for a
construction permit or Sec. 52.80(b) if the application is for a
combined license. The other part must
[[Page 49412]]
contain the information required by Sec. Sec. 50.34(a) and 50.34a if
the application is for a construction permit, or Sec. 52.79 and Sec.
52.80(a) if the application is for a combined license. In addition, the
part that is filed first must contain the information required by Sec.
50.33, Sec. 50.34(a)(1) if the application is for a construction
permit, Sec. 52.79(a)(1) if the application is for a combined license,
and Sec. 50.37. There are no considerations unique to combined
licenses which would weigh against allowing a combined license
applicant to submit a two part application under paragraph (a)(5) of
Sec. 2.101. Accordingly, the Commission is adopting this change in the
final rulemaking. Inasmuch as the revisions are to the Commission's
rules of procedure and practice, the Commission may adopt them in final
form without further notice and comment, under the rulemaking
provisions of the APA, 5 U.S.C. 553(b)(A).
Paragraph (a-1) of Sec. 2.101 allows applicants for combined
licenses, as well as applicants for construction permits, to submit an
application in parts to allow for early consideration and a presiding
officer's partial initial decision on those site suitability matters
for which the applicant seeks NRC resolution. The provisions governing
early consideration of site suitability issues in a combined license
proceeding are set forth in paragraph (a-1)(2). Under this paragraph, a
combined license application may be submitted in three parts, with the
first part containing information on the site suitability issues which
the applicant wishes to have resolved first. The second and third
parts, which constitute the remainder of the application as described
in paragraph (a-1)(2)(ii) and (iii), must be submitted during the
period that the partial decision on part one is effective, viz., 5
years under new Sec. 2.627 in subpart F of part 2. There are no
considerations unique to combined licenses which would weigh against
allowing a combined license applicant to obtain early consideration of
site suitability issued under paragraph (a-1). As with the change to
paragraph (a)(5), this revision to paragraph (a-1) constitutes
revisions to the Commission's rules of procedure and practice.
Accordingly, the Commission may adopt them in final form without
further notice and comment, under the rulemaking provisions of the APA,
5 U.S.C. 553(b)(A).
5. Section 2.102, Administrative Review of Application
This section is revised by adding conforming references in Sec.
2.102(a) to applications for early site permits, standard design
approvals, combined licenses, and manufacturing licenses under part 52.
Under the revised section, the NRC staff will establish a review
schedule for an application for these processes, thereby treating the
applications the same as applications for construction permits or
operating licenses.
6. Section 2.104, Notice of Hearing
Section 2.104 sets forth the NRC's requirements regarding
publication in the Federal Register of notice of hearings. The former
rule, as well as the proposed part 52 rule, specified the nature of the
issues that the presiding officer must address in both uncontested and
contested proceedings. The NRC has decided, based upon its experience
in noticing hearings in the last decade (in which the Commission's
notices for more significant proceedings have varied from requirements
in this section), as well as its consideration of the nature of
mandatory hearings under Section 189 of the AEA, that much of this
detailed prescription of the content of the notice of hearing should be
removed from Sec. 2.104.
Accordingly, the language of Sec. 2.104 has been considerably
truncated from the former rule. Paragraph (a) is largely the same as
former paragraph (a). However, paragraph (b) has been modified to
specify only the requirements of the notice of hearing which are common
to all proceedings. All provisions in the former Sec. 2.104 specifying
the issues to be addressed by the presiding officer are removed in the
final rule. Inasmuch as this revision is to the NRC's rules of
procedure and practice, the NRC may adopt them in final form without
further notice and comment, under the rulemaking provisions of the APA,
5 U.S.C. 553(b)(A).
Paragraph (c), (paragraph (m) in the proposed rule, former
paragraph (e)) requires the NRC to transmit a notice of hearing on an
initial application of a license for a production or utilization
facility to an appropriate state official and the chief executive of
the municipality or county in which the facility is to be located or an
activity is to be conducted. In addition to the redesignation,
paragraph (c) is revised to clarify that the notice must be provided
for applications for early site permits, combined licenses, but not
manufacturing licenses. Manufacturing licenses are excluded from the
notification provisions because the NRC is not licensing any particular
location or site where manufacturing may occur (see discussion of the
manufacturing license concept).
7. Section 2.105, Notice of Proposed Action
Section 2.105 contains the NRC's procedures for notices of proposed
actions where a hearing is not required by law and if the Commission
has determined that a hearing is in the public interest. Inasmuch as
amendments to combined licenses and manufacturing licenses do not
require a mandatory hearing under the AEA, Sec. 2.105(a)(4) is revised
to clarify that the procedures in Sec. 2.105 also apply to
applications for amendments of combined licenses and manufacturing
licenses. Furthermore, because the AEA does not require a mandatory
hearing for the initial issuance of manufacturing licenses, paragraph
(a)(13) is added in the final rule to provide for publication of a
notice of proposed action in connection with an application for a
manufacturing license under subpart F of part 52.
Under Sec. 52.103(a), which implements Section 189.a(1)(B)(i) of
the AEA, the NRC is required to publish in the Federal Register a
notice of intended operation and an opportunity to request a hearing
with respect to compliance of the facility with inspections, tests, and
acceptance criteria in a part 52 combined license. Accordingly, the NRC
is revising Sec. 2.105 by adding Sec. 2.105(a)(12) which addresses
the information to be contained in the Federal Register notice required
by Sec. 52.103(a).
Because the Commission's authorization for a combined license
holder to operate under Sec. 52.103 does not constitute ``issuance''
of a license or amendment under Sec. 2.106, Sec. 2.105(b)(3) is added
indicating that the Commission will publish a notice of intended
operation in the Federal Register that identifies the proposed Agency
action as making the finding under Sec. 52.103(g). Paragraph
(b)(3)(iii) of the proposed rule, which would have required that the
Commission publish, as part of that Federal Register notice, a finding
that ITAAC have been met, has not been included in the final rule. This
is because Commission may not have made, at the time of the Federal
Register notice, the finding that all ITAAC have been met. After
careful review of the language of Section 189 of the AEA, the
Commission concludes that the Federal Register notice required by
Section 189.a(1)(B)(i) need not include a finding that ITAAC have been
met. Accordingly, Sec. 2.105(b)(3) of the final rule does not include
a requirement for such a finding to be
[[Page 49413]]
included in the Federal Register notice of intended operation.
8. Section 2.106, Notice of Issuance
Section 2.106(a) formerly provided that the NRC will publish in the
Federal Register a notice of issuance of a license or amendment of a
license where a notice of proposed action has been previously
published, and notice of amendment of a nuclear power plant license.
However, that language did not require publication in the Federal
Register that the Commission has made the finding under Sec.
52.103(g). Although the AEA does not require publication of a notice of
the Commission finding under Sec. 52.103, the Commission believes that
this publication is desirable as a matter of public transparency and
consistency with past practice of the Federal Register publication of
Commission action with similar effects (i.e., the issuance of a nuclear
power plant operating license). Accordingly, Sec. 2.106(a) is revised
to require Federal Register publication of the Commission finding under
Sec. 52.103.
Section 2.106(b)(2) is also revised to set forth the minimum
requirements for the contents of a Federal Register notice of action,
e.g., the manner in which copies of the safety analyses, if any, may be
obtained and examined, and a finding that the prescribed inspections,
tests, and analyses have been performed and that the acceptance
criteria prescribed in the combined license have been met, and that the
license complies with the requirements of the AEA and the NRC's
regulations. These provisions are the same as the existing requirements
with respect to notices of issuance for licenses and license
amendments, but adds the requirements with respect to ITAAC mandated by
Section 185 of the AEA and part 52. The NRC disagrees with the
contention raised by the nuclear industry that Section 185 of the AEA
limits the NRC to a finding of compliance with respect to ITAAC under
Sec. 52.103(g). Nothing in the legislative history suggests that by
adopting Section 185 of the AEA, Congress intended to override the
NRC's long-standing practice of making findings of compliance with the
Act and the Commission regulations when issuing nuclear power plant
licenses.
9. Section 2.109, Effect of Timely Renewal Application
Section 2.109 is revised to add conforming references to a combined
license under subpart C of part 52. The revised language clarifies that
an application for a combined license filed no later than 5 years
before its expiration will not be deemed to have expired until the
renewal application has been finally determined.
10. Section 2.110, Filing and Administrative Action on Submittals for
Standard Design Approval or Early Review of Site Suitability Issues
In a conforming change, paragraphs (a) and (b) of Sec. 2.110 are
revised to refer to subpart E of part 52 and appendix Q of part 50.
Paragraph (c) is corrected by adding Sec. 2.110(c)(2) to address the
procedures applicable to administrative determinations of submittals
for early review of site suitability issues; formerly, paragraph (c)
only refers to standard designs.
11. Section 2.111, Prohibition of Sex Discrimination
This section prohibits sex discrimination against certain persons
with respect to, inter alia, a license under the AEA. This section is
revised to include standard design approvals under part 52, and
petitions for rulemaking, including an application for a design
certification under part 52.
12. Section 2.202, Orders
This section is revised by redesignating Sec. 2.202(e) as Sec.
2.202(e)(1), and adding Sec. Sec. 2.202(e)(2) through (5), to indicate
the backfitting provisions in part 52 applicable to the various
licensing processes under part 52. No provisions were deemed necessary
to address issuance of orders representing backfitting of NRC approvals
such as standard design approvals.
13. Section 2.309, Hearing Requests, Petitions To Intervene,
Requirements for Standing, and Contentions
Section 2.309, which establishes the NRC requirements governing
requests for hearing and petitions to intervene--including submission
of contentions--is revised to add three conforming and clarifying
changes. First, paragraph (a) is revised, consistent with a change to
Sec. 52.103(c), to make clear that in a proceeding under Sec. 52.103,
the Commission itself will act as the presiding officer, will consider
and act upon a request for a hearing under Sec. 52.103, and will also
determine whether a period of interim operation may be permitted, as
provided for under Section 189.a(1)(B)(iii) of the AEA. Inasmuch as the
Commission itself will make the contention admission determination,
there should be no need for further Commission review of the contention
admission decision at the end of the hearing.
Second, paragraph (f)(1)(i) has been revised to make clear that
contentions in Sec. 52.103(b) requests for hearing must raise issues
in law or fact with respect to whether one or more of the acceptance
criteria in a combined license have not been, or will not be met, and
that the specific operational consequences of nonconformance would be
contrary to providing reasonable assurance of adequate protection to
public health and safety. This is consistent with the statutory
limitation on the scope of a hearing in Section 189.a(1)(B)(ii) of the
AEA.
Third, a new paragraph (f)(1)(vii) has been added to set forth the
specific requirements for a contention under Section 189.a(1)(B)(ii)
and 10 CFR 52.103(b). The new paragraph provides that, in a request for
hearing under Sec. 52.103(b), the information submitted must be
sufficient and include supporting information showing, prima facie,
that: (i) One or more of the acceptance criteria in a combined license
have not been, or will not be met, and (ii) the specific operational
consequences of nonconformance would be contrary to providing
reasonable assurance of adequate protection to public health and
safety. The revision also makes clear that the information in support
of a contention that an acceptance criterion is not, or will not be
met, must identify the specific portions of the Sec. 52.99(c) report
which is inaccurate, incorrect, or incomplete. The terms,
``inaccurate,'' and ``incorrect,'' while somewhat overlapping, are
intended to cover a broad range of situations. ``Inaccurate'' is
intended to address a situation where information contained in,
referenced by, or relied upon (either explicitly or implicitly) as a
supporting basis for a representation in a Sec. 52.99(c) report, is
erroneous (e.g., an erroneous computation, or inaccurate data entry of
a test result). By contrast, ``incorrect'' focuses on a situation where
such information is the result of a cognitive inadequacy or failure
(even if, under the circumstances, the inadequacy or failure is
justifiable), poor judgement, negligence, or deliberate wrongdoing. By
``incomplete,'' the NRC means that the report does not provide the
information which must be provided in the report as required by Sec.
52.99. Furthermore, if the requestor contends that the Sec. 52.99(c)
report is incomplete, and the requestor contends that the incomplete
portion prevents the requestor from making the necessary prima facie
showing, then the requestor must also, as provided by Sec.
2.309(f)(1)(vii), explain why the deficiency (viz., the incomplete
nature of the report) prevents the requestor from making the necessary
prima facie
[[Page 49414]]
showing. The NRC believes that these changes to Sec. 2.309 will help
ensure that any 10 CFR 52.103 hearing on whether the acceptance
criteria in ITAAC have been, or will be met, is focused only on the
matters which Congress intended to be adjudicated at this juncture, as
directed by Section 189.a.(1)(B) of the AEA.
Fourth, paragraph (g) is revised to conform with the change in: (i)
10 CFR 52.103(c), which now provides that the Commission will act as
the presiding officer in determining whether to grant or deny a request
for hearing with respect to whether acceptance criteria in ITAAC have
been or will be met; and (ii) 10 CFR 2.310, which provides that the
Commission, acting as the presiding officer, will determine the hearing
procedures to be utilized in a Sec. 52.103 hearing. Under the revised
paragraph (g), a request for hearing under Sec. 52.103 shall not
address the hearing procedures to be utilized.
Fifth, paragraph (h) is revised to prohibit a reply by a requestor
for a hearing under Sec. 52.103. The NRC believes that Congress
intended the Commission's initial decision to grant the hearing and the
determination of interim operation to be based upon the same set of
information. The Commission's view is based upon the language of
Section 189.a.(1)(B)(iii), which refers to a Commission determination
to allow a period of interim operation based upon the ``petitioner's
prima facie showing and any answers thereto. * * *'' That the statute
only refers to a request and the answers thereto suggests that Congress
did not intend that a reply was necessary. This is understandable given
Congress'' explicit direction that any hearing granted be completed
``to the maximum possible extent * * * within 180 days of the
publication of the notice [of opportunity to request a hearing under
Section 189.a(10)(B)(i)] or the anticipated date for initial loading of
fuel into the reactor, whichever is later.'' While the relevant
statutory language literally applies only to the Commission
determination of interim operation, the NRC believes that as a matter
of logic, Congress must have intended that it would also apply to the
threshold question of granting or denying the hearing request. It is
unclear why Congress would allow more information to be considered in
the threshold question of the hearing request, but limit the
information to be considered in the interim operation determination.
The NRC concludes that it would be closer to Congress' intention to
prohibit a requestor for a Sec. 52.103 hearing from replying to any
answers filed by the applicant and/or the NRC staff.
Finally, in a conforming change associated with the revision to
Sec. 52.103(c), paragraph (i) is revised to prohibit any ``appeal''
under Sec. 2.311 of a Commission decision to grant or deny a request
for hearing. Inasmuch as the Commission is acting as a presiding
officer, there can be no further ``appeal'' to a higher agency
decisionmaker. Moreover, an adversely affected party may seek
reconsideration of the Commission's decision under Sec. 2.345, and it
would be duplicative to afford an adversely-affected party a Sec.
2.311 ``review'' right in addition to the opportunity to seek
reconsideration under Sec. 2.345.
Inasmuch as these revisions are to the NRC's rules of procedure and
practice, the NRC may adopt them in final form without further notice
and comment, under the rulemaking provisions of the APA, 5 U.S.C.
553(b)(A).
14. Section 2.310, Selection of Hearing Procedures
Section 2.310 is revised, in part to conform with the change in 10
CFR 52.103(c), which now provides that the Commission will act as the
presiding officer in determining whether to grant or deny a request for
hearing with respect to whether acceptance criteria in ITAAC have been
or will be met. The revised Sec. 2.310 now provides that the
Commission will determine the hearing procedures to be utilized in its
determination on a hearing request under Sec. 52.103, as well as the
hearing procedures to be utilized in resolving admitted contentions
under Sec. 52.103(c) and (g).\8\
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\8\ The NRC notes that 10 CFR 2.309 does not apply, by its
terms, to petitions to modify the terms and conditions of a combined
license under 10 CFR 52.103(f). Such petitions must meet the
requirements of 10 CFR 2.206.
---------------------------------------------------------------------------
Inasmuch as this revision is to the NRC's rules of procedure and
practice, the NRC may adopt it in final form without further notice and
comment, under the rulemaking provisions of the APA, 5 U.S.C.
553(b)(A).
15. Section 2.340, Initial Decision in Certain Contested Proceedings;
Immediate Effectiveness of Initial Decisions; Issuance of
Authorizations, Permits, and Licenses
Section 2.340 addresses several different matters relating to the
presiding officer's initial decision and its effect. The final rule
reorganizes the paragraphs in this section in order to better
distinguish among these matters, reserves paragraphs (g) and (h) for
future use by the Commission, and makes substantial changes to these
matters addressed in this section, as discussed below. These changes
are to the NRC's rules of procedure and practice, and the NRC is
adopting the changes in final form without further notice and comment,
under the rulemaking provisions of the APA, 5 U.S.C. 5, 553(b)(A).
Scope of Presiding Officer's Initial Decision
Formerly, paragraph (a) limited the scope of the presiding
officer's findings and conclusions of law in initial decisions in
contested proceedings for production or utilization facility operating
licenses to matters put into controversy by the parties. Matters not
put into controversy by the parties could only be examined by the
presiding officer by direction of the Commission, either on its own
initiative or upon the presiding officer's referral of the matter to
the Commission. In a conforming change, a new paragraph (b) is added to
apply the limitation in contested hearings under Sec. 52.103(g) with
respect to whether the acceptance criteria in a combined license ITAAC
have been, or will be met.
The Sec. 2.340(a) limitation did not apply to a contested
utilization facility construction permit proceeding. Although the
statement of considerations for the original rulemaking adopting this
limitation (in former Sec. 2.760a) does not directly address the basis
for this limitation (see January 17, 1975; 40 FR 2973), the underlying
rationale may be gleaned from the Commission's order in Consolidated
Edison Co. of New York (Indian Point Nuclear Generating Unit 3), 8 AEC
7 (1974) which engendered the rulemaking. In explaining that the
Licensing Board has no obligation at the operating license stage to
inquire into matters which parties have not raised and the Licensing
Board itself has no reason to inquire, the Commission stated:
To have a Licensing Board engage in an idle exercise examining
issues just for the sake of examination--when the parties have not
raised such matters, and the Board is satisfied that there is
nothing to inquire about--would serve no useful purpose. This is
particularly true since an operating license proceeding is not to be
used to rehash issues already well ventilated and resolved at the
construction permit stage. Alabama Power Co. (Joseph M. Farley
Nuclear Plant, Units 1 and 2), CLI-74-12 (RAI-74-3-203).
Id. at 8. Thus, the limitation was based, in part, upon the broader
scope of inquiry for the presiding officer at construction permit
stage, which is a ``mandatory hearing'' required by
[[Page 49415]]
Section 189.a(1)(A). This rationale continues to apply today, and
consequently the NRC does not propose to alter the NRC's practice by
extending the Sec. 2.340(a)/Sec. 2.760a limitation to construction
permit (including early site permit) proceedings. Nor should the Sec.
2.340(a)/Sec. 2.760a limitation apply in a part 52 combined license
proceeding with respect to matters that would otherwise be addressed
and resolved in a construction permit issuance proceeding.
The final part 52 rule includes several changes to implement the
NRC's conclusions in this regard. Section 2.340(a) is revised to
provide that the presiding officer in a contested operating license
proceeding shall make findings of fact and conclusions of law to, inter
alia, those matters put into controversy or otherwise directed by the
Commission. Paragraphs (b), (c), and (d) are revised to address the
scope of the presiding officer's initial decision in a combined license
proceeding (including a renewal or amendment proceeding), in a
proceeding under Sec. 52.103(g), and in a manufacturing license
proceeding (including a renewal or amendment proceeding).
As discussed previously, the former Sec. 2.340(a)/Sec. 2.760a
limitation applied only to operating license proceedings, and did not
apply to other contested proceedings which do not require a ``mandatory
hearing,'' which includes most materials licensing proceedings (with
the notable exception of the licensing of a uranium enrichment
facility). The statement of consideration in this document merely
states that the rule codifies the Commission's Indian Point decision.
(see January 17, 1975; 40 FR 2973 (first column)). Inasmuch as the
Indian Point proceeding involved a utilization facility license, it is
likely that the Commission simply did not consider as part of the
rulemaking the possibility of applying the limitation to non-production
or utilization facility proceedings, as opposed to making a deliberate
decision not to apply the limitation to non-production or utilization
facility proceedings. Currently, the NRC believes that with 30
additional years of hearing experience, there is no practical,
compelling policy-based, or legal reason why the Sec. 2.340(a)
limitation should not be extended to non-production or utilization
facility proceedings. Accordingly, the NRC is revising Sec. 2.340 by
adding a new paragraph (e), which extends the existing limitation on
the presiding officer's initial decision in contested proceedings to
all other proceedings not covered by paragraphs (a) or (b) of Sec.
2.340. Although this change is not related to the part 52 rulemaking
effort, the NRC is adopting this change as part of the part 52 final
rule to ensure that stakeholders understand the provisions of Sec.
2.340 as an integrated whole.
Immediate Effectiveness of Presiding Officer's Initial Decision in
Production and Utilization Facility Proceedings
The remainder of former Sec. 2.340 was an amalgam of the
Commission's original rule (10 CFR 2.764 \9\) a presiding officer's
initial decision in certain proceedings was immediately effective upon
issuance, combined with newer provisions--first adopted in 1979 and
modified in 1981--which suspended the immediate effectiveness rule. The
``automatic stay'' provisions were adopted following the accident at
TMI-2, in order to provide for the Commission's direct involvement in
the issuance of nuclear power plant licenses. The Commission first
issued an Interim Statement of Policy and Procedure in October 1979,
which first noted that the TMI-2 accident was being investigated by the
NRC and may result in ``significant changes in the Commission's
regulatory policy and in the procedures it employs to license nuclear
power facilities.'' The Policy Statement then indicated that ``new
construction permits, limited work authorizations, or operating
licenses for any nuclear power plants shall be issued only after action
of the Commission itself.'' (See October 10, 1979; 44 FR 58559.) Soon
thereafter, on November 9, 1979 (44 FR 65049), the NRC issued a
Suspension of Sec. 2.764 and Statement of Policy on the Conduct of
Adjudicatory Proceedings. As part of this final rulemaking, the NRC
adopted a new appendix B to part 2 addressing the suspension of
immediate effectiveness provisions in Sec. 2.764, and providing for
both Atomic Safety and Licensing Appeal Board review and Commission
review of the presiding officer's initial decision.
---------------------------------------------------------------------------
\9\ 31 FR 12774 (September 30, 1966).
---------------------------------------------------------------------------
On May 28, 1981 (46 FR 28627), the NRC issued a final rule which
removed the need for the Appeal Board review of a presiding officer's
initial decision, but retained a minimum 60-day period for Commission
review. The final rule was almost immediately amended to exclude from
Commission review presiding officer decisions authorizing fuel load and
low-power testing (September 30, 1981; 46 FR 47764). In 2004, the
provisions in Sec. 2.764 were transferred without substantive change
to a new Sec. 2.340 as part of the general revision to 10 CFR part 2
(January 14, 2004; 69 FR 2182).
While the NRC's 1979 and 1981 rulemakings were justified in light
of the circumstances at that time, other factors now lead the NRC to
believe that the oversight provisions adopted in 1981 are no longer
necessary or desirable. In the 25 years since the adoption of the 1981
provisions, the NRC's regulatory framework and requirements for nuclear
power plants has evolved and strengthened. The NRC's technical
requirements for nuclear power reactors were substantially augmented in
the years immediately following the TMI accident, and thereafter have
evolved to reflect lessons learned, new information, and the increasing
acceptance of risk-informed methodologies. Similarly, the NRC's
oversight of nuclear power plants has evolved to reflect lessons
learned, new information, and the maturation of risk assessment
methodologies. Thus, the NRC believes its regulations may be revised to
remove the regulatory requirement for direct Commission involvement in
all production and utilization licensing proceedings. The Commission's
words in the May 1981 final rulemaking apply with more force today:
This amendment does not compromise the Commission's commitment
to the protection of public health and safety or to a fair hearing
process. Thorough technical safety reviews of license applications
by the NRC staff and the Advisory Committee on Reactor Safeguards,
the availability of public hearings on license applications, and the
Commission's inherent supervisory authority form the basis of the
network of procedural safeguards intended to implement this
commitment to a fair decision process and public health and safety.
(May 28, 1981; 46 FR 28628 first column)
The NRC's commitment remains unchanged, and the NRC's safeguards
have been strengthened since that time, for example, by refocusing the
regulatory process to include considerations of risk. In addition, the
NRC's rules of practice in part 2 provide several procedural safeguards
within the NRC's administrative process, including: (1) A petition for
presiding officer reconsideration under Sec. 2.345; (2) a petition for
Commission review under Sec. 2.341; and (3) a motion for a stay with
the presiding officer or the Commission under Sec. 2.342.
By removing the ``automatic stay'' provisions in former Sec.
2.340(f) and (g), the NRC's administrative process will be completed in
less time, thereby benefitting all parties from the reduction in
litigation resources without compromising the fairness of the overall
hearing process. Faster completion of
[[Page 49416]]
the adjudication will also enable aggrieved parties to more quickly
seek relief via an appeal to a U.S. Circuit Court of Appeals. The NRC
believes that Congress intends the Commission to conduct fair, but
efficient, hearings with respect to licensing, and to remove
unnecessary hearing procedures which do not contribute to such a
hearing process. This is evidenced by Section 189 of the AEA, as
amended by the Energy Policy Act of 1992, which directs the Commission
to issue, ``to the maximum possible extent,'' a final decision on
issues raised with respect to acceptance criteria by the anticipated
date for initial loading of fuel. The Commission concludes that the
changes to Sec. 2.340 are consistent with applicable law, and will
provide tangible benefits to all parties in NRC adjudications.
Immediate Effectiveness of Presiding Officer's Initial Decision in
Other, Non-Production or Utilization Facility Proceedings
As noted previously, the 1981 final rulemaking provided for an
``automatic stay'' to provide for direct Commission involvement in the
issuance of nuclear power plant licenses. Since that time, the NRC has
extended the ``automatic stay'' provisions in Sec. 2.340 to other
licensing contexts, such as independent spent fuel storage facilities
(ISFSIs) at sites away from nuclear power reactors, monitored
retrievable storage (MRO) licenses, and provided for a parallel
provision in 10 CFR part 61 for low-level waste (LLW) facilities, see
10 CFR 2.1211. The NRC did not explain the basis for requiring direct
Commission involvement in the issuance of a part 61 LLW license (see 47
FR 57446; December 27, 1982), although one could surmise from the
timing of the rulemaking that the factors underlying the 1981
rulemakings also were the basis for the 1982 rulemaking's provision
providing for direct Commission involvement in part 61 license
issuances. The NRC's original intent in requiring direct Commission
involvement in the issuance of specific ISFSI licenses and a MRS
license was the lack of regulatory experience (see, e.g., 60 FR 20879
and 20883; April 28, 1995), and, therefore, is somewhat different from
the motivating factors for the 1981 rulemakings. In any event, the NRC
now has had the benefit of experience in licensing a specific ISFSI, as
well as several specific ISFSIs located at reactor sites. Thus, the NRC
has come to a recognition that the safety, security and regulatory
issues associated with these licenses are of less complexity than those
associated with nuclear power plants, and that the NRC has greater time
to respond to potentially adverse situations. Compare 46 FR 47764,
47765 (issuance of licenses for activities involving minimal risk to
public health and safety, and greater time to take corrective action,
do not require Commission involvement). Furthermore, the Commission
possesses general supervisory authority over the NRC staff and may
direct the staff to keep the Commission appraised of licensing status
and issues for such licenses. Accordingly, the NRC concludes that there
is little regulatory benefit to be provided by a rule requiring direct
Commission involvement in the issuance of these licenses and that the
provisions in Sec. 2.340 providing for such involvement should also be
removed as part of this streamlining of the regulatory process.
Issuances of Authorizations, Permits, Licenses, and Sec. 52.103(g)
Findings
Former paragraph (c) of Sec. 2.340 provided that the appropriate
staff Office Director was authorized to issue certain delineated
licenses, including license amendments, construction permits, and
construction authorizations, within 10 days from the date of issuance
of an initial decision. The former language could be erroneously read
as requiring the Director to issue a license following an initial
decision on a contested matter, even if other issues not contested had
yet to be resolved by the NRC staff. In addition, paragraph (c) did not
address the issuance of a finding under Sec. 52.103(g). To resolve
these concerns, new paragraphs (i), (j), and (k) are added to Sec.
2.340. In general, each paragraph authorizes the appropriate staff
Office Director to issue the delineated license, permit, authorization
or finding within 10 days from the issuance of an initial decision, if
all other safety and environmental findings necessary for issuance of
the license, permit, authorization or finding have been made,
notwithstanding the pendency of various petitions or motions for
reconsideration, review or stay before the presiding officer or the
Commission.
Paragraph (i) authorizes the Director of Nuclear Reactor Regulation
(NRR) or the Director of the Office of New Reactors (NRO), as
appropriate, to issue nuclear power plant licenses, including
amendments, permits and authorizations, within 10 days of the initial
decision. Paragraph (j) authorizes the Commission or the appropriate
staff Office Director to make the finding under 10 CFR 52.103(g) that
the acceptance criteria in a combined license have been met. Finally,
paragraph (k) addresses the issuance of other licenses that are issued
by the Director of Nuclear Material Safety and Safeguards (NMSS).
Typical licenses of this type would be materials licenses for, inter
alia, medical uses, well logging, radiography, irradiators, and
research.
16. Section 2.341, Review of Decisions and Actions of a Presiding
Officer
This section addresses requests for review and appeals to the
Commission from a presiding officer's decision or actions in a hearing.
In a conforming change associated with the revision to Sec. 52.103(c),
paragraph (a)(1) of Sec. 2.341 is revised to explicitly prohibit a
party from seeking a ``review'' or an ``appeal'' of the Commission's
determination to allow a period of interim operation under Sec.
52.103(c), separate from and in addition to a request for
reconsideration under Sec. 2.345. Inasmuch as the Commission is acting
as the presiding officer in the Sec. 52.103(c) determination, there
can be no further ``appeal'' to a higher agency decisionmaker.
Moreover, it would be duplicative to afford a Sec. 2.341 ``review'' or
``appeal'' right in addition to the opportunity to seek reconsideration
under Sec. 2.345.
Inasmuch as this revision is to the NRC's rules of procedure and
practice, the NRC may adopt it in final form without further notice and
comment, under the rulemaking provisions of the APA, 5 U.S.C.
553(b)(A).
17. Section 2.347, Ex Parte Communications
Section 2.347, which sets forth the NRC's requirements governing ex
parte communications with the Commission and its adjudicatory
employees, is revised in this final rule to address several problems
with the current rule.
First, Sec. 2.347 is revised to make clear that ex parte
communication restrictions are not applicable in uncontested
proceedings. The APA requirements in 5 U.S.C. 557(d)(1) governing ex
parte communications apply only to communications ``relevant to the
merits of the proceeding * * *,'' which are made to and from
``interested persons outside the agency.'' In an uncontested
proceeding, there are no ``interested persons outside the agency,'' in
the sense that there are no persons for which a hearing has been
requested or intervention in a hearing has been granted. Hence, ex
parte communication restrictions do not apply. Moreover, as the NRC has
stated in the 2004 rulemaking revising 10 CFR part 2, Section 189 of
the AEA does not require NRC hearings under that section to be ``on the
record.'' See 69 FR 2183-2185, 2192-2193 (January 14, 2004).
[[Page 49417]]
Accordingly, Sec. 2.347 is revised to explicitly provide that ex parte
restrictions do not apply to uncontested proceedings.
Second, Sec. 2.347 is revised to exclude undisputed (i.e.,
uncontested) issues in contested proceedings from the application of ex
parte restrictions. It makes little sense to require the Commission to
inform parties to the proceeding of the Commission's communications
with the applicant or licensee on matters for which those parties have
not been admitted (and may have no interest in litigating). In
addition, the NRC believes that uncontested matters are not, for
purposes of applying the ex parte limitations in Section 557(d)(1) of
the APA, either ``a fact in issue'' or a matter which is ``relevant to
the merits of the [contested] proceeding.'' The NRC also believes, as
stated above, that the ex parte limitations in Section 557(d) of the
APA do not apply to NRC proceedings, and therefore the application of
ex parte restrictions in NRC proceedings is a matter of discretion on
the part of the NRC. The NRC believes that it is appropriate to exclude
undisputed issues from the application of ex parte limitations in
contested proceedings, inasmuch as there appears to be little, if any,
public confidence benefit from extending ex parte limitations to
``undisputed issues,'' i.e., matters which have not been raised by any
party in the proceeding.
Finally, Sec. 2.347 is also revised to make clear that ex parte
restrictions apply to matters which are the subject of a presiding
officer referral to the Commission under Sec. 2.340(a), and the
presiding officer's examination of that matter following Commission
approval under Sec. 2.340(a) (referred to as ``sua sponte'' issues at
53 FR 10361; March 31, 1988). The application of ex parte restrictions
to Sec. 2.340(a) ``sua sponte'' matters does not represent a change in
NRC practice, cf., 53 FR 10360, 10361 (first and second column) (March
31, 1988). Nonetheless, upon further reflection the NRC believes it is
inaccurate to treat Sec. 2.340(a) ``sua sponte'' matters as a
``disputed issue'' for purposes of applying Sec. 2.347. Accordingly,
the NRC is revising Sec. 2.347 to explicitly state that consideration
of Sec. 2.340(a) ``sua sponte'' matters are to be subject to ex parte
restrictions.
Inasmuch as these Sec. 2.347 revisions are to the NRC's rules of
procedure and practice, the NRC may adopt them in final form without
further notice and comment under the rulemaking provisions of the APA,
5 U.S.C. 553(b)(A).
18. Section 2.348, Separation of Functions
This section sets forth the NRC's requirements governing separation
of functions of the Commission and its adjudicatory employees when
acting in their adjudicatory capacity. The rule prohibits an NRC
officer or employee engaged in the performance of investigative or
litigation function in that proceeding from participating in or
advising the Commission and its adjudicatory employees about ``any
disputed issue in that proceeding * * *,'' with certain delineated
exceptions (10 CFR 2.348(a)).
The NRC believes that there are two problems with the current
language. First, the rule does not explicitly state that in an
uncontested proceeding, separation of functions does not apply. More
importantly, the rule applies separation of functions in circumstances
where it is not required by Section 554(d), viz., determinations
involving initial licenses (5 U.S.C. 554(d)(2)(A) of the APA). The NRC
recognizes that public confidence considerations may favor compliance
with separation of functions restrictions in contested initial
licensing proceedings. However, there is little apparent value in
applying separation of functions to the NRC's resolution of uncontested
(i.e., ``undisputed'') issues in contested proceedings. The NRC also
notes that (as in the case of the APA restrictions on ex parte
communications) the APA separation of functions requirements apply only
to adjudications which are required to be ``on the record.'' As
discussed above, NRC licensing proceedings are not required by the AEA
or any other statute to be on the record. Thus, there is no legal
requirement to apply separation of functions in initial licensing
proceedings. Although the NRC could voluntarily, as a matter of
discretion, apply separation of functions in circumstances where it is
not required by law, such a course of action seems unjustified in view
of the lack of a clear public confidence benefit--which is the primary
objective of separation of functions restrictions. For these reasons,
the final part 52 rule revises Sec. 2.348 to make explicit that
separation of functions requirements do not apply to either uncontested
proceedings, or to an undisputed issue in contested initial licensing
proceedings.
Section 2.348 is also revised to make clear that separation of
functions applies to matters which are the subject of a presiding
officer referral to the Commission under Sec. 2.340(a), and the
presiding officer's examination of that matter following Commission
approval under Sec. 2.340(a). As with the change in Sec. 2.347 with
respect to ex parte restrictions, this change in Sec. 2.348 does not
depart from the NRC's current practice of applying separation of
function restrictions to ``sua sponte'' matters under Sec. 2.340(a).
The NRC believes that it is more accurate to explicitly state that sua
sponte matters under Sec. 2.340(a) are subject to separation of
functions restrictions, rather than characterizing such matters as
``disputed issues.''
Inasmuch as these Sec. 2.348 revisions are to the NRC's rules of
procedure and practice, the NRC may adopt them in final form without
further notice and comment under the rulemaking provisions of the APA,
5 U.S.C. 553(b)(A).
19. Section 2.390, Public Inspections, Exemptions, Requests for
Withholding
Section 2.390 governs the availability of NRC records and documents
regarding a license, permit or order, and implements the Freedom of
Information Act (FOIA). This section is revised to make clear that its
provisions also applies to NRC records and documents regarding standard
design approvals under part 52.
20. Subpart D--Additional Procedures Applicable to Proceedings for the
Issuance of Licenses To Construct and/or Operate for Nuclear Power
Plants of Identical Design at Multiple Sites
Formerly, subpart D of part 2 set forth the Commission's
administrative and hearing procedures for proceedings for issuance of
construction permits and operating licenses under part 52 for nuclear
power plants of ``duplicate'' design at multiple sites. The
requirements governing the content of such applications and the
technical consideration of such applications are set forth in 10 CFR
part 50, appendix N, which was ``transferred'' to part 52 as part of
the 1989 part 52 rulemaking. However, the 1989 rulemaking did not
remove appendix N from part 50, nor did the NRC make conforming changes
to appendix N in part 52 to make its provisions applicable to combined
licenses under subpart C of part 52. As discussed elsewhere, in the
March 2006 proposed rule the NRC proposed deleting appendix N in part
52, and retaining these provisions in part 50. Although no comment was
received on this proposal, the NRC has decided to withdraw its proposal
to delete appendix N in part 52. Instead, the NRC is revising appendix
N in part 52 to apply only to proceedings for combined licenses under
subpart C of part 52
[[Page 49418]]
(appendix N in part 50 will continue to address proceedings for
construction permits and operating licenses under that part).
To reflect the expanded scope of appendix N of part 52 and to
ensure that all of the NRC's regulations use consistent terminology,
the NRC is revising subpart D of part 2 as part of this final
rulemaking. Inasmuch as the changes to the provisions in subpart D
constitute revisions to the NRC's rules of procedure and practice, the
NRC may adopt them in final form without further notice and comment,
under the rulemaking provisions of the APA, 5 U.S.C. 553(b)(A).
21. Section 2.400, Scope of Subpart
This section is revised to refer to both appendix N of both part 50
and part 52, in order to reflect the Commission's determination that
the appendix should be retained in both parts, and that the procedures
in the appendices (both of which refer to this subpart) should apply to
applications for construction permits, operating reactors, and combined
licenses of identical design. In addition, Sec. 2.400 is revised to
use the term ``identical design,'' instead of the former ``essentially
the same design,'' so that subpart D and appendix N of part 50 and part
52 use identical terminology.
22. Section 2.401, Notice of Hearing on Construction Permit or Combined
License Applications Pursuant to Appendix N of 10 CFR Parts 50 or 52
Paragraph (a) of Sec. 2.401 is revised to indicate that notices of
hearing will be published for both construction permits under part 50
and combined licenses under part 52. Notices of the issuance of
operating licenses is addressed, as was the case under the former
provisions of subpart D, in Sec. 2.403. No other substantive changes
are intended by this revision. Paragraph (b) remains unchanged.
23. Section 2.402, Separate Hearings on Separate Issues; Consolidation
of Proceedings
Both paragraphs of this section are revised to refer to
applications under part 50 and part 52. No other substantive changes
are intended by this revision.
24. Section 2.403, Notice of Proposed Action on Applications for
Operating Licenses Pursuant to Appendix N of 10 CFR Part 50
This section is revised to refer to operating licenses issued under
part 50, rather than part 52. This reflects the Commission's
determination that appendix N of part 50 applies to construction
permits and operating licenses, whereas appendix N of part 52 applies
to combined licenses under subpart C of part 52.
25. Section 2.404, Hearings on Applications for Operating Licenses
Pursuant to Appendix N of 10 CFR Part 50
This section is revised to make clarifying changes by adding
references to a presiding officer, correctly referring to the Chief
Administrative Judge, and removing a reference to the atomic safety and
licensing board. No substantive changes are intended by this revision.
26. Section 2.405, Initial Decisions in Consolidated Hearings
This section is revised by requiring the presiding officer to issue
a separate partial initial decision on the common design. Section 2.405
is also revised by clarifying that the presiding officer may, if
otherwise determined under the consolidation provisions of Sec.
2.317(b), issue a consolidated decision for those proceedings. No other
substantive changes are intended by this revision.
27. Section 2.406, Finality of Decisions on Separate Issues
This section is revised to refer to both appendix N of both part 50
and part 52. No other substantive changes are intended by this
revision.
28. Section 2.407, Applicability of Other Sections
This section is revised to correctly reference subparts C, L, and N
of part 2. No other substantive changes are intended by this revision.
29. Section 2.500, Scope of Subpart
This section is revised by adding a conforming reference to subpart
F of part 52 on manufacturing licenses.
30. Section 2.501, Notice of Hearing on Application Under Subpart F of
Part 52 for a License To Manufacture Nuclear Power Reactors
This section is revised by adding a conforming reference to subpart
F of part 52 on manufacturing licenses. In addition, paragraph (b) of
this section is revised by removing the detailed requirements governing
the content of the notice of hearing published in the Federal Register,
and instead referencing proposed Sec. 2.104(f). As previously
discussed, the Commission is consolidating in Sec. 2.104 the
requirements governing the content of a notice of hearing with respect
to part 52 licensing and regulatory approval processes (with the
exception of standard design certifications, which are addressed in
subpart H of part 2).
31. Sections 2.502, 2.503, and 2.504
The text of these sections is removed, and their places are
reserved in the final rule, because the matters addressed in these
sections, regarding finality and the referencing of a manufactured
reactor in a combined license, are addressed with greater specificity
in the revisions to subpart F of part 52.
32. Subpart F, Additional Procedures Applicable to Early Partial
Decisions on Site Suitability Issues in Connection with an Application
for a Construction Permit or Combined License for Certain Utilization
Facilities
Subpart F provides special procedures for the acceptance,
docketing, administrative consideration, the conduct of hearings, and
the presiding officer's issuance of a partial initial decision in
licensing proceedings where there is early submittal of site
suitability information in connection with an application for a
construction permit or operating license, as described in Sec.
2.101(a-1). As discussed earlier, the NRC has revised Sec. 2.101(a-1)
to allow applicants for combined licenses under part 52, as well as
applicants for construction permits under part 50, to submit their
applications in two parts, and to allow for early consideration and
presiding officer's partial initial decision on those site suitability
matters for which the applicant seeks early resolution in accordance
with subpart F of part 2.
The NRC has reorganized subpart F in an attempt to improve its
usability (the reorganization is reflected in the provisions of Sec.
2.600, Scope of subpart). Requirements applicable to partial decisions
in construction permit proceedings continue to be addressed in
Sec. Sec. 2.602 through 2.606; a new subheading is added before Sec.
2.602 to reflect the subject matter of these sections. The new
requirements applicable to partial decisions in combined license
proceedings are in Sec. Sec. 2.621 through 2.629; a new subheading is
also added before Sec. 2.621 to reflect the subject matter covered by
these sections. Section 2.629, which has no analogous provisions in
Sec. Sec. 2.602 through 2.606, is added by the NRC to ensure that the
finality of a presiding officer's partial initial decision in a
combined license proceeding is clearly addressed using regulatory
language similar to that used in the finality provisions in part 52,
e.g., Sec. Sec. 52.39, 52.63, 52.98.
[[Page 49419]]
Section 2.601 is revised to correctly list subparts A, C, G, L, and
N of part 2 as subparts which are either applicable to or may be
utilized in proceedings under subpart F.
33. Section 2.800, Scope and Applicability
Subpart B of part 52 sets out the requirements applicable to
Commission issuance of regulations granting standard design
certification for nuclear power facilities. Standard design
certifications are approved through a rulemaking proceeding, and, in
concept, the applicant for a design certification may be considered as
a petitioner for rulemaking. However, subpart H of part 2, which sets
forth the Commission's procedures governing rulemaking, including
petitions for rulemaking, did not specifically address design
certification. Furthermore, based upon the Commission's experience with
three final design certification rules and a proposed design
certification rule, it is clear that some of the procedural
requirements applicable to petitions for rulemaking are not well-suited
to the administrative process for determining a design certification
application, e.g., the existing prohibition against pre-application
consultation with the NRC. These consultations between potential
license applicants and the NRC staff are not currently prohibited and
indeed are encouraged by the Commission to enhance NRC resource
planning and to facilitate early identification and resolution of
technical and regulatory issues. An application for design
certification is more like a license application than a traditional
petition for rulemaking, and the current prohibition against pre-
application consulting appears to be inconsistent with the Commission's
strategic objectives of safety, effectiveness, and management
excellence. The Commission also believes, based upon its experience,
that administrative provisions ordinarily applied in the context of
licensing (e.g., docketing and acceptance review, denial of application
for failure to supply information), should also be available for
application as appropriate in its determination of design certification
applications.
For these reasons, the Commission is revising subpart H of part 2
to address standard design certifications. Section 2.800 is revised to
delineate which provisions of subpart H are applicable to all petitions
for rulemaking, and which provisions are applicable only to initial
applications for design certification and applications for amendments
to existing design certification rules filed by the original applicant
(or successors in interest). The title of Sec. 2.800 is revised to
reflect the additional function of this section. New Sec. Sec. 2.811
through 2.819 are added to address initial applications for design
certification as well as applications for amendments to existing design
certifications filed by the original applicant (or successors in
interest), and are based upon Sec. Sec. 2.101, 2.107, and 2.109.
Petitions for amendment of existing design certification, which are
filed by third parties other than the original applicant for that
design certification (or successor in interest), will be treated as an
amending petition for rulemaking under the provisions of Sec. Sec.
2.801 through 2.810.
34. Section 2.801, Initiation of Rulemaking
In a conforming change, Sec. 2.801 is revised to refer to
applications for standard design certification rulemaking.
35. Section 2.811, Filing of Standard Design Certification Application;
Required Copies
New Sec. 2.811 clarifies the requirements that are related to the
filing of applications for standard design certifications. The
requirements in this section are derived from procedural requirements
for license applications located in several different regulations in
part 50. Section 2.811(a), which is analogous to Sec. 50.4(a),
identifies the NRC addresses where an application for a standard design
certification must be filed, and provides the requirements for
electronic submission of a design certification application. Section
2.811(b), which is analogous to Sec. 50.30(a)(1) and (3), provides
that a standard design certification application must meet the written
communications requirements in Sec. 2.813. Section 2.811(c), which is
analogous to Sec. 50.30(a)(2), requires the applicant to have the
capability to make and supply additional copies of the application upon
NRC request. Section 2.811(d), which is analogous to the requirement in
Sec. 50.30(a)(4), requires the applicant to make a copy of the updated
application for use by any party in a hearing conducted under subpart O
of part 2 (a legislative-style hearing). Section 2.811(e), which
addresses pre-application consultation with the NRC staff, provides
that the potential applicant for a design certification may consult
with the NRC on the subject matters listed in Sec. 2.802(a)(1)(i)
through (iii), including the procedure and process for filing and
processing an application for a design certification. However, Sec.
2.811(e) also allows the prospective standard design certification
applicant to consult with the NRC staff on substantive technical and
regulatory matters relevant to the design certification; the
prohibitions in Sec. 2.802(a)(2) do not apply to these consultations.
36. Section 2.813, Written Communications
New Sec. 2.813 contains procedural and ``housekeeping''
requirements governing written communications with the NRC, and are
derived from analogous requirements located in several different
regulations in part 50. Section 2.813(a) is analogous to Sec. 50.4(a).
Section 2.813(b) is analogous to Sec. 50.4(c), and sets forth the
requirement that written copies be submitted in permanent form on
unglazed paper. Section 2.813(c) is analogous to Sec. 50.4(d), and
expresses the Commission's preference that the upper right corner of
the first page of the applicant's submission set forth the specific
regulation or other basis which instigated the written communication.
37. Section 2.815, Docketing and Acceptance Review
New Sec. 2.815 is analogous to Sec. 2.101(a)(2), and permits the
NRC to conduct a review to determine whether the application is
complete (i.e., addresses all matters specifically required by NRC
regulation to be addressed in an application) and acceptable for
docketing. Section 2.815(a) provides that the NRC may determine, in its
discretion, the acceptability for docketing of an application based on
the technical adequacy of the application, not just on the completeness
of the application.
38. Section 2.817, Withdrawal of Application
New Sec. 2.817 is analogous to Sec. 2.107, and addresses the
procedures that the NRC will follow if a design certification applicant
withdraws its application. Section 2.817 also provides for a notice of
action on the withdrawal on the NRC Web site if the notice of
application was published on the NRC Web site.
39. Section 2.819, Denial of Application for Failure To Supply
Information
New Sec. 2.819 is analogous to Sec. 2.108, and states in
paragraph (a) that the NRC may deny an application for a standard
design certification if the applicant fails to respond to an NRC
request for additional information concerning its application within 30
days of the request. Section 2.819(b) provides that the NRC will
publish in the Federal Register a document denying the application.
Section 2.819(b) also states that the NRC will publish a notice on
[[Page 49420]]
the NRC's Web site denying the application if the NRC previously
published a notice of receipt of the application on the NRC Web site.
40. Section 2.1202, Authority and Role of NRC Staff
Paragraph (a) of Sec. 2.1202 acknowledges and confirms the
authority of the NRC staff to take regulatory (including licensing)
action during the pendency of a hearing, with several delineated
exceptions in numbered paragraphs (a)(1) through (5). Most of these
exceptions are mandated by Section 189.a.(1)(A) of the AEA, which
requires that the NRC hold a ``mandatory hearing,'' after 30 days
notice and publication once in the Federal Register, on any application
for a construction permit for a facility to be licensed under Section
103 or 104b. Paragraph (a)(1) is revised by adding specific references
to applications for limited work authorizations and combined licenses
under 10 CFR part 52. A limited work authorization is considered to be
a partial construction permit, and a combined license under part 52
includes a construction permit. Therefore, they are both subject to the
strictures of Section 189.a.(1)(A). Paragraphs (2), (3), and (4) are
redesignated as paragraphs (4), (5), and (6), and a new paragraph (2)
is added for early site permits applications. An early site permit is
considered to be a partial construction permit, and therefore is also
subject to Section 189.a(1)(A). A new paragraph (3) is added for
manufacturing licenses, as a matter of NRC discretion. The Section
189.a.(1)(A) requirement for a mandatory hearing applies only to
construction permits; a manufacturing license is not a construction
permit. Hence, the remaining provisions of Section 189.a.(1)(A),
including the NRC's authority to issue an operating license or
amendment to a construction permit without a hearing but only upon 30
days notice and publication once in the Federal Register of the NRC's
intent to do so, are inapposite and do not constrain the NRC's
authority to issue manufacturing licenses despite a pending hearing.
Nonetheless, as a matter of discretion, the NRC has decided to treat
manufacturing licenses similar to construction permits in this regard,
although the NRC reserves the right to change its practice in the
future.
G. Changes to 10 CFR Part 10
1. Section 10.1, Purpose; and Sec. 10.2, Scope
Part 10, which contains the NRC's requirements and procedures for
determining eligibility for granting access to Restricted Data and
National Security Information, did not reflect the licensing and
approval processes in part 52. Accordingly, the Commission made two
changes to ensure that there are defined criteria and procedures
governing requests for access to Restricted Data and National Security
Information by individuals with respect to a license or approval under
part 52.
Section 10.1 is revised by adding a new paragraph (a)(3), which
refers to the eligibility of individuals for employment with NRC
licensees and applicants, and holders of standard design approvals
under part 52. Section 10.2(b) is revised so that it refers to standard
design approvals under part 52 and applicants for consultants. This
change will address the provision of services associated with design
approvals, who may not, per se, be ``employees.''
H. Changes to 10 CFR Part 19
Part 19, entitled Notices, Instructions and Reports to Workers:
Inspection and Investigations, establishes the NRC's requirements for
notices, instructions and reports to persons participating in NRC
licensed and other regulated activities. For example, it requires
licensees and applicants for licenses to post a copy of, inter alia,
the regulations in 10 CFR parts 19 and 20, and NRC Form 3. NRC Form 3
provides a statement of rights and responsibilities to employees with
respect to NRC requirements. Part 19 also establishes the rights and
responsibilities of the NRC and individuals during interviews compelled
by subpoena as part of a NRC inspection or investigation under Section
161.c of the AEA. Finally, part 19 prohibits, on the grounds of sex,
the exclusion from participation in, or being subjected to
discrimination under any program or activity licensed by the NRC. The
regulatory authority for part 19 stems from Sections 211 and 401 of the
Energy Reorganization Act of 1974, as amended (1974 ERA).
The NRC has identified a number of weaknesses with the former
regulatory language in part 19. Formerly, part 19's regulatory
requirements and proscriptions applied only to licensees who receive,
possess, use or transfer material licensed under the NRC's regulations,
including persons licensed to operate a production or utilization
facility under 10 CFR part 50, but did not cover holders of 10 CFR part
52 licenses such as combined licenses, early site permits, and
manufacturing licenses. Moreover, part 19 applied only to licensees who
receive, possess, use or transfer materials licensed under 10 CFR parts
30 through 36, 39, 40, 60, 61, 63, 70 or 72 (including persons licensed
to operate a production or utilization facility under part 50). Thus,
the former regulations did not appear to address discrimination against
an employee during ``non-operational'' activities such as manufacturing
or construction of a nuclear power plant. Because the NRC's regulatory
scheme relies upon the proper design, manufacture, siting, and/or
construction of a production or utilization facility; discrimination
against an employee at any of these stages could have significant
adverse public health and safety or common defense and security
implications and effects. One would therefore expect that part 19 would
apply to such non-operational activities. Finally, part 19 applied only
to a ``licensee'' and activities authorized by a ``license'' (see,
e.g., Sec. Sec. 19.1, 19.2, 19.11, 19.20, 19.32), and did not extend
to part 52's non-licensing regulatory approvals, i.e., standard design
approvals and standard design certifications. Inasmuch as these non-
licensing activities regulated under part 52 are not different in kind
from the licensing which are currently subject to part 19 requirements,
the NRC concludes that they should also be subject to the requirements
in part 19.
Accordingly, the NRC is amending various provisions in part 19 to
ensure that its provisions extend to applicants for and holders of part
50 construction permits, and combined licenses, early site permits and
manufacturing licenses under part 52. In addition, the NRC extends part
19 to cover applicants for and holders of standard design approvals and
standard design certifications. The NRC believes that its regulatory
authority under Section 211 and Section 401 of the 1974 ERA is much
broader than the former scope of part 19. The anti-discrimination
proscriptions in Section 211 of the ERA apply to any ``employer,''
which the NRC regards as including non-licensee entities otherwise
regulated by the NRC, such as applicants for and holders of standard
design approvals, and applicants for standard design certifications.
The Commission believes that the use of the term, ``includes,'' in
paragraph (a)(2) of Section 211 of the 1974 ERA was not intended to be
an exclusive list of the persons and entities subject to the anti-
discrimination provisions in that section. The House Report on H.R.
776, which was adopted by Congress as the Energy Policy Act of 1992,
states:
[Title V] also broadens the coverage of existing whistle blower
protection provisions to include * * * any other employer engaged
[[Page 49421]]
in any activity under the Energy Reorganization Act of the Atomic
Energy Act of 1954. (H. Rep. No. 102-474, part 8, 102d Congress, 2d
Sess., at 78-79 (1992) (emphasis added))
There was no discussion of the statutory language in the conference
report. (H.R. Conf. Rep. No. 102-1018, 102d Cong., 2d Sess. (1992)).
The provisions in Section 401 of the ERA, prohibiting sex
discrimination apply to ``any program or activity carried on * * *
under any title of this Act.'' Accordingly, the NRC concludes that it
has the authority to extend the former scope of part 19 to address the
non-licensing regulatory approvals in part 52.
To implement the NRC's broadening of the scope of part 19,
Sec. Sec. 19.1 and 19.2 are revised to explicitly refer to: (1)
applicants for and holders of licenses and permits under part 52; (2)
applicants for and holders of final design approvals; and (3)
applicants for standard design certifications. The NRC notes that the
existing provision in Sec. 19.2 excluding part 19 from applying to NRC
employees and NRC contractors remains unchanged in the final rule. To
provide a convenient term for referring to persons and entities
applying for, or granting non-licensed regulatory approvals in part 52,
as well as any future regulatory processes, the NRC is amending Sec.
19.3 to the terms, regulated activities, and regulated entities.
Regulated entities are defined to include (but not be limited to)
applicants for and holders of standard design approvals under subpart E
of part 52, and applicants for standard design certifications under
subpart B of part 52.
Section 19.11 establishes requirements for posting of notices to
workers. Because Sec. Sec. 19.11(a)(2) and (a)(4) contain posting
requirements which are not relevant to early site permits,
manufacturing licenses, standard design approvals, and standard design
certifications, the NRC delineated in Sec. 19.11(b) the applicable
posting requirements for those regulatory processes. Section 19.11(c)
is reserved for future Commission use.
Sections 19.14 and 19.20 are revised to apply to regulated
entities, as well as licensees.
Section 19.31, governing exemptions from part 19, is revised to use
language consistent with Sec. 50.12 and Sec. 52.7. Unlike the former
regulation, which limits a request for exemption to a ``licensee,'' the
final rule allows ``interested persons,'' as well as licensees to
request an exemption from one or more provisions of part 19. This will
allow applicants for and holders of non-license regulatory vehicles in
part 52 (standard design approvals and design certifications) to
request exemptions from part 19. The broadened scope of persons that
will be allowed to request an exemption is consistent with most of the
exemption provisions throughout the NRC's regulations in Title 10 of
the CFR, including the specific exemption provision in part 50 (i.e.,
Sec. 50.12).
Section 19.32 is revised to more closely track the broad scope of
statutory language in Section 401 of the 1974 ERA, which is not limited
to licensing, but extends the sex discrimination prohibition to ``any *
* * activity carried on * * * under any title'' of the ERA. By using
the statutory language in the proposed rule, the NRC believes that the
regulations cover not only the existing non-license regulatory vehicles
in part 52, but any other regulatory approaches that the NRC may adopt
in the future (Section 401 of the 1974 ERA applies to NRC regulatory
activities under the AEA, inasmuch as the 1974 ERA transferred the AEA
regulatory authority from the old AEC to the NRC, see 1974 ERA, Sec.
104(c)).
I. Changes to 10 CFR Part 20
1. Section 20.1002, Scope
10 CFR part 20 applies to persons licensed by the NRC to receive,
possess, use, transfer, or dispose of byproduct, source, or special
nuclear material or to operate a production or utilization facility.
Accordingly, Sec. 20.1002 is revised by adding a conforming reference
to part 52, which sets forth a process for licensing a utilization
facility.
2. Section 20.1401, General Provisions and Scope
This section on decommissioning of facilities is revised to add a
conforming reference to facilities licensed under 10 CFR part 52.
3. Section 20.1406, Minimization of Contamination
Section 20.1406 requires applicants for licenses, other than
renewals, after August 20, 1997, to describe in the application how
facility design and procedures for operation will minimize, to the
extent practicable, contamination of the facility and the environment,
facilitate eventual decommissioning, and minimize, to the extent
practicable, the generation of radioactive waste. The NRC is adding
conforming changes to Sec. 20.1406 in the final rule. These conforming
changes to address part 52 were inadvertently overlooked in the
proposed rule. Section 20.1406 contains requirements that relate both
to design and operation of a facility and therefore applies in whole or
in part to design approvals, design certifications, manufacturing
licenses, and combined licenses. The final rule divides Sec. 20.1406
into two paragraphs. Paragraph (a) addresses applicants for licenses,
other than early site permits and manufacturing licenses, and contains
all of the requirements in former Sec. 20.1406. Paragraph (b)
addresses applicants for standard design certifications, standard
design approvals, and manufacturing licenses and only contains the
requirements related to design. If a combined license applicant
references a design approval, design certification, or a reactor
manufactured under a manufacturing license that has addressed the
design portion of this requirement under paragraph (b), then it would
only need to address the remaining ``operational'' requirements under
paragraph (a).
4. Section 20.2203, Reports of Exposures, Radiation Levels, and
Concentrations of Radioactive Material Exceeding the Constraints or
Limits
Sections 20.2203(c) and (d) are revised to add a reference to
holders of combined licenses to the procedures on submitting reports.
J. Changes to 10 CFR Part 21
Part 21 implements the reporting requirements in Section 206 of the
ERA. The proposed part 52 rule published in 2003 set forth the NRC's
proposals as to how Section 206 reporting and, therefore, part 21
applicability should be extended to early site permits, standard design
certifications, and combined licenses. However, the 2003 proposed rule
did not address Section 206 reporting requirements with respect to
standard design approvals or manufacturing licenses. Moreover, the
proposals were developed without the benefit of the NRC's in-depth
consideration of the issues as applied in the context of the early site
permit applications that are currently before the NRC. Accordingly, NRC
withdrew its earlier proposal and developed a more complete and
integrated rule on Section 206 reporting under part 21 and Sec.
50.55(e). As discussed previously, Sec. 50.55(e) sets forth the
Section 206 reporting requirements applicable to holders of
construction permits, combined licenses, and manufacturing licenses.
Key Principles of Reporting Under Section 206 of the ERA
The NRC believes that the extension of NRC's reporting requirements
[[Page 49422]]
implementing Section 206 of the ERA to part 52 licensing and approval
processes should be consistent with three key principles. First, NRC
regulatory requirements implementing Section 206 of the ERA should be a
legal obligation throughout the entire ``regulatory life'' of an NRC
license, a standard design approval, or standard design certification.
Second, reporting of defects or failures to comply associated with
substantial safety hazards should occur whenever the information on
potential defects would be most effective in ensuring the integrity and
adequacy of the NRC's regulatory activities under part 52 and the
activities of entities \10\ subject to the part 52 regulatory regime.
Third, each entity conducting activities within the scope of part 52
should develop and implement procedures and practices to ensure that it
fulfills its Section 206 of the ERA reporting obligation in an accurate
and timely manner.
---------------------------------------------------------------------------
\10\ Throughout this discussion, reference to entities,
licensees and/or applicants includes the contractors and
subcontractors of those entitles, licensees and/or applicants.
---------------------------------------------------------------------------
First Principle--Section 206 of the ERA Applies Throughout ``Regulatory
Life''
The first principle, that NRC regulatory requirements implementing
Section 206 must extend throughout the entire ``regulatory life'' of a
part 52 process, reflects the regulatory pattern inherent in part 52,
whereby certain designated licenses or approvals--e.g., an early site
permit, nuclear power reactor manufactured under a manufacturing
license, or a design certification--are capable of being referenced in
a subsequent nuclear power plant licensing application. Under the part
52 regulatory scheme, a referenced NRC approval constitutes the NRC's
basis for the licensing action within the scope of the prior Commission
approval, and becomes part of the ``licensing basis'' for that plant.
However, if Section 206 of the ERA reflects that effective NRC
decision-making and regulatory oversight require accurate and timely
information about defects and failures to comply associated with
substantial safety hazards, then Section 206 of the ERA should apply
whenever necessary to support effective NRC decision-making and
regulatory oversight of the referencing licenses and regulatory
approvals. To put it in different terms, if the NRC decision that it
may safely issue a license depends in part upon an earlier NRC safety
determination for a referenced license, standard design approval, or
standard design certification, it follows that a safety issue with
respect to the referenced license, design approval, or design
certification has safety implications for the referencing license or
design certification, and the continuing validity of the NRC's
licensing decision. Thus, the NRC concludes that the need for Section
206 reporting should not be limited to those licenses and approvals
under part 52 which are referenced or ``relied upon'' in a subsequent
nuclear power plant licensing application (viz., early site permits,
standard design approvals, standard design certifications, and
manufacturing licenses), but rather should extend to licenses and
approvals that are capable of being referenced in a future licensing
application. In other words, they must extend until there can be no
further potential safety implications for a referencing license or
approval.
The NRC believes that the beginning of the ``regulatory life'' of a
referenced license, standard design approval, or standard design
certification under part 52 occurs when an application for a license,
design approval, or design certification is docketed. Docketing of an
application marks the start of the NRC's formal safety and
environmental review of the application, and therefore the initiation
of the NRC's need for accurate and timely information to support its
regulatory review and approval. However, the NRC cautions that this
does not mean that an applicant is without Section 206 responsibilities
for pre-application activities. As the NRC staff discussed in a June
22, 2004, letter to the Nuclear Energy Institute (NEI) (ML040430041) in
the context of an early site permit, there are two aspects, namely, a
``backward looking'' or retrospective aspect with respect to existing
information, and a ``forward looking'' or prospective aspect with
respect to future information. The retrospective obligation is that the
early site permit holder and its contractors, must report all known
defects or failures to comply in ``basic components,'' as defined in
part 21. The prospective obligation is that the early site permit
holder and its contractors must report all defects or failures to
comply in basic components discovered subsequent to early site permit
issuance. The early site permit holder and its contractors are required
to meet these requirements, and must continue to meet them throughout
the term of the early site permit. Accordingly, safety-related design
and analysis or consulting services should be procured and controlled,
or dedicated, in a manner sufficient to allow the early site permit
holder and its contractors, as applicable, to comply with the above
described reporting requirements of Section 206, as implemented by 10
CFR 50.55(e) and part 21.
The NRC believes that the end of regulatory life occurs at the
later of: (1) The termination or expiration of the referenced license,
standard design approval, or standard design certification; or (2) the
termination or expiration of the last of the license or design
certification directly or indirectly referencing the (referenced)
license, design approval, or design certification. For example, if the
NRC approves a standard design approval, which is subsequently
referenced in a final standard design certification rule, and that
standard design certification is, in turn referenced in a combined
license issued by the NRC, the ``end'' of the regulatory life occurs
when the authorization to operate under the combined license is
terminated (ordinarily, under the provisions of Sec. 52.110). As long
as a referenced combined license continues to be effective, the
``regulatory life'' of a referenced license, standard design approval,
standard design certification, or manufactured reactor (as applicable)
must also continue and cannot be deemed to have ended.
Some commenters argued that the NRC's regulatory interests would be
met if reporting under Section 206 of the ERA were limited to the
referencing applicant/licensee, and that there should be no ongoing
part 21 reporting obligation imposed on the early site permit holder,
original applicant for a standard design certification, or holder of a
part 52 regulatory approval. Under this proposal the referencing
applicant and licensee would satisfy its obligation by an appropriate
contractual provision between the referencing applicant/licensee and
the entity ``supplying'' the referenced license or regulatory approval.
Although this could be a viable alternative for some combined licenses,
early site permits, and standard design approvals, the approach would
not be effective in the following contexts. This approach would not
result in reporting of defects to the NRC by the applicant of the early
site permit or standard design certification, which violates the NRC's
second principle (discussed more fully in the next section). In
addition, this approach would not result in reporting where there is no
contractual relationship between the combined license applicant/
licensee and the original applicant of the standard design
certification. Because the approach suggested by these commenters does
not
[[Page 49423]]
satisfy the NRC's regulatory objectives, it is not adopted.
One of the original applicants for the current standard design
certifications stated that any arguable Section 206 requirements must
logically end upon expiration of the standard design certification,
inasmuch as expiration marks the end time that the standard design
certification may be referenced. The NRC disagrees with this position.
Under Sec. 52.55(b) of the current regulations, a standard design
certification continues to be effective in a hearing for a combined
license or operating license docketed before the expiration date, and
in a hearing under Sec. 52.103 for authority to load fuel and operate.
At minimum, the original standard design certification applicant should
be subject to Section 206 requirements until the proceeding is
completed. Beyond the minimum requirements, the NRC also believes that
the original design certification applicant's Section 206 obligations
should continue until operation is no longer authorized in accordance
with Sec. 50.82(a)(2) for the last operating license or combined
license referencing that standard design certification. The NRC
believes that the regulatory need for information concerning defects in
a standard design certification continues throughout the operating life
of a license referencing that design certification; the relevance of
and the NRC's need for this information, if subsequently discovered by
the original design certification applicant, does not diminish simply
because the standard design certification may no longer be referenced.
Second Principle--Notification Occurs When Information Is Needed
The second principle is focused on ensuring that the NRC, its
licensees, and license applicants receive information on defects at the
time when the information would be most useful to the NRC in carrying
out its regulatory responsibilities under the AEA, and to the licensee
or applicant when engaging in activities regulated by the NRC. A result
of this principle is that reporting may be delayed if there is no
immediate consequence or regulatory interest in prompt reporting, and
that delayed reporting will actually occur when necessary to support
effective, efficient, and timely action by the NRC, its licensees and
applicants. Applying the second principle and its result to part 52
processes, the NRC believes that immediate reporting is required
throughout the period of pendency of an application, be it a license, a
standard design approval, or a standard design certification. Allowing
an applicant to delay the reporting of a defect would appear to be
inconsistent with the NRC's statutory mandate to provide adequate
protection to public health and safety and common defense and security.
Even if delayed reporting would allow the NRC an opportunity to modify
its prior safety finding with respect to the license, design approval,
or design certification, the delayed consideration is inconsistent with
one of the fundamental purposes of part 52, viz., to provide for early
consideration and resolution of issues in a manner that avoids the
potential for delay during licensing of a facility. Accordingly, the
Commission has determined that the NRC's requirements implementing
Section 206 of the ERA must extend to applicants (and their contractors
and subcontractors) for all part 52 processes (licenses, early site
permits, design approvals, and design certifications). Some commenters
stated that part 21 should not apply to applicants and claimed that the
NRC's proposal was contrary to the ERA. For the reasons stated
previously, the Commission does not agree with that position. However,
once an application has been granted, the Commission has decided that
immediate reporting of subsequently-discovered defects is not necessary
in certain circumstances. For those part 52 processes which do not
authorize continuing activities required to be licensed under the AEA,
but are intended solely to provide early identification and resolution
of issues in subsequent licensing or regulatory approvals, the
reporting of defects or failures to comply associated with substantial
safety hazards may be delayed until the time that the part 52 process
is first referenced. The Commission's view is based upon its
determination that a defect with respect to part 52 processes should
not be regarded as a ``substantial safety hazard,'' because the
possibility of a substantial safety hazard becomes a tangible
possibility necessitating NRC regulatory interest only when those part
52 processes are referenced in an application for a license, such as a
combined license or manufacturing license.
Some commenters believe that these reporting requirements should
not apply to a holder of an early site permit or a vendor of a standard
design until the ESP or standard design is referenced in a COL
application. As stated previously, the Commission agrees that reporting
may be delayed until the approval, certification, or permit is
referenced. After referencing, the holder (or in the case of a design
certification, the applicant who submitted the application leading to
the final design certification regulation) must make the necessary
notifications to the NRC as well as provide final engineering. The
notification must address the period from the Commission adoption of
the final design certification regulation up to the filing of the
application referencing the final design certification regulations.
Thereafter, notice must be made in the ordinary manner. The
notification obligation ends when the last license referencing the
design certification is terminated.
Third Principle--Procedures and Practices Must Be Implemented To Ensure
Accurate and Timely Reporting
The third principle (viz., each entity conducting activities under
the purview of part 52, should develop and implement procedures and
practices to ensure that the entity accurately and timely fulfils its
reporting obligation as delineated in the NRC's regulations), is
intended to ensure the effectiveness of each entity's reporting
processes. This is especially true where there is a potential for
substantial passage of time between the discovery of a defect and the
reporting of the defect, as may be allowed by the NRC consistent with
the second principle. For example, following issuance of a final
standard design certification regulation, if the original applicant
determines that there is a substantial safety hazard, that applicant
need not report the discovery until the time that the design
certification rule is referenced--which may be as long as 15 years from
the date of the final rule. Given the substantial time that may pass
between the time of discovery and the date of reporting, it is
imperative that the original standard design certification applicant
develop and implement procedures from the time of effectiveness of the
final design certification regulations.
The result of the third principle, consistent with part 21's
current requirements, is that licensees, license applicants, and other
entities seeking a design approval or design certification, must have
contractual provisions with their contractors, subcontractors,
consultants, and other suppliers which notify them that they are
subject to the NRC's regulatory requirements on reporting and the
development and implementation of reporting procedures. This result is
set forth in Sec. Sec. 21.31 and 50.55(e)(7).
[[Page 49424]]
Division of Implementing Requirements Between Part 21 and Sec.
50.55(e)
Under the Commission's current regulatory structure, persons and
entities engaged in construction (or the functional equivalent of
construction) are subject to reporting requirements under Sec.
50.55(e). Persons and entities engaged in all other activities within
the purview of Section 206 of the ERA are subject to the requirements
in part 21 and/or Sec. 50.55(e). The revised part 21 and Sec.
50.55(e) reflect the Commission's determination to retain this divided
regulatory structure. The NRC believes that the only part 52 processes
that authorize ``construction'' or its functional equivalent are
manufacturing licenses and combined licenses before the Commission
makes the finding under Sec. 52.103(g). Therefore, the reporting
requirements with respect to Section 206 of the ERA for manufacturing
licenses and combined licenses before the Commission makes the finding
under Sec. 52.103(g) are contained in Sec. 50.55(e). The requirements
in part 21 apply after the Commission makes the finding under Sec.
52.103(g) for a combined license. Part 21 was revised to explicitly
apply to the remaining part 52 processes, i.e., early site permits,
standard design approvals, and standard design certifications. Table A-
1 provides a summary of the applicability of part 21 and Sec. 50.55(e)
to each of the various approvals under part 52.
Table A-1.--Applicability of NRC Requirements Implementing Section 206 of the Energy Reorganization Act to Part
52 Licensing and Approval Processes
----------------------------------------------------------------------------------------------------------------
Applicable NRC requirement Sanctions
Part 52 licensing or approval processes implementing section 206 of the ---------------------------
ERA Civil Criminal
----------------------------------------------------------------------------------------------------------------
Early Site Permit (ESP)
Application................................. part 21........................... 21.61 21.62
Issuance of ESP............................. part 21........................... 21.61 21.62
Standard Design Approval (SDA)
Application................................. part 21........................... 21.61 21.62
Issuance of SDA............................. part 21........................... 21.61 21.62
Standard Design Certification Rule (DCR)
Application................................. part 21........................... 21.61 21.62
Final DCR Rulemaking........................ part 21........................... 21.61 21.62
Combined License (COL)
Application................................. 50.55(e).......................... 50.110 50.111
COL before Sec. 52.103 Authorization...... 50.55(e).......................... 50.110 50.111
COL after Sec. 52.103 Authorization....... part 21........................... 21.61 21.62
Manufacturing License (ML)
Application................................. 50.55(e).......................... 50.110 50.111
Issuance of ML.............................. 50.55(e).......................... 50.110 50.111
----------------------------------------------------------------------------------------------------------------
Reporting Requirements for Early Site Permits
If the ESP holder becomes aware of a significant safety concern
with respect to its site (e.g., that the specified site characteristics
for seismic acceleration is less than the projected acceleration due to
new information), the concern should be reported to the NRC so that it
may be considered in the review of any future application referencing
the ESP. As stated previously, the reporting may be delayed until the
ESP is referenced. This reporting attains special importance given the
NRC's proposal not to impose an updating requirement for ESP
information other than that related to emergency preparedness. In order
for the applicant for an ESP to have the capability to report to the
NRC any known significant safety concerns with respect to its site, or
any safety concerns of which it may subsequently become aware (i.e., to
be able to report any defects or failures to comply associated with
substantial safety hazards under part 21) the ESP applicant would have
to have a program in place for implementing the requirements of 10 CFR
part 21. The applicant's program may be inspected by the NRC as part of
the application review. Approval of the ESP application would be
subject to approval of the part 21 program.
Some commenters claimed that there is no practicable method for ESP
applicants or holders to determine whether an error in siting
information creates a substantial safety hazard and, therefore, part 21
should not be applicable to ESP applicants or holders. The Commission
does not agree with this position. As stated previously, the ESP holder
and its contractors can determine defects or failures to comply with
``basic components,'' as defined in part 21. This information is
necessary in order to support effective NRC decisionmaking and
regulatory oversight of the referencing licenses and approvals.
Applicability of Part 21 to Contractors or Subcontractors of an ESP
Applicant or Holder
In accordance with 10 CFR 21.31, the purchaser of a basic component
must state in the procurement documents for the basic component that
part 21 is applicable to that procurement. As explained previously,
services that are required to support an early site permit application
(e.g., geologic or seismic analyses, etc.) that are safety-related and
could be relied upon in the siting, design, and construction of a
nuclear power plant, are to be treated as basic components as defined
in part 21. Therefore, these services must be either purchased as basic
components, requiring the service provider to have an appendix B to
part 50 QA program, as well as its own part 21 program, or the early
site permit applicant could dedicate the service in accordance with
part 21, which requires the dedication process itself to be controlled
under an appendix B to part 50 QA program.
Reporting Requirements for Standard Design Approvals
A standard design approval represents the NRC staff's determination
regarding the acceptability of the design for a nuclear power reactor
(or major portions thereof). Although a standard design approval does
not represent the NRC's final determination as to the acceptability of
the design, it nonetheless represents a substantial expenditure of
agency resources in reviewing the design. A standard design
[[Page 49425]]
approval may be referenced in a subsequent application for a design
certification, construction permit, operating license, combined
license, or manufacturing license. Accordingly, consistent with the
first principle, the final rule imposes requirements implementing
Section 206 of the ERA on applicants for and holders of standard design
approvals.
A standard design approval does not authorize construction of a
nuclear power plant; it merely constitutes the NRC staff's approval of
the design of a nuclear power reactor (or major portion thereof).
Therefore, the requirements implementing Section 206 of the ERA, which
are applicable to standard design approvals, were placed in part 21, as
opposed to Sec. 50.55(e).
Reporting Requirements for Standard Design Certification Regulations
A standard design certification represents the NRC's approval by
rulemaking of an acceptable nuclear power reactor design, which may
then be referenced in a subsequent combined license or manufacturing
license application. Consistent with the first principle, the
Commission imposed Section 206 of the ERA reporting requirements on
applicants for design certifications, including applicants whose
designs are certified in a final design certification rulemaking. As
with a standard design approval, a design certification does not
actually authorize construction. Accordingly, the NRC revised
Sec. Sec. 21.2, 21.3, 21.21, 21.51, and 21.61 to explicitly refer to
an applicant for a standard design certification, rather than Sec.
50.55(e).
Some commenters have asserted that because there is no ``holder''
or licensee, the NRC is without authority under Section 206 of the ERA
to impose part 21 and/or Sec. 50.55(e) evaluation and reporting
requirements on applicants for standard design certification. The NRC
disagrees with this assertion. The statute by its terms does not limit
its reach to licensees; rather, the statute applies to any individual
or responsible officer of a firm ``constructing, owning, operating, or
supplying the components of any facility or activity which is licensed
or otherwise regulated * * *.'' The NRC believes that an applicant for
a standard design certification, by submitting its application, is
constructively ``supplying'' a ``component'' (the nuclear power plant)
for use in a future ``facility * * * licensed'' by the NRC. One of the
consequences of the design certification provisions in part 52 is the
ability of the applicant to subsequently offer its design with
additional, value-added services. Thus, applying for and facilitating
NRC adoption of a final standard design certification regulation is
simply a partial step in the overall activity of ``supplying'' the
certified design to potential nuclear power plant license applicants.
Alternatively, one could treat the standard design certification
applicant as supplying a component of an ``activity'' which is
``otherwise regulated'' by the NRC. Under this interpretation, the
``activity * * * otherwise regulated by the NRC'' can be viewed as the
design certification rulemaking, and/or the entire part 52 regulatory
regime whereby a design certification rule is referenced in a
subsequent licensing application. The NRC concludes that under either
interpretation, Section 206 of the ERA provides ample statutory
authority for the NRC to impose regulations implementing Section 206 on
design certification applicants, during the pendency of the application
before the NRC, as well as after NRC adoption of a final design
certification regulation (for those applicants whose application is
granted).
As with standard design approvals, a standard design certification
does not authorize construction of a nuclear power plant; it
constitutes the NRC's approval of the design of a nuclear power plant.
Therefore, the requirements implementing Section 206 of the ERA which
are applicable to design certifications were placed in part 21, as
opposed to Sec. 50.55(e).
Reporting Requirements for Combined Licenses
A combined license authorizes both construction of a nuclear power
plant, and loading of fuel and operation if the NRC makes the findings
specified in Sec. 52.103. As such, the application of the first and
second principles to combined licenses is the most straightforward of
all the part 52 processes. Under the final rule, the NRC's requirements
implementing Section 206 of the ERA would apply throughout the
regulatory life of the combined license, i.e., from docketing of the
application until termination of the combined license.
To maintain the current division between Sec. 50.55(e) and part 21
with respect to NRC requirements implementing Section 206 of the ERA,
the NRC revised Sec. 50.55(e) to make its provisions applicable to
each holder of a combined license under part 52 before the effective
date of the NRC's finding under Sec. 52.103(g), and to revise part 21
to clarify that its provisions apply to each holder of a combined
license on the effective date of the Commission's authorization under
Sec. 52.103(g).
Reporting Requirements for Manufacturing Licenses
Under subpart F of part 52, a manufacturing license would
constitute both the NRC's approval of a final nuclear power reactor
design, as well as approval to manufacture one or more reactors in
accordance with approved programs and procedures. The manufactured
reactors would then be transported offsite and incorporated into
nuclear power facilities by holders of combined licenses--who may be
different entities than the holder of a manufacturing license. Given
the possibility that the manufacturing license holder is different from
the combined license holder whose facility uses the manufactured
reactor, the NRC believes that the combined license holder must be kept
informed of any significant issue with design or manufacture of the
reactor, to ensure that they evaluate the significance of these matters
for their facility and undertake any necessary action to assure public
health and safety and common defense and security. Furthermore, unlike
a standard design certification, the financial resources necessary to
obtain a manufacturing license will, as a practical matter, result in
manufacturing beginning immediately after issuance of the manufacturing
license. There will be no interim period similar to a design
certification where there is no activity occurring under the
manufacturing license. Accordingly, in compliance with the first and
second principles, the NRC proposes that Section 206 of the ERA
requirements should apply continuously from the filing of the
application, until the manufacturing license expires or is otherwise
terminated by the NRC.
A manufacturing license holder would essentially be conducting the
same activities as a construction permit holder, albeit with several
differences.\11\ Nonetheless, the NRC believes that manufacturing is
similar to construction such that the NRC's requirements implementing
Section 206 of the ERA which are applicable to manufacturing licenses,
are contained in Sec. 50.55(e).
[[Page 49426]]
Accordingly, the NRC revised Sec. 50.55(e) to specifically apply its
provisions to holders of manufacturing licenses.
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\11\ These key differences are, first, the design of the
manufactured plant would be approved before manufacturing commences,
unlike the historical practice with construction permits. Second, a
single manufacturing license may authorize the manufacture of
multiple reactors, with the manufacturing process to be accomplished
in a controlled setting rather than as a ``field'' operation. This
is unlike the historical approach where non-standardized nuclear
power facilities were constructed onsite using a ``roving''
workforce. Third, the manufacturing license will specify the
inspections, tests, and acceptance criteria for determining
successful manufacturing.
---------------------------------------------------------------------------
K. Change to 10 CFR Part 25
1. Section 25.35, Classified Visits
Part 25 sets forth the NRC's requirements governing the granting of
access authorization to classified information to certain individuals.
Section 52.35, which requires that licensees and certificate holders
minimize the number of classified visits, did not, by its terms, apply
to applicants for standard design certifications, and applicants for or
holders of standard design approvals. Accordingly, Sec. 25.35 is
revised to refer to an applicant for a standard design certification
under part 52 (including the applicant after the NRC adopts a final
standard design certification rule), and the applicant for or holder of
a standard design approval under part 52.
L. Changes to 10 CFR Part 26
1. Section 26.2, Scope, Sec. 26.10, General Performance Objectives;
and Appendix A to Part 26
Part 26, which sets forth the NRC's requirements governing fitness-
for-duty, currently uses a two-part regulatory regime for the
application of fitness-for-duty requirements. A holder of an operating
license for a nuclear power plant is required to implement all of the
provisions in part 26. By contrast, a holder of a construction permit
is required to comply with Sec. Sec. 26.10, 26.20, 26.23, 26.70, and
26.73, and also implement a chemical testing program, including random
tests, and make provisions for employee assistance programs, imposition
of sanctions, appeals procedures, the protection of information, and
record keeping.
The NRC has extended the applicability of parts 26 to 52, in
keeping with the existing two-part regulatory regime, so that the full
array of requirements in part 26 apply to a combined license holder
after the date that the NRC authorizes makes the finding under Sec.
52.103(g), analogous to holder of an operating license under part 50.
By contrast, holders of combined licenses, before the date that the NRC
makes the Sec. 52.103(g) findings, are required to comply with the
part 26 provisions currently applicable to construction permit holders.
Similarly, holders of manufacturing licenses under subpart F of part 52
are treated the same as holders of construction permits. Finally,
persons authorized to conduct the limited construction activities
allowed under Sec. 50.10(e)(3) are also treated the same as a
construction permit holder. The final rule accomplishes this by: (1)
Revising Sec. 26.2(a) to refer to combined license holders after the
date that the NRC makes the finding under Sec. 52.103(g); (2) revising
Sec. 26.2(c) to refer to a holder of a combined license before the
date that the NRC makes the finding under Sec. 52.103(g), a holder of
a manufacturing license under subpart F of part 52, and a person
authorized to conduct the activities under Sec. 50.10(e)(3); (3)
revising Sec. 26.10(a) to refer to the personnel of a holder of a
manufacturing license and those authorized to conduct the activities
under Sec. 50.10(e)(3); and (4) revising appendix A to part 26,
paragraph 1.1(1) to include a reference to a holder of combined license
after the date that the NRC makes the finding under Sec. 52.103(g).
The NRC believes that part 26 need not be extended to cover
applicants for and holders of early site permits, standard design
approvals, and applicants for standard design certifications. These
activities present less of a concern with respect to public health and
safety, and common defense and security, as compared with construction
permits, manufacturing licenses, operating licenses, and combined
licenses. None of these regulatory approvals or design certification
regulations authorize the construction, manufacture, or operation of a
facility, nor do they authorize possession of special nuclear material
(SNM). The adverse impacts on public health and safety or common
defense and security attributable to any fitness-for-duty issues are
likely to be of a much lower level of significance, as compared to
issues that may occur during construction, manufacture, operation, or
possession of SNM. The NRC believes that the potential benefits of
imposing the fitness-for-duty requirements are not justified in view of
the regulatory burden to be imposed upon such applicants and holders.
Accordingly, these requirements will not be imposed on applicants for
and holders of standard design approvals and applicants for standard
design certifications under part 52.
M. Changes to 10 CFR Part 51
The NRC is making several conforming changes to part 51 to clarify
the environmental protection regulations applicable to the various part
52 licensing processes.
NEPA Compliance for Design Certifications
For each of the four design certification rules in appendices A, B,
C, and D of part 52, the NRC prepared an environmental assessment
which: (1) Provides the bases for a Commission finding of no
significant environmental impact (FONSI) for issuance of the design
certification regulation; and (2) identifies and addresses the need for
incorporating SAMDAs into the design certification rule. Based upon
this experience, the NRC is making changes to part 51 to accomplish two
objectives.
First, the NRC is eliminating the need for the NRC to prepare
essentially repetitive discussions in environmental assessments
supporting a FONSI on issuance of a final standard design certification
regulation. Each of the environmental assessments and FONSIs prepared
to date conclude that there is no significant environmental impact
associated with NRC issuance of a final design certification regulation
because a design certification does not authorize either the
construction or operation of a nuclear power facility. Design
certification represents the NRC's pre-approval of the design for the
nuclear power facility, but does not authorize manufacture or
construction. For the design certification to have practical effect, it
must be referenced in an application for a combined license. The NRC is
revising part 51 to eliminate the need for the NRC to make repetitive
findings of no significant environmental impact for future design
certifications and amendments to design certifications.
Second, the NRC is requiring that SAMDAs be addressed at the design
certification stage. SAMDAs are alternative design features for
preventing and mitigating severe accidents, which may be considered for
incorporation into the proposed design. The SAMDA analysis is that
element of the severe accident mitigation alternatives analysis dealing
with design and hardware issues. At the design certification stage, the
NRC's review is directed at determining if there are any cost
beneficial SAMDAs that should be incorporated into the design, and if
it is likely that future design changes would be identified and
determined to be cost-justified in the future based on cost/benefit
considerations. It is most cost effective to incorporate SAMDAs into
the design at the design certification stage. Retrofitting a SAMDA into
a design certification once site-specific design and engineering for a
nuclear power facility have been completed would increase the cost of
implementing a SAMDA. The retrofitting costs continue to increase in
ensuing stages of facility construction and operation. For these
reasons, the NRC believes that environmental
[[Page 49427]]
assessments for design certifications should address SAMDAs. However,
under the former provisions of part 51, both the environmental
information submitted by the design certification applicant, and the
environmental assessment prepared by the NRC, are directed either at
determining whether an EIS must be prepared, or that a FONSI is
justified. Accordingly, the NRC is requiring that SAMDAs be addressed
in environmental reports and environmental assessments for design
certifications.
The NRC is making a number of changes to accomplish these two
objectives. The NRC is redesignating existing Sec. 51.55 as Sec.
51.58, and is adding new Sec. 51.55 to indicate that an environmental
report submitted by the design certification applicant must be directed
towards addressing the costs and benefits of possible SAMDAs, and
presenting the bases for not incorporating identified SAMDAs into the
design to be certified. The environmental report for an applicant
seeking to amend an existing design certification would be somewhat
narrower by focusing on if the design change which is the subject of
the amendment, renders a SAMDA previously rejected to become cost-
beneficial, and if the design change results in the identification of
new SAMDAs that may be reasonably incorporated into the design
certification.
The NRC is revising Sec. 51.30 to provide for a new Sec. 51.30(d)
establishing the scope of an environmental assessment for a design
certification. The NRC is adding Sec. Sec. 51.32(b)(1) and (2) to set
forth the NRC's generic determination of no significant environmental
impact associated with issuance of a final or amended design
certification rule. This is, essentially, the legal equivalent of a
categorical exclusion. The NRC is including an explicit statement of no
significant environmental impact in Sec. 51.32. The NRC believes that
external stakeholders will better understand the nature of the
Commission's action by doing so. The NRC is modifying Sec. 51.31 by
adding Sec. 51.31(b) specifying the information on the environmental
assessment to be included in the proposed rulemaking on the design
certification published in the Federal Register.
The NRC is revising Sec. 51.50(c)(2) to indicate that if a
combined license application references a design certification then the
combined license applicant's environmental report may reference the
SAMDA discussion in the design certification environmental assessment
as part of its SAMDA analysis, but must contain information
demonstrating that the site characteristics for the combined license
site falls within the site parameters in the design certification
environmental assessment.\12\
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\12\ The design certification applicant may have chosen to
specify site parameters for the design certification safety review
under Sec. 52.79 which differ from the site parameters specified in
the environmental report for its design. If such a design
certification is referenced in a combined license application, the
combined license applicant must demonstrate that the two differing
sets of site parameters are met, in order for the full panoply of
issue finality provisions in Sec. 52.63 to apply in the combined
license proceeding.
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Finally, the NRC is adding Sec. 51.75(c)(2) to provide that if a
combined license application references a design certification, then
the combined license EIS will incorporate by reference the design
certification environmental assessment, and summarize the SAMDA
analysis and conclusions of the environmental assessment.
NEPA Compliance for Manufacturing Licenses
The NRC believes that its current approach for meeting the
Commission's NEPA responsibilities for standard design certifications
should be extended to manufacturing licenses for nuclear power
reactors. Under subpart F to part 52, a manufacturing license is
similar to a standard design certification in that a final nuclear
power reactor design would be approved. Therefore, the NRC is requiring
that the environmental effects of construction and operation of a
nuclear power facility using a manufactured reactor would be addressed
in the EIS for the combined license application for a nuclear power
facility using a manufactured reactor, rather than in an environmental
assessment or EIS at the manufacturing license stage.
Further, the NRC does not believe that NEPA requires the NRC to
address the environmental impacts of actually manufacturing a nuclear
power reactor licensed under subpart F of part 52, either at the
manufacturing license stage or at the combined license stage where an
application proposes to use a manufactured reactor. The manufacturing
license approves the final design of the manufactured reactor, the
organization and technical procedures for designing and manufacturing
the reactor, and the ITAAC that are to be used by the licensee in
determining whether the reactor has been properly manufactured in
accordance with NRC requirements and the manufacturing license, and the
possession (but not the use or transport offsite) of the manufactured
reactor. The manufacturing license does not approve any specific
location, building, or facility where the actual manufacture of the
reactors may occur,\13\ and the NRC does not require the applicant for
the manufacturing license to submit any information on these matters as
part of its application. These matters are commercial matters generally
unrelated to the NRC's regulatory jurisdiction. The Federal Aviation
Administration (FAA) does not prepare an EIS when issuing a production
certificate under 14 CFR part 21, subpart G, authorizing the production
of an aircraft or component in conformance with a type certificate. See
Federal Aviation Agency Order 1050.1E, Sec. 308c (June 8, 2004).
Because the NRC does not approve any specific location or facility in
which to manufacture any component of or the reactor licensed under the
manufacturing license, it would be speculative for the NRC to describe
and assess the environmental impacts of manufacturing. NEPA does not
require that an EIS address speculative impacts. The NRC also notes
that EISs prepared in the past for construction permits and operating
licenses under part 50, as well as current environmental assessments
for nuclear power plant license amendments, have never considered the
offsite environmental impacts of fabricating systems and components by
vendors and subcontractors, even for circumstances where the
fabrication activities are subject to NRC regulatory jurisdiction
(e.g., under applicable provisions of parts 19 and 21). For these
reasons, the NRC concludes that NEPA does not require the NRC to
address, either at the manufacturing license stage or at the combined
license stage where the application proposes to use a manufactured
reactor, the speculative impacts of manufacturing a reactor offsite at
a location or in a facility not specified or approved in the
manufacturing license.
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\13\ A reactor manufactured outside of the United States would
not be within the scope of a manufacturing license under subpart F
of part 52, by virtue of proposed Sec. 52.9, which states that no
license shall be deemed to have been issued for activities which are
not under or within the jurisdiction of the United States.
---------------------------------------------------------------------------
The NRC is making a number of changes to part 51, in some cases
parallel to those described previously with respect to design
certifications, consistent with its views on manufacturing licenses.
The NRC is revising existing Sec. 51.54 to clarify that an
environmental report for a manufacturing license must address the costs
and benefits of SAMDAs and the bases for not incorporating SAMDAs
[[Page 49428]]
into the design of the reactor to be manufactured, and to state that
the environmental report need not address the impacts of manufacturing
a reactor under the manufacturing license. The NRC is removing both
Sec. 51.20(b)(6), which formerly required preparation of an EIS for
issuance of a manufacturing license, and Sec. 51.76, which formerly
addressed the subject matter of an EIS for a manufacturing license,
from part 51.
The NRC is revising Sec. 51.30(e) to establish the scope of an
environmental assessment prepared for a manufacturing license. The NRC
is adding Sec. Sec. 51.32(b)(3) and (4) to state the NRC's generic
determination of no significant environmental impact associated with
issuance of a final or amended manufacturing license. As with the
parallel provisions governing design certifications in Sec.
50.32(b)(1) and (2), the NRC is including an explicit statement of no
significant environmental impact for manufacturing licenses in Sec.
51.32(b)(3) and (4) to facilitate external stakeholders' understanding
of the nature of the Commission's action. The NRC is adding Sec.
51.31(c) to describe the NRC's process for determining the
manufacturing license with respect to environmental issues covered by
NEPA.
The NRC is adding Sec. 51.50(c)(3) to provide that if a combined
license application proposes using a manufactured reactor, then the
combined license environmental report may incorporate by reference the
environmental assessment for the manufacturing license under which the
reactor is to be manufactured and, if so, must include information
demonstrating that the site characteristics for the combined license
site fall within the site parameters specified in the manufacturing
license environmental assessment. This section also states that the
environmental report need not address the environmental impacts
associated with manufacturing the reactor under the manufacturing
license.
Finally, the NRC is adding Sec. 51.75(c)(3) to indicate that if
the proposed combined license application to use a manufactured reactor
and the site characteristics of the combined license's site fall within
the site parameters specified in the manufacturing license
environmental assessment,\14\ then the combined license EIS must
incorporate by reference the manufacturing license environmental
assessment. As in the case where the combined license application
references a design certification, Sec. 51.75(c)(3) requires the
combined license EIS to summarize the findings and conclusions of the
environmental assessment with respect to SAMDAs. Finally, Sec.
51.75(c)(3) explicitly provides that the combined license EIS will not
address the environmental impacts of manufacturing the reactor under
the manufacturing license.
---------------------------------------------------------------------------
\14\ Analogous to design certifications, it is possible that an
applicant for a manufacturing license may have chosen to specify
site parameters for the manufacturing license safety review under
Sec. 52.79 which differ from the site parameters specified in the
environmental report for its design. If the combined license
application proposes to use such a manufactured reactor, then the
combined license applicant must demonstrate that the two differing
sets of site parameters are met, in order for the full division of
issue finality provisions in Sec. 52.171 to apply in the combined
license proceeding.
---------------------------------------------------------------------------
NEPA Obligations Associated With Sec. 52.103(g) Findings on ITAAC
Formerly, neither part 51 nor subpart C of part 52 explicitly
addressed whether an environmental finding under NEPA is needed in
connection with an NRC finding under Sec. 52.103(g) that combined
license ITAAC have been met. Nor does part 51 or subpart C of part 52
explicitly address whether contentions on environmental matters may be
admitted in a hearing under Sec. 52.103(b). The NRC never intended to
make an environmental finding in connection with the Sec. 52.103(g)
finding on ITAAC, and the NRC does not believe that NEPA requires such
a finding. The Sec. 52.103(g) finding that ITAAC have been met is not
a ``major Federal action significantly affecting the environment.'' The
major Federal action occurs when the NRC issues the combined license,
which includes the authority to operate the nuclear power plant--
subject to an NRC finding of successful completion of ITAAC. This is
the reason why the environmental impacts of operation under the
combined license are evaluated and considered by the NRC in determining
whether to issue the combined license even under the former provisions
of part 52, see Sec. 52.89. By contrast, the scope and nature of the
NRC finding that ITAAC have been met is constrained by the ITAAC itself
(indeed, the NRC has always recognized the possibility that ITAAC could
be written such that the ``inspections and tests'' exception in Section
554(a)(3) of the APA could be invoked to preclude the need to provide
an opportunity for hearing on Sec. 52.103(g) findings). The safety
consequences of operation are not considered when making the Sec.
52.103(g) findings; these issues are addressed by the NRC in
determining whether to issue the combined license in the first place.
Therefore, the NRC does not view the Sec. 52.103(g) finding as
constituting a ``major Federal action,'' and makes no environmental
findings in connection with that finding. It, therefore, follows that
no contentions on environmental matters should be admitted in any
hearing under Sec. 52.103(b).
Accordingly, the NRC is adding Sec. 51.108 to clarify that: (1)
The Commission will not make any environmental findings in connection
with the finding under Sec. 52.103(g); and (2) contentions on any
environmental matters, including the adequacy of the combined license
EIS and any referenced environmental assessment, may not be admitted
into any Sec. 52.103(b) hearing on compliance with ITAAC. Those issues
are essentially challenges to the continuing validity of the combined
license or any referenced design certification or manufacturing
license. Accordingly, these challenges should be raised with the
Commission using relevant Commission-established processes for
requesting Commission action. A challenge on environmental grounds with
respect to the combined license or manufacturing license must be filed
under the provisions of Sec. 2.206. A challenge to an existing design
certification on environmental grounds must be filed as a petition for
rulemaking to modify the existing design certification under subpart H
of part 2.
NEPA Compliance for Combined Licenses Referencing an Early Site Permit
The NRC has made several changes in the final rule based on public
comments regarding the requirements for a combined license application
referencing an early site permit and further consideration of the NRC's
obligations under NEPA for such actions. Several commenters believed
that an ESP and COL met the definition of ``connected actions,'' under
NEPA case law and Council on Environmental Quality (CEQ) regulations,
and should therefore not require the preparation of a new EIS for the
second of the two connected actions, or a revalidation of previous
findings if neither the applicant nor others identify new and
significant information. Commenters stated that under applicable NEPA
case law, there was no requirement to prepare a new EIS for the latter
of the two connected actions that were previously evaluated together in
a single EIS. The commenters stated that the EIS prepared at the ESP
stage serves as the EIS for issuance of both the ESP and COL.
Commenters stated that the ESP EIS included an evaluation of the
environmental impacts related to
[[Page 49429]]
issuance of a COL inasmuch as it considered the environmental impact of
plant construction and operation.
The NRC continues to believe that it is not necessary to require
that all topics be covered in a single EIS at the ESP stage, and that
topics such as alternative energy sources and need for power may be
treated in an EIS supplement at the COL application stage when the
detailed planning for the project is completed. As the commenters note,
new and significant information may also prompt the preparation of a
supplement to the ESP EIS in connection with the COL application. Since
the NRC believes that some issues may not be ripe for consideration at
the ESP stage, and an ESP EIS need not address such issues, the
Commission is declining to take a position on whether the granting of
an ESP and the granting of a COL referencing that ESP are connected
actions. Nevertheless, the Commission believes that, inasmuch as an
early site permit and a combined license are major Federal actions
significantly affecting the quality of the human environment, both
actions require the preparation of an EIS. However, 10 CFR part 52 does
provide finality for previously resolved issues. Under NEPA, the
combined license environmental review is informed by the EIS prepared
at the ESP stage and the NRC staff intends to incorporate by reference
the ESP EIS in the combined license supplemental EIS. A description of
what the combined license applicant must address in this situation can
be found under the discussion of changes to Sec. 51.50(c)(1).
More specific changes to individual sections in part 51 are
discussed as follows:
1. Section 51.20, Criteria for and Identification of Licensing and
Regulatory Actions Requiring Environmental Impact Statements
The NRC is revising Sec. 51.20(b) to identify the part 52
licensing processes that require an EIS or a supplement to an EIS.
Specifically, the NRC is revising Sec. 51.20(b)(1) to indicate that
issuance of an early site permit requires an EIS. The NRC is revising
Sec. 51.20(b)(2) to indicate that issuance of a combined license
requires an EIS. Also, paragraph (b)(6) is being removed and reserved
because, under the Commission's proposed revision to the requirements
for manufacturing licenses, only an environmental assessment is
required at this stage.
2. Section 51.22, Criterion for Categorical Exclusion; Identification
of Licensing and Regulatory Actions Eligible for Categorical Exclusion
or Otherwise Not Requiring Environmental Review
The NRC is revising Sec. 51.22(c) to identify part 52 licensing
processes that are eligible for categorical exclusion or otherwise do
not require environmental review.
3. Section 51.23, Temporary Storage of Spent Fuel After Cessation of
Reactor Operation--Generic Determination of No Significant
Environmental Impact
The NRC is revising Sec. Sec. 51.23(b) and(c) to indicate that the
provisions of these paragraphs also apply to combined licenses.
4. Section 51.26, Requirement To Publish Notice and Conduct Scoping
Process
The NRC is adding a new paragraph (d) to this section to provide
requirements for publication of a notice of intent when the NRC
determines that a supplement to an EIS will be prepared. This new
provision also states that, in such cases, the NRC staff need not
conduct a scoping process, provided, however, that if scoping is
conducted, then the scoping must be directed at matters to be addressed
in the supplement. If scoping is conducted in a proceeding for a
combined license referencing an ESP under part 52 , then the scoping
must be directed at matters to be addressed in the supplement as
described in Sec. 51.92(e).
5. Section 51.27, Notice of Intent
The NRC is adding a new paragraph (b) to this section to provide
requirements for the contents of a notice of intent when the NRC
determines that a supplement to an EIS will be prepared. Paragraph (b)
states that the notice of intent will, among other things, describe the
matters to be addressed in the supplement to the final EIS and describe
any proposed scoping process that the NRC staff may conduct.
6. Section 51.29, Scoping-Environmental Impact Statement and Supplement
to Environmental Impact Statement
The NRC is revising paragraph (a)(1) of this section in the final
rule to include requirements for supplements to an ESP EIS prepared for
a combined license application.
7. Section 51.45, Environmental Report
The NRC is revising Sec. 51.45(c) to indicate that the analysis in
an environmental report prepared for an ESP need not include
consideration of the economic, technical, and other benefits and costs
of the proposed action and of energy alternatives. This change is being
made for consistency with the provisions of Sec. 51.50(b), which state
that an environmental report included in an ESP application need not
include an assessment of the benefits (e.g., need for power) of the
proposed action and with the Commission's denial of a Petition for
Rulemaking (See PRM-52-02 (October 28, 2003; 68 FR 55905)).
8. Section 51.50, Environmental Report--Construction Permit, Early Site
Permit, or Combined License Stage
The NRC is revising the title of Sec. 51.50 to ``Environmental
Report Construction Permit, Early Site Permit, or Combined License
Stage,'' and including separate paragraphs with specific requirements
for environmental reports for early site permit and combined license
applications which are based on existing requirements in part 51 for
construction permits and operating licenses and requirements for early
site permits and combined licenses in part 52.
The NRC is revising the requirements from former Sec. 52.17(a)(2)
to clarify that an early site permit applicant has the flexibility of
either addressing the matter of alternative energy sources in the
environmental report supporting its early site permit application, or
deferring consideration of alternative energy sources to the time that
the early site permit is referenced in a licensing application. The NRC
believes the former regulations already afforded the early site permit
applicant such flexibility, inasmuch as former Sec. 52.17(a)(2) stated
that the environmental report submitted in support of an early site
permit application must ``focus on the environmental effects of
construction and operation of a reactor, or reactors * * *.'' The
environmental report's discussion of alternative energy sources does
not, per se, address the ``environmental effects of construction and
operation of a reactor,'' which is one of the matters which must be
addressed in an environmental impact statement (EIS). [See 10 CFR
51.71(d); National Environmental Policy Act of 1969 (NEPA), Sec.
102(2)(C)(i), (ii), and (v).] Rather, alternative energy sources
constitute part of the discussion of reasonable alternatives to the
proposed action, which is required by Section 102(2)(C)(iii) of NEPA.
[See 10 CFR 51.71(e) n.4; 46 FR 39440 (August 3, 1981) (proposed rule
that would eliminate consideration of need for
[[Page 49430]]
power and alternative energy sources at operating license stage), at
39441 (first column) (final rule published March 26, 1982; 47 FR
12940).] See Exelon Generation Company, LLC et al., CLI-05-17, 62 NRC
5, where the Commission ruled that:
[T]he ``reasonable alternatives'' issue does not apply with full
force to ESP (or ``partial'' construction permit) cases. At the ESP
stage of the construction permit process, the boards' ``reasonable
alternatives'' responsibilities are limited because the proceeding
is focused on an appropriate site, not the actual construction of a
reactor. Thus, boards must merely weigh and compare alternative
sites, not other types of alternatives (such as alterative energy
sources). (Id. at 48 (citations omitted).)
Accordingly, the NRC believes that former Sec. 52.17(a)(2) already
provided the early site permit applicant the flexibility of choosing to
defer consideration of alternative energy sources to the time that the
early site permit is referenced in a combined license or a construction
permit application. The revisions in Sec. 51.50(b) clarify that the
early site permit applicant may either include a discussion of
alternative energy sources in its environmental report, or defer
consideration of the matter. The NRC made conforming amendments
elsewhere in part 51 to clarify that the NRC's EIS need not address the
need for power or alternative energy sources (and therefore these
matters may not be litigated) if the early site permit applicant
chooses not to address these matters in its environmental report. The
environmental report and EIS for an early site permit must address the
benefits associated with issuance of the early site permit (e.g., early
resolution of siting issues, early resolution of issues on the
environmental impacts of construction and operation of a reactor(s)
that fall within the site characteristics, and ability of potential
nuclear power plant licensees to ``bank'' sites on which nuclear power
plants could be located without obtaining a full construction permit or
combined license). The benefits (and impacts) of issuing an early site
permit must always be addressed in the environmental report and EIS for
an early site permit, regardless of whether the early site permit
applicant chooses to defer consideration of the benefits associated
with the construction and operation of a nuclear power plant that may
be located at the early site permit site. This is because the
``benefits * * * of the proposed action'' for which the discussion may
be deferred are the benefits associated with the construction and
operation of a nuclear power plant that may be located at the early
site permit site; the benefits which may be deferred are entirely
separate from the benefits of issuing an early site permit. The
proposed action of issuing an early site permit is not the same as the
``proposed action'' of constructing and operating a nuclear power plant
for which the discussion of benefits (including need for power) may be
deferred under Sec. 51.50(b).
The NRC is further modifying Sec. 51.50(b) in the final rule based
on public comments. This section addresses requirements for
environmental reports at the early site permit stage. In the proposed
rule, Sec. 51.50(b) stated that environmental reports ``must focus on
the environmental effects of construction and operation of a reactor,
or reactors, which have characteristics that fall within the postulated
site parameters.'' Commenters pointed out that the use of ``postulated
site parameters'' was not consistent with the terminology the NRC had
used elsewhere in the proposed rule. Consequently, the NRC is revising
this provision in the final rule to require that the environmental
report ``must focus on the environmental effects of construction and
operation of a reactor, or reactors, which have design characteristics
that fall within the site characteristics and design parameters for the
early site permit application.'' A similar change is being made to the
same language in final rule Sec. 51.75(b) [proposed Sec. 51.71(d)].
The NRC is making additional changes to Sec. 51.50(b) to further
clarify the scope of the environmental review at the early site permit
stage. Final Sec. 51.50(b)(2) states that an early site permit
environmental report may address one or more of the environmental
effects of construction and operation of a reactor, or reactors, which
have design characteristics that fall within the site characteristics
and design parameters for the early site permit application, but that
the environmental report must address all environmental effects of
construction and operation necessary to determine whether there is any
obviously superior alternative to the site proposed. The purpose of
this change is to clearly delineate that the scope of the environmental
review at the early site permit stage is, at a minimum, to address all
issues needed for the NRC to perform its evaluation of the alternative
sites. In addition, the applicant may choose to address one or more
issues related to construction and operation of the facility with the
goal of achieving finality on those issues at the early site permit
stage.
In addition, the NRC is modifying Sec. Sec. 51.50(b) and 51.50(c)
in the final rule to reflect comments made at the NRC's public
workshops during the public comment period on the proposed rule. These
discussions related to the requirement to include a proposed list of
activities and a redress plan in license applications that request
authority to perform activities under Sec. 50.10(e). The NRC concluded
that it is preferable to include both the list of proposed activities
and the redress plan as separate documents in the application, outside
of both the final safety analysis report (or site safety analysis
report in the case of an early site permit) and the environmental
report. The NRC's conclusion is based on the fact that the requirements
in Sec. 50.10(e) address both safety and environmental issues.
Additional changes were made to Sec. Sec. 52.17(c), 52.79(a), and
52.80 to implement this concept.
The NRC is also revising Sec. 51.50(c) based on public comments in
the final rule. These revisions address the situation where a combined
license applicant is referencing an early site permit and provide for a
clearer link to the finality provisions in Sec. 52.39, eliminate
language that attempted to define ``new and significant,'' and provide
greater consistency with related requirements elsewhere in part 51. The
revisions also provide requirements for addressing environmental terms
and conditions. The discussion that follows reflects the language in
the final rule.
The NRC is adding a requirement in Sec. 51.50(c)(1) that the
applicant's environmental report need not contain information or
analyses submitted to the Commission in the early site permit
environmental report or resolved in the Commission's early site permit
environmental impact statement, but must contain, in addition to the
environmental information and analyses otherwise required: (1)
Information to demonstrate that the design of the facility falls within
the site characteristics and design parameters specified in the early
site permit; (2) information to resolve any significant environmental
issue that was not resolved in the early site permit proceeding; (3)
any new and significant information for issues related to the impacts
of construction and operation of the facility that were resolved in the
early site permit proceeding; (4) a description of the process used to
identify new and significant information regarding the NRC's
conclusions in the early site permit environmental impact statement,
including a requirement that the process use a reasonable
[[Page 49431]]
methodology for identifying such new and significant information; and
(5) a demonstration that all environmental terms and conditions that
have been included in the early site permit will be satisfied by the
date of issuance of the combined license. Any terms or conditions of
the early site permit that cannot be met by the time the combined
license is issued must be set forth as terms or conditions of the
combined license.
For an early site permit, the NRC prepares an EIS that resolves
numerous issues within certain bounding conditions. These issues have
issue preclusion at the combined license or CP stage provided certain
conditions are met. A combined license or CP application must
demonstrate that the design of the facility falls within the site
characteristics and design parameters specified in the early site
permit. In addition, the application must include any new and
significant information for issues related to the impacts of
construction and operation of the facility (i.e., the issue being
addressed at the combined license stage) that were resolved in the
early site permit proceeding. Documentation related to the applicant's
search for new information and its determination about the significance
of the new information should be maintained in an auditable form by the
applicant. The NRC staff may also use the environmental scoping process
to assist it in determining if there is new and significant information
regarding issues that were resolved in the early site permit
proceeding. Although the NRC is ultimately responsible for completing
any required NEPA review under 10 CFR 51.70(b), for example, an
evaluation of the impact of new and significant information on the
conclusions for a resolved early site permit environmental issue, the
combined license applicant must identify whether there is new and
significant information on such an issue. A combined license applicant
should have a reasonable process to ensure it becomes aware of new and
significant information that may have a bearing on the earlier NRC
conclusion, and should document the results of this process in an
auditable form. The NRC staff will verify that the applicant's process
for identifying new and significant information is effective.
The NRC, in the context of a combined license application that
references an early site permit, has defined the term ``new'' in the
phrase ``new and significant information'' as any information that was
both (1) not considered in preparing the ESP environmental report or
EIS (as may be evidenced by references in these documents, applicant
responses to NRC requests for additional information, comment letters,
etc.) and (2) not generally known or publicly available during the
preparation of the EIS (such as information in reports, studies, and
treatises). For new information to be ``significant,'' it must be
material to the issue being considered, that is, it must have the
potential to affect the finding or conclusions of the NRC staff's
evaluation of the issue. The COL applicant need only provide
information about a previously resolved environmental issue if it is
both new and significant.
The combined license applicant referencing an early site permit is
also required to provide information sufficient to resolve any other
significant environmental issue not considered in the early site permit
proceeding (e.g., need for power) and the information contained in the
application should be sufficient to aid the staff in its development of
an independent analysis (see 10 CFR 51.45).
Finally, the combined license applicant referencing an early site
permit must demonstrate that all environmental terms and conditions
included in the early site permit will be satisfied by the date of
issuance of the combined license. In some cases, this may require
adding a condition to the combined license to adequately address the
environmental issue raised in the early site permit condition. Note
that this provision was added to Sec. 51.50(c)(1) in the final rule.
Requirements to include environmental conditions in an early site
permit environmental report were addressed in the proposed rule in
Sec. 51.50(b), but the associated provision to ensure any conditions
included in the permit would be met was inadvertently left out of Sec.
51.50(c)(1).
In the past, the NRC staff has attempted to explain the
relationship between the environmental review of an early site permit
application to that of a combined license application referencing the
early site permit by analogy to the license renewal environmental
review process. The NRC believes the analogy especially useful because
the license renewal process is well-established and clearly understood.
Because there appears to be some confusion regarding this analogy, NRC
believes a brief explanation of the similarities of the two processes
is warranted.
For license renewal, the NRC prepared a generic EIS (GEIS) that
resolved more than 60 issues for all plants based on certain bounding
assumptions. These were termed Category 1 issues. If a license renewal
applicant identifies new and significant information with respect to a
Category 1 issue, it documents its assessment of that information in
its application. If the applicant determines that this new information
is not significant, or that there is no new information, the applicant
documents the bases for these determinations in an auditable form and
makes the documentation available for staff inspection. If there is new
and significant information on a Category 1 issue, the NRC staff limits
its inquiry to determine if this information changes the Commission's
earlier conclusion set forth in the GEIS. The NRC staff may inquire if
the applicant has a reasonable process for identifying new and
significant information on Category 1 issues.
Similarly, in the NRC environmental review process for a combined
license application, the combined license EIS brings forward the
Commission's earlier conclusions from the early site permit EIS and
articulates the activities undertaken by the NRC staff to ensure that
an issue that was resolved can remain resolved. If there is new and
significant information on a previously resolved issue, then the staff
will limit its inquiry to determine if the information changes the
Commission's earlier conclusion. Environmental matters subject to
litigation in a combined license proceeding mainly include (1) those
issues that were not considered in the previous proceeding on the site
or the design; (2) those issues for which there is new and significant
information; and (3) those issues subject to the change or exemption
processes in 10 CFR part 52.
Notwithstanding that, in the context of renewal, the GEIS resolves
Category 1 issues through rulemaking and an early site permit resolves
environmental issues through an individual licensing proceeding, the
staff believes that the license renewal practice is similar to the part
52 process in which a combined license application references an early
site permit.
The NRC has determined that a combined license is a major Federal
action significantly affecting the quality of the human environment
and, in accordance with 10 CFR 51.20, the NRC must prepare an EIS on
that action. If there is no new and significant information for matters
resolved at the ESP stage, then the staff will rely upon (``tier off'')
the ESP EIS at the combined license stage and disclose the NRC
conclusion for matters covered in the early site permit review. Such
matters
[[Page 49432]]
will not be subject to litigation at the combined license stage.
9. Section 51.51, Uranium Fuel Cycle Environmental Data--Table S-3
The NRC is revising Sec. 51.51 to require that every environmental
report prepared for the early site permit stage or combined license
stage of a light-water-cooled nuclear power reactor use Table S-3,
Table of Uranium Fuel Cycle Environmental Data, as the basis for
evaluating the contribution of the environmental effects of the uranium
fuel cycle to the environmental costs of licensing light-water-cooled
nuclear power reactors. If the application for a combined license
references an early site permit in which the environmental impacts and
costs related to the uranium fuel cycle were already evaluated and
resolved, then the repetition of this information in the environment
report for the combined license is not required unless the applicant
has identified new and significant information regarding these
environmental impacts and costs.
10. Section 51.52, Environmental Effects of Transportation of Fuel and
Waste--Table S-4
The NRC is revising Sec. 51.52 to require that every environmental
report prepared for the early site permit stage or combined license
stage of a light-water-cooled nuclear power reactor contain a statement
concerning transportation of fuel and radioactive wastes to and from
the reactor. If the application for a combined license references an
early site permit in which the transportation of fuel and radioactive
wastes to and from the reactor has already been evaluated and resolved,
then the repetition of this information in the environment report for
the combined license is not necessary unless the applicant has
identified new and significant information regarding the associated
environmental impacts.
11. Section 51.53, Postconstruction Environmental Reports
The NRC is revising Sec. 51.53(a) to clarify that any
postconstruction environmental report may incorporate by reference any
information contained in a prior environmental report or supplement
thereto that relates to the site or any information contained in a
final environmental document previously prepared by the NRC staff that
relates to the site. This change reflects the recognition that
environmental documents will be prepared at the early site permit stage
and may be referenced in environmental documents for future licensing
actions. The NRC is also revising Sec. 51.53(a) to clarify that
documents that may be referenced in post-construction environmental
reports include those prepared in connection with an early site permit
or a combined license. In addition, the NRC is revising Sec.
51.53(c)(3) to clarify that the requirements for the content of
environmental reports submitted in applications for renewal of a
combined license are the same as those for renewal of an operating
license.
12. Section 51.54, Environmental Report--Manufacturing License
The NRC is revising this section by adding two paragraphs to
delineate the difference in the matters with respect to SAMDAs that
must be addressed in an environmental report for issuance of a
manufacturing license under subpart F of part 52, versus that for an
amendment to the manufacturing license. Section 51.54(a) provides that
the environmental report for the manufacturing license must address the
costs and benefits of SAMDAs, and the bases for not incorporating into
the design of the manufactured reactor any SAMDAs identified during the
applicant's review. Section 51.54(b) reflects the narrower scope of an
environmental report submitted in connection with a proposed amendment
to a manufacturing license, by providing that the report need only
address whether the design change which is subject of a proposed
amendment either renders a SAMDA previously identified and rejected to
become cost beneficial, or results in the identification of new SAMDAs
that may be reasonably incorporated into the design of the manufactured
reactors.
As discussed earlier, the environmental impacts of manufacturing a
reactor under a manufacturing license are not considered by the NRC,
and Sec. 51.54 indicates that the environmental report need not
include a discussion of the environmental impacts of manufacturing a
reactor.
13. Section 51.55, Environmental Report--Standard Design Certification
The NRC is transferring the provisions in current Sec. 51.55 to a
new Sec. 51.58 (discussed in Sec. 51.58), and the NRC is revising
this section to address the contents of environmental reports for
design certifications under subpart B of part 52. The structure of new
Sec. 51.55 is similar to that of Sec. 51.54, reflecting the fact that
the environmental review for either manufacturing licenses or design
certifications is limited to SAMDAs. Section 51.55(a) provides that the
environmental report for the design certification must address the
costs and benefits of SAMDA, and the bases for not incorporating into
the design certification any SAMDAs identified during the applicant's
review. Section 51.55(b) provides that the environmental report
submitted in support of a request to amend a design certification need
only address whether the design change which is the subject of a
proposed amendment either renders a SAMDA previously identified and
rejected to become cost beneficial, or results in the identification of
new SAMDAs that may be reasonably incorporated into the design
certification.
14. Section 51.58, Environmental Report--Number of Copies; Distribution
The matters previously addressed in Sec. 51.55 are addressed in a
new Sec. 51.58. The NRC is adding conforming references to Sec.
51.58(a) for early site permits and combined licenses. Section 51.58(b)
contains a conforming reference to subpart F of part 52.
15. Section 51.71, Draft Environmental Impact Statement--Contents
The NRC is revising Sec. 51.71(d) to include a reference to Sec.
51.75 in the first sentence because Sec. 51.75 also includes
exceptions to the provisions in Sec. 51.71(d). This represents a
change the NRC is making in the final rule to move the specific
discussions on early site permits and combined licenses from Sec.
51.71(d) to their associated paragraphs in Sec. 51.75. The NRC is also
revising associated footnote 3 to include references to early site
permits and combined licenses.
16. Section 51.75, Draft Environmental Impact Statement--Construction
Permit, Early Site Permit, or Combined License
The NRC is adding Sec. Sec. 51.75(b) and (c) to include separate
requirements for the preparation of draft EISs at the early site permit
and combined license stages. In the final rule, the NRC is also moving
information related to early site permits that was contained in
proposed Sec. 51.71(d) to Sec. 51.75(b). In addition, the NRC is
providing further clarification in the final rule on the scope of the
environmental review at the early site permit stage. Section 51.75
requires that the draft environmental impact statement must include an
evaluation of alternative sites to determine whether there is any
obviously superior alternative to the site proposed. The draft
environmental impact statement must also include an evaluation of the
environmental effects of construction
[[Page 49433]]
and operation of a reactor, or reactors, which have design
characteristics that fall within the site characteristics and design
parameters for the early site permit application, but only to the
extent addressed in the early site permit environmental report or
otherwise necessary to determine whether there is any obviously
superior alternative to the site proposed. The purpose of this change
is to clearly delineate that the scope of the environmental review at
the early site permit stage is, at a minimum, to address all issues
needed for the NRC to perform its evaluation of the alternative sites.
In addition, the applicant may choose to address one or more issues
related to construction and operation of the facility with the goal of
achieving finality on those issues at the early site permit stage. The
NRC also notes that, where the early site permit application identifies
a specific nuclear power reactor design (i.e., a standard design
certification or manufacturing license) under Sec. 52.17(a)(1)(i), the
environmental report for an early site permit may address the
applicability of the severe accident mitigation design alternatives
(SAMDA) evaluation for that reactor design to the proposed site. In
this situation, the early site permit EIS must determine whether the
site characteristics bound the site parameters relevant to the SAMDA
analysis, as specified in the environmental assessment for the
identified nuclear power reactor design.
The requirements for combined licenses are organized into separate
paragraphs (c)(1), (c)(2), and (c)(3) which address the contents of the
combined license environmental impact statement if the combined license
application references an early site permit or standard design
certification, or proposes to use a manufactured reactor. For example,
Sec. 51.75(c)(3) provides that the combined license EIS will not
address the environmental impacts associated with manufacturing the
reactor under the manufacturing license.
In the final rule, Sec. 51.75(c)(1) states that if a combined
license application references an early site permit, then the NRC staff
shall prepare a supplement to the early site permit EIS. Paragraph
(c)(1) also requires that the supplement be prepared in accordance with
Sec. 51.92. Section 51.92 contains the requirements for the content of
a supplemental EIS prepared for a combined license application that
references an early site permit.
17. Section 51.92, Supplement to the Final Environmental Impact
Statement
The NRC is revising Sec. 51.92 in the final rule to provide
requirements for NRC staff preparation of a supplement to the final
environmental impact statement for an early site permit as required by
Sec. 51.75(c)(1). Paragraph (b) of Sec. 51.92 states that, in a
proceeding for a combined license application referencing an early site
permit, the NRC staff shall prepare a supplement to the final
environmental impact statement for the referenced early site permit in
accordance with Sec. 51.92(e). In the final rule, the NRC is moving
information related to combined licenses that was contained in proposed
Sec. 51.71(d) to Sec. 51.92(e) and is revising the wording of this
provision. In the proposed rule, Sec. 51.71(d) stated that the draft
supplemental environmental impact statement prepared at the combined
license stage when an early site permit is referenced need not include
detailed information or analyses that were resolved in the final
environmental impact statement prepared by the Commission in connection
with the early site permit, provided that the design of the facility
falls within the design parameters specified in the early site permit,
the site falls within the site characteristics specified within the
early site permit, and there is no new and significant environmental
issue or information not considered on the site or the design only to
the extent that they differ from that discussed in the final
environmental impact statement prepared by the Commission in connection
with the early site permit. In the final rule, the NRC has modified
these provisions and moved them to Sec. 51.92(e). The revised language
in paragraph (e) provides a clearer link to the finality provisions in
Sec. 52.39, eliminates language in the proposed rule that attempted to
define ``new and significant,'' and provides greater consistency with
related requirements elsewhere in part 51. Specifically, paragraph (e)
requires that a supplement to an early site permit final environmental
impact statement must: (1) Identify the proposed action as the issuance
of a combined license for the construction and operation of a nuclear
power plant as described in the combined license application at the
site described in the early site permit referenced in the combined
license application; (2) incorporate by reference the final
environmental impact statement prepared for the early site permit; (3)
contain no separate discussion of alternative sites; (4) include an
analysis of the economic, technical, and other benefits and costs of
the proposed action, to the extent that the final environmental impact
statement prepared for the early site permit did not include an
assessment of these benefits and costs; (5) include an analysis of
other energy alternatives, to the extent that the final environmental
impact statement prepared for the early site permit did not include an
assessment of energy alternatives; (6) include an analysis of any
environmental issue related to the impacts of construction or operation
of the facility that was not resolved in the proceeding on the early
site permit; and (7) include an analysis of the issues related to the
impacts of construction and operation of the facility that were
resolved in the early site permit proceeding for which new and
significant information has been identified, including, but not limited
to, new and significant information demonstrating that the design of
the facility falls outside the site characteristics and design
parameters specified in the early site permit.
18. Section 51.95, Postconstruction Environmental Impact Statements
The NRC is revising Sec. 51.95(a) to indicate that documents that
may be referenced in a supplement to a final environmental impact
statement include documents prepared in connection with an early site
permit or combined license. In addition, the NRC is revising Sec.
51.95(c) to add provisions for renewal of combined licenses and to
correct the address for the NRC Public Document Room. The NRC is
revising Sec. 51.95 to indicate that the NRC will prepare a
supplemental environmental impact statement in connection with the
amendment of a combined license authorizing decommissioning activities
or with the issuance, amendment, or renewal of a license to store spent
fuel at a nuclear power reactor after expiration of the combined
license, and that the supplement may incorporate by reference any
information contained in the final environmental impact statement for
the combined license or in the records of decision prepared in
accordance with an early site permit or combined license. Finally, the
NRC is revising Sec. 51.95(d) to indicate that, unless otherwise
required by the Commission, in accordance with the provisions of Sec.
51.23(b), a supplemental environmental impact statement for the post
combined license stage will address the environmental impacts of spent
fuel storage only for the term of the license, amendment, or renewal
applied for.
[[Page 49434]]
19. Section 51.105, Public Hearings in Proceedings for Issuance of
Construction Permits or Early Site Permits
The NRC is revising the section heading and Sec. 51.105(a) to
indicate that the requirements for presiding officers in public
hearings on construction permits also apply to public hearings on early
site permits. In addition, the NRC is adding Sec. 51.105(b) to
indicate that the presiding officer in an early site permit hearing
shall not admit contentions concerning the benefits assessment (e.g.,
need for power), or alternative energy sources if the applicant did not
address those issues in the early site permit application. This change
is being made for consistency with the provisions of Sec. 51.50(b),
which state that an environmental report included in an early site
permit application need not include an assessment of the benefits
(e.g., need for power) of the proposed action, and with the
Commission's denial of a Petition for Rulemaking (See PRM-52-02
(October 28, 2003; 68 FR 55905)). The NRC notes that the environmental
report and EIS for an early site permit must address the benefits
associated with issuance of the early site permit (e.g., early
resolution of siting issues, early resolution of issues on the
environmental impacts of construction and operation of a reactor(s)
that fall within the site characteristics, and ability of potential
nuclear power plant licensees to ``bank'' sites on which nuclear power
plants could be located without obtaining a full construction permit or
combined license). The benefits (and impacts) of issuing an early site
permit must always be addressed in the environmental report and EIS for
an early site permit, regardless of whether the early site permit
applicant chooses to defer consideration of the benefits associated
with the construction and operation of a nuclear power plant that may
be located at the early site permit site. This is because the
``benefits * * * of the proposed action'' for which the discussion may
be deferred are the benefits associated with the construction and
operation of a nuclear power plant that may be located at the early
site permit site; the benefits which may be deferred are entirely
separate from the benefits of issuing an early site permit. The
presiding officer needs to be mindful of whether the applicant has
addressed only the benefits of issuing the early site permit or whether
the applicant has also addressed all of the benefits of construction
and operation of the facility. This is because the presiding officer,
in accordance with Sec. 51.105(a)(3), must determine, after weighing
the environmental, economic, technical, and other benefits against
environmental and other costs, and considering reasonable alternatives,
whether the early site permit should be issued, denied, or
appropriately conditioned to protect environmental values. If the
applicant has addressed all of the costs and benefits associated with
construction and operation of the facility in its environmental report,
the final balancing between costs and benefits needs to occur at the
early site permit stage.
The NRC also notes that, where the early site permit application
identifies a specific nuclear power reactor design (i.e., a standard
design certification or manufacturing license) under Sec.
52.17(a)(1)(i), the environmental report for an early site permit may
address the applicability of the severe accident mitigation design
alternatives evaluation for that reactor design to the proposed site.
In this situation, the early site permit EIS must determine whether the
site characteristics bound the site parameters relevant to the SAMDA
analysis, as specified in the environmental assessment for the
identified nuclear power reactor design. In addition, in accordance
with Section 52.107(c), the presiding officer shall not admit
contentions proffered by any party concerning severe accident
mitigation design alternatives unless the contention demonstrates that
the site characteristics fall outside of the site parameters in the
standard design certification or underlying manufacturing license for
the manufactured reactor.
20. Section 51.105a, Public Hearings in Proceedings for Issuance of
Manufacturing Licenses
The NRC is adding Sec. 51.105a to provide requirements for public
hearings in proceedings for issuance of manufacturing licenses.
Specifically, Sec. 51.105a establishes that the presiding officer in a
proceeding for a manufacturing license will determine whether the
manufacturing license should be issued as proposed by the appropriate
NRC staff director.
21. Section 51.107, Public Hearings in Proceedings for Issuance of
Combined Licenses
The NRC is adding Sec. 51.107 to set out the requirements for
public hearings in proceedings for issuance of combined licenses. The
requirements parallel the associated requirements for public hearings
on construction permits and operating licenses, as appropriate, and
provide requirements unique to the combined license process that are
derived from various provisions in part 52, namely Sec. Sec. 52.39 and
52.103. The NRC is making changes to the language in Sec. 51.107 in
the final rule to more clearly define the role of the presiding officer
in a proceeding for the issuance of a combined license where an early
site permit is being referenced. Specifically, paragraph (b) addresses
the situation where a combined license application references an early
site permit and a supplement to the early site permit environmental
impact statement is prepared in accordance with Sec. 51.75(c)(1) and
Sec. 51.92(e). In such cases, the presiding officer in the combined
license hearing shall not admit any contention proffered by any party
on environmental issues which have been accorded finality under Sec.
52.39 unless the contention: (1) Demonstrates that the nuclear power
reactor proposed to be built does not fit within one or more of the
site characteristics or design parameters included in the early site
permit; (2) raises any significant environmental issue that was not
resolved in the early site permit proceeding; or (3) raises any issue
involving the impacts of construction and operation of the facility
that was resolved in the early site permit proceeding for which new and
significant information has been identified.
N. Changes to 10 CFR Part 54
1. Section 54.1, Purpose
This part applies to renewed operating licenses for nuclear power
plants. A conforming change is made to this section to include renewed
combined licenses.
2. Section 54.3, Definitions
The definition for renewed combined license is added to explain the
meaning of the new phrase as it is used in this part.
3. Section 54.17, Filing of Application
Section 54.17(c) is revised to add a conforming reference to
combined licenses issued under 10 CFR part 52.
4. Section 54.27, Hearings
This section is revised to include a conforming reference to
renewed combined license issued under 10 CFR part 52.
5. Section 54.31, Issuance of a Renewed License
Sections 54.31(a), (b), and (c) are revised to include conforming
references to combined licenses in this procedure on issuance of
renewed licenses.
[[Page 49435]]
6. Section 54.35, Requirements During Term of Renewed License
This section is revised to include conforming references to holders
of combined licenses and the regulations in part 52 into the
requirements for a renewed license.
7. Section 54.37, Additional Records and Recordkeeping Requirements
Section 54.37(a) is revised to include a conforming reference to a
renewed combined license.
O. Changes to 10 CFR Part 55
Part 55 establishes the NRC's requirements for licensing of
operators of utilization facilities in accordance with the statutory
requirements in Section 202 of the ERA. Formerly, the provisions in
part 55 referred only to utilization facilities licensed under part 50,
and therefore, do not address utilization facilities licensed for
operation under a combined license issued under subpart C of part 52.
Section 202 of the ERA, however, does not limit its mandate to
operators of facilities licensed under part 50; the statutory
requirement would also appear to apply to operators of facilities
licensed under part 52 (i.e., combined licenses under subpart C of part
52).
Accordingly, Sec. Sec. 55.1 and 55.2 are revised by adding a
reference to part 52. This clarifies that each operator of a nuclear
power reactor licensed under a part 52 combined license or renewed
under part 54 must first obtain an operator's license under part 55. In
addition, the conforming changes clarify that these operators, as well
as holders of combined licenses issued under part 52 or renewed under
part 54, are subject to the requirements in part 55 (e.g., part E of
part 55, Written Examinations and Operating Tests, set forth
requirements which are directed, for the most part, at the holders of
operating licenses for utilization facilities).
P. Changes to 10 CFR Part 72
1. Section 72.210, General License Issued
Part 72 sets forth the requirements for independent spent fuel
storage facilities. This section is revised to include a conforming
reference to persons authorized to operate nuclear power reactors under
10 CFR part 52 (i.e., a combined license holder).
2. Section 72.218, Termination of Licenses
Section 72.218(b) is revised to include a conforming reference to
combined licenses issued under part 52.
Q. Changes to 10 CFR Part 73
Part 73 establishes the NRC's requirements for the physical
protection of production and utilization facilities licensed by the
NRC. It provides requirements for the physical protection of licensed
activities, for personnel access authorization, and for criminal
history checks of individuals granted unescorted access to a nuclear
power facility or access to Safeguards Information. Formerly, the
language of Sec. 73.1, Purpose and scope, Sec. 73.2, Definitions,
Sec. 73.50, Requirements for physical protection of licensed
activities, Sec. 73.56, Personnel access authorization requirements
for nuclear power plants, and Sec. 73.57, Requirements for criminal
history checks of individuals granted unescorted access to a nuclear
power facility or access to Safeguards Information by power reactor
licensees, and Appendix C, Licensee Safeguards Contingency Plans, did
not refer to combined licenses issued under part 52. However, part 73
was formerly applicable to combined licenses under the provisions of
Sec. 52.83, Applicability of part 50 provisions, which states that all
provisions of 10 CFR part 50 and its appendices applicable to holders
of operating licenses also apply to holders of combined licenses.
Accordingly, Sec. 73.1 is revised to clarify that the regulations in
part 73 apply to persons who receive combined licenses under part 52,
and Sec. 73.2 is revised to state that terms defined in part 52 have
the same meaning when used in part 73. The NRC has addressed combined
licenses in Sec. 73.57 by making the provisions that are required
before receiving an operating license under part 50 applicable before
the date that the Commission makes the finding under Sec. 52.103 for a
combined license. Additional conforming changes to include part 52
licenses are made for Sec. Sec. 73.50 and 73.56, and appendix C to
part 73.
R. Change to 10 CFR Part 75
1. Section 75.6, Maintenance of Records and Delivery of Information,
Reports, and Other Communications
Part 75 sets forth NRC requirements intended to implement the
agreement between the United States and the International Atomic Energy
Agency (IAEA) with respect to safeguards of nuclear material. Various
provisions throughout part 75 require certain licensees and other
individuals and entities regulated by the NRC to submit to the NRC
various reports and communications. Section 75.6 specifies the NRC
officials to whom these reports and communications are to be sent.
However, Sec. 75.6(b)--the provision applying to, inter alia, nuclear
power plants--refers only to holders of a construction permit or an
operating license, and does not include holders of combined licenses.
Accordingly, Sec. 75.6(b) is revised to reference combined licenses.
The NRC notes that early site permits and manufacturing licenses need
not be referenced, inasmuch as the U.S.-IAEA Safeguards Agreement does
not extend to early site permits or manufacturing licenses.
S. Changes to 10 CFR Part 95
The following discussion explains the requirements in part 95
generically and covers Sec. Sec. 95.5, 95.13, 95.19, 95.20, 95.23,
95.31, 95.33 through 95.37, 95.39, 95.43, 95.45, 95.49, 95.51, 95.53,
95.57, and 95.59.
Part 95 sets forth the NRC requirements governing what individuals
and entities may be provided access to National Security Information
(NSI) and/or Restricted Data (RD) received or developed in connection
with activities licensed, certified, or regulated by the NRC, and how
this information and data is to be protected by these individuals and
entities against unauthorized disclosure.
Although requirements for protection of NSI and RD must, by
statute, apply to all individuals and entities provided access to such
information, various sections in part 95 use slightly different wording
to delineate the relevant set of individuals and entities. To ensure
consistency, the Commission is revising its regulations to refer to
``licensee, certificate holder, or other person,'' to describe the
individuals and entities subject to the applicable requirements. In
adopting this phrase, the NRC intends to ensure that its regulatory
requirements for protection of NSI and RD in part 95 extend as broadly
as the NRC's authority provided under applicable law. The term,
``licensee,'' includes both holders of all NRC licenses, including (but
not limited to) combined licenses, as well as holders of permits such
as construction permits and early site permits. The term, ``certificate
holder,'' includes (but is not limited to) all certificates of approval
that the Commission may issue, such as a certificate of compliance for
spent fuel casks under 10 CFR part 72. Finally, the term, ``or other
person,'' is intended to include individuals and entities who are
subject to the regulatory authority of the Commission, including
applicants for standard design approvals and standard design
certifications under part 52. For the same reasons, the Commission is
revising Sec. 95.39 to use the phrase, ``NRC
[[Page 49436]]
license, certificate, or standard design approval or standard design
certification under part 52.''
T. Changes to 10 CFR Part 140
Part 140 addresses the NRC requirements applicable to nuclear
reactor licensees with respect to financial protection and indemnity
agreements to implement Section 170 of the AEA, commonly referred to as
the Price-Anderson Act. In general, the indemnification and financial
protection requirements in part 140 become applicable when a holder of
a 10 CFR part 50 construction permit who also possesses a materials
license under 10 CFR part 70 brings fuel onto the site. However, part
140 did not address the indemnification and financial protection
requirements of combined license holders. Accordingly, the final rule
revises various sections in part 140 to address combined licenses under
part 52.
The NRC does not believe that part 140 must be revised to address
any part 52 licensing process other than a combined license. Neither an
early site permit nor a manufacturing license authorizes the possession
or use of nuclear fuel or other nuclear materials, and the NRC would
not issue these licenses with a materials license under part 70. The
NRC also believes that part 140 need not be revised to address standard
design approvals or standard design certifications, because neither of
these processes authorize the possession or use of nuclear fuel or
other nuclear materials.
U. Changes to 10 CFR Part 170
Part 170 sets out the fees charged for licensing services performed
by the NRC. The NRC is revising Sec. 170.2(g) and (k) to add
conforming references to manufacturing licenses and standard design
approvals issued under part 52, revise the existing reference to
appendix Q to part 52 to be a reference to appendix Q to part 50, and
delete the reference to a manufacturing license issued under part 50
(which is being removed from part 50 because of its transfer to part 52
in the 1989 rulemaking adopting part 52).
V. Changes to 10 CFR Part 171
Part 171 sets out the annual fees charged to persons who hold
licenses issued by the NRC. The NRC is revising Sec. 171.15 to add
conforming references to combined licenses issued under part 52. Note
that for combined licenses, the requirements of part 171 are not
applicable until after the Commission has made the finding under Sec.
52.103(g). This section also provides fee requirements for each person
holding a part 50 power reactor license that is in decommissioning or
possession only status and each person holding a part 72 license who
does not hold a part 50 license. The NRC also added conforming changes
to include references in part 52 in these provisions.
VI. Section-by-Section Analysis
Part 52, General Provisions
Section 52.0 Scope; Applicability of 10 CFR Chapter I Provisions
This section, formerly designated as Sec. 52.1, has been expanded
to: (1) address all licensing and regulatory processes covered in part
52; and (2) more clearly define the relationship between part 52 and
remaining provisions of 10 CFR Chapter I. Paragraph (a), which
establishes the scope of part 52, is revised by referring to all
licensing and regulatory processes covered in part 52. In addition,
paragraph (a) is revised to give notice to contractors, subcontractors
or consultants of applicants for or holders of licenses or regulatory
approvals under part 52 that they are subject to NRC enforcement action
for violations of the deliberate misconduct proscriptions in Sec.
52.4. The Commission notes, as discussed below in the section-by-
section analysis of Sec. 52.4, that deliberate misconduct under Sec.
52.4 may occur as the result of a violation of any Commission rule and
regulation throughout 10 CFR Chapter I, not just a violation of a
requirement in part 52.
Paragraph (b) is a new provision that supersedes former Sec.
52.83. The first sentence of paragraph (b) is intended to make clear
that the Commission's regulations in 10 CFR Chapter I apply to
applicants and holders of licenses, permits and other regulatory
approvals in part 52 (e.g., design approvals and standard design
certifications). Accordingly, applicants, licensees and holders of
regulatory approvals under part 52 should review the regulations in 10
CFR Chapter I to ensure that they are in compliance with applicable
Commission requirements throughout 10 CFR Chapter I. The second
sentence of paragraph (b) reinforces the applicability of the
Commission's requirements throughout 10 CFR Chapter I to part 52
licenses, permits, and other regulatory approvals. As part of this
final rule, the Commission is making conforming changes as necessary
throughout Chapter I to ensure that relevant regulations clearly set
forth their applicability to part 52 licenses and approvals, and to
part 52 entities such as applicants, licensees, and holders.
Nonetheless, the Commission is adopting paragraph (b) in order to
clearly and unambiguously impose applicable regulatory requirements
that exist throughout 10 CFR Chapter I.
Section 52.1 Definitions
This section, formerly designated as Sec. 52.2, has been
supplemented by: (1) adding definitions of terms that are used in part
52 but were undefined in the previous rule; and (2) providing
definitions of new terms that were added in this rulemaking to provide
greater clarity and precision. New definitions which are noteworthy are
discussed individually as follows.
A definition of modular design is added to explain the type of
modular reactor design to which the Commission intended to refer to in
the second sentence of the current Sec. 52.103(g). This special
provision for modular designs was added to part 52 to facilitate the
licensing of nuclear plants, such as the Modular High Temperature Gas-
Cooled Reactor (MHTGR) and Power Reactor Innovative Small Module
(PRISM) designs, that consisted of three or four nuclear reactors in a
single power block with a shared power conversion system. During the
period that the power block is under construction, the Commission could
separately authorize operation for each nuclear reactor when each
reactor and all of its necessary support systems were completed. The
Commission believes that the term ``modular design'' needs to be
defined to aid future use of the current Sec. 52.103(g) by
distinguishing the intended definition from other currently used
definitions for ``modular design.'' Also, future combined license
applicants for a multi-unit site that would be similar to current
multi-unit sites (where each unit is similar in design but independent
of all other units) could use this provision.
Definitions of the terms design characteristics, design parameters,
site characteristics, and site parameters were added to Sec. 52.1 to
clarify their meaning and use in the licensing and approval processes
of part 52. Design characteristics are defined as the actual features
of a nuclear reactor or reactors. Design characteristics are specified
in the final safety analysis report for a standard design approval, a
standard design certification, a combined license application, or a
manufacturing license. Design parameters are defined as the postulated
features of a nuclear reactor or reactors that could be built at the
proposed site. Design parameters are specified in an early site permit
application. Site characteristics are defined as the actual physical,
environmental, and demographic
[[Page 49437]]
features of a site. Site characteristics are specified in an early site
permit or combined license application. Site parameters are defined as
the postulated physical, environmental, and demographic features of an
assumed site. Site parameters are specified in a standard design
approval, standard design certification, or manufacturing license.
The values for the characteristics and parameters will be used in
the NRC's review of combined license applications that reference design
approvals, design certifications, manufacturing licenses, or early site
permits. For example, Sec. 52.79(b) requires that a combined license
application referencing an early site permit contain information
sufficient to demonstrate that the actual design characteristics of the
nuclear facility fall within the design parameters and site
characteristics specified in the early site permit. Also, Sec.
52.79(d) requires that a combined license application referencing a
design certification rule must contain information sufficient to
demonstrate that the actual site characteristics fall within the site
parameters specified in the design certification.
The above terms are also used in Sec. Sec. 52.39 and 52.93.
Because the NRC is relying on certain design parameters specified in
the early site permit applications to reach its conclusions on site
suitability, these design parameters will be included in any early site
permit issued. The NRC believes that its review of a combined license
application that references an early site permit will involve a
comparison to ensure that the actual characteristics of the design
chosen by the combined license applicant fall within the design
parameters specified in the early site permit. A combined license
application that references a design certification will involve a
comparison to ensure that the actual characteristics of the site chosen
by the combined license applicant fall within the site parameters in
the design certification. Similarly, if a combined license applicant
references both an early site permit and a design certification, the
NRC will review the application to ensure that the site characteristics
in the early site permit fall within the site parameters in the
referenced design certification and that the actual design
characteristics fall within the design parameters in the early site
permit.
A new definition of major features of the emergency plans is added
to explain what aspects of emergency preparedness--short of full and
integrated emergency plans--an early site permit applicant may seek
approval of under Sec. 52.17(b)(2)(i). A major feature may consist of
a specific aspect of a plan necessary to address in whole or part 1 or
more of the 16 planning standards in 10 CFR 50.47(b). Additional
requirements for each of the planning standards are set forth in part
50, appendix E, and the applicant may choose to demonstrate compliance
with one or more provisions in appendix E, either in addition to or
without a full demonstration of compliance with a planning standard in
Sec. 50.47(b), when seeking approval of part of a major feature. A
major feature may also be a description of one or both of the emergency
planning zones (EPZs) required by 10 CFR 50.33(g). Regulatory
considerations governing EPZs are set forth in Sec. 50.33(g); a major
feature need not address all of these considerations.
A definition of prototype plant is added to explain the type of
nuclear power plant that the Commission intended in the former Sec.
52.47(b) (new Sec. 50.43), and Sec. 52.157(e)). A prototype plant is
a licensed nuclear reactor test facility that is similar to and
representative of either the first-of-a-kind or standard nuclear plant
design in all features and size, but may have additional safety
features. The purpose of the prototype plant is to perform testing of
new or innovative safety features for the first-of-a-kind nuclear plant
design, as well as being used as a commercial nuclear power facility.
Section 52.2 Interpretations
This section, formerly designated as Sec. 52.5, remains unchanged.
It provides that the only interpretations of part 52 that are legally
binding on the Commission are interpretations provided by the General
Counsel. These written interpretations, which are rarely provided by
the General Counsel, are set forth in 10 CFR part 8.
Section 52.3 Written Communications
This new section, which is analogous to Sec. 50.4, sets forth
administrative requirements regarding written communications with the
NRC, including the addressing of such communications, and listings of
the various NRC offices and officials who must receive copies of
different types of communications (e.g., applications for licenses and
license amendments, security plan and related submissions, quality
assurance related submissions). The administrative requirements
themselves are identical to those in Sec. 50.4; they are reproduced in
Sec. 52.3 to make clear that they apply to applicants for and holders
of permits, licenses, and regulatory processes that are contained in
part 52.
Section 52.4 Deliberate Misconduct
This section, formerly designated as Sec. 52.9, has been
substantially rewritten in order to more clearly delineate the
applicability of the proscriptions against deliberate misconduct to all
delineated part 52 entities, including applicants for and holders of
standard design approvals, and applicants for standard design
certifications (including those applicants whose designs are certified
by the Commission in a standard design certification rulemaking).
Although the regulatory language in Sec. 52.4 differs from former
Sec. 52.9, no substantive change in any aspect of the Commission law
or the underlying policy considerations is being made by the
Commission's adoption of Sec. 52.4. The relevant law and policy
considerations for former Sec. 52.9 are merely clarified and extended
in Sec. 52.4 to cover applicants for and holders of permits, licenses,
and regulatory processes that are contained in part 52.
Section 52.5 Employee Protection
This new section, which is analogous to Sec. 50.7, prohibits
discrimination against employees for engaging in protected activities
established in Section 211 of the Energy Reorganization Act of 1974, as
amended (1974 ERA). These protected activities, which are listed in
Sec. 52.5(a)(1), include (but are not limited to) providing the
Commission or the employer information about alleged violations of the
AEA or 1974 ERA, of any of the Commission's regulations. No substantive
change in any aspect of the Commission law or the underlying policy
considerations with respect to employee protection is being made by the
Commission adoption of Sec. 52.5; the relevant law and policy
considerations for former Sec. 50.7 are merely clarified and extended
in Sec. 52.5 to cover applicants for and holders of permits, licenses,
and regulatory processes that are contained in part 52 (currently,
standard design approvals and standard design certifications).
Section 52.6 Completeness and Accuracy of Information
This new section, which is analogous to Sec. 50.9, requires that
all information submitted to the NRC by the delineated part 52 entities
be complete and accurate, and imposes a reporting requirement on such
entities with respect to information with respect to the regulated
activity having a significant implication for public health and safety
or common defense and security. No substantive change in any aspect of
the Commission law or the
[[Page 49438]]
underlying policy considerations is being made by the Commission
adoption of Sec. 52.6; the relevant law and policy considerations
underlying Sec. 50.9 are merely clarified and extended to cover
applicants for and holders of permits, licenses and regulatory
processes that are contained in part 52. For example, Sec. 50.9 does
not impose a positive obligation on licensees to seek out new
information meeting the reporting thresholds in the rule. In applying
Sec. 52.6, the Commission would extend this interpretation to part 52
entities such as combined license holders and standard design
certification applicants (including applicants whose applications were
approved, for the regulatory life of the certification rule).
Section 52.7 Specific Exemptions
This new section, which is analogous to Sec. 50.12, provides for
specific procedures and criteria for Commission grants of exemptions
from the provisions of part 52. No substantive change in any aspect of
the Commission law or the underlying policy considerations is being
made by the Commission adoption of Sec. 52.7; the relevant law and
policy considerations underlying Sec. 50.12 are merely extended to
part 52.
The NRC notes that the exemption provisions in Sec. 52.7 do not
supercede or otherwise diminish more specific exemption provisions that
are in part 52, such as the provision of a specific design
certification rule or Sec. 52.63(b)(1) governing exemptions from one
or more elements of a design certification rule. An applicant or
licensee referencing a standard design certification rule who wishes to
obtain an exemption from one or more elements must meet the criteria in
the specific design certification rule or Sec. 52.63(b)(1). If the
applicant or licensee is unable to demonstrate compliance with those
criteria, then it may request an exemption under the more encompassing
authority of Sec. 52.7. However, the exemption request must then
demonstrate compliance with the additional criteria in Sec. 52.7.
The Commission also notes that Sec. 52.7 does not supercede the
applicability of more specific dispensation provisions in other parts
of Chapter I. For example, a holder of a combined license would not
require a separate part 52 exemption in order to obtain approval of an
alternative to a provision of an applicable ASME Code provision that is
otherwise required under 10 CFR 50.55a; the licensee need only satisfy
the criteria in Sec. 50.55a(a)(3). However, in the absence of a more
specific dispensation provision, the Commission intends to utilize
Sec. 52.7 as a means for granting dispensation from compliance with
Commission requirements in other parts of 10 CFR Chapter I. The person
requesting an exemption need only address the Sec. 52.7 criteria as
applied to the underlying requirement for which dispensation from
compliance is sought, and need not also address dispensation from
compliance with the relevant part 52 requirement. For example, the
holder of the combined license who wishes dispensation from compliance
with a fire protection requirement in 10 CFR 50.48 need only address
the relevant criteria in Sec. 52.7 with respect to the reasons for
dispensation from compliance with Sec. 50.48. The holder need not
address dispensation from compliance with Sec. 52.0, which otherwise
makes applicable the provisions of Sec. 50.48 on the licensee. Any
exemption granted by the Commission would address the reasons for
dispensation with the underlying requirement--in this case, Sec.
50.48, and would also provide dispensation from compliance with Sec.
52.0.
Section 52.8 Combining Licenses; Elimination of Repetition
This new section includes provisions analogous to Sec. Sec. 50.31,
50.32, and 50.52 and is added to clarify that these regulatory
provisions also apply to part 52 licenses. Paragraph (a), which is
analogous to Sec. 50.31, is added to make clear that an applicant for
a license under part 52 may combine in one application, several
applications for different kinds of licenses under various regulations
in 10 CFR Chapter I. Section 50.31 currently provides that an applicant
may combine in one application, several applications for different
kinds of licenses under various regulations in 10 CFR Chapter I. The
plain reading of this language, given that this provision is located in
part 50, is that a part 50 application may contain in one application
other applications for different licenses in other parts of 10 CFR
Chapter I. Thus, Sec. 50.31 would not appear to allow a part 52
application (as for a combined license) to combine in one application
other applications for different license in other parts of 10 CFR
Chapter I. Accordingly, paragraph (a) makes clear that a part 52
application may be combined with application for different licenses in
other parts of 10 CFR Chapter I.
Paragraph (b), which is analogous to Sec. 50.32, is added to make
clear that an applicant for a license, standard design certification,
or design approval under part 52 may incorporate by reference in its
application information contained in other documents provided to the
Commission, but that such incorporation must clearly specify the
information to be incorporated.
Paragraph (c), which is analogous to Sec. 50.52, is added to
clarify the Commission's authority under Section 161.h of the AEA to
combine NRC licenses, such as a special nuclear materials license under
part 70 for the reactor fuel, with a combined license under part 52.
Analogous to the situation with respect to Sec. 50.31, the language in
Sec. 50.52 would not appear to allow the Commission to combine into a
single part 52 license, other non-part 52 licenses. No substantive
change in any aspect of the Commission law or the policy considerations
underlying Sec. Sec. 50.31, 50.32, and 50.52 is being made by the
Commission adoption of Sec. 52.8; the relevant law and policy
considerations underlying Sec. Sec. 50.31, 50.32, and 50.52 are merely
extended to part 52.
Section 52.9 Jurisdictional Limits
This new section, which is analogous to Sec. 50.53, makes clear
that no approval provided by the Commission under part 52 addresses or
approves in any manner activities which are not under or within the
territorial jurisdiction of the United States. As a practical matter,
this means that an approval or license issued by the NRC under part 52
has no legal effect outside the territorial jurisdiction of the United
States. No substantive change in any aspect of the Commission law or
the policy considerations underlying Sec. 50.53 is being made by the
Commission adoption of Sec. 52.9; the relevant law and policy
considerations are merely extended to part 52.
Section 52.10 Attacks and Destructive Acts
This new section, which is analogous to Sec. 50.13, applies the
existing Commission law and policy that a licensee need not provide for
design features or other measures to protect against certain attacks
and destructive acts, or the use or deployment of weapons incident to
U.S. defense activities, to the applicants for and holders of permits,
licenses and other approvals under part 52. No substantive change in
any aspect of the Commission law or the underlying policy
considerations is being made by the Commission adoption of Sec. 52.10;
the relevant law and policy considerations for the Sec. 50.13
exclusion are merely extended to cover applicants for and holders of
permits, licenses, and regulatory processes that are contained in part
52.
[[Page 49439]]
Section 52.11 Information Collection Requirements: OMB Approval
This section, formerly designated as Sec. 52.8, remains unchanged.
It gives notice that all information collection and reporting
requirements in part 52 have been approved by the Office of Management
and Budget. No requirement, action or responsibility is imposed on part
52 entities by this section.
Subpart A--Early Site Permits
Section 52.12 Scope of Subpart
This section describes the scope of this licensing process. Under
this subpart an applicant can request pre-approval of a site (so-called
site banking), separate from other licensing actions, and subsequently
reference that early site permit in a future application to build a
nuclear power plant. This process was created for proposed sites that
the applicant may not plan to use in the near term.
Section 52.13 Relationship to Other Subparts
This section explains the relationship of the early site permit
process to the construction permit process under 10 CFR part 50 and to
the combined license process under part 52.
Section 52.15 Filing of Applications
This section explains who can file, how to file, and the fees for
NRC review of an application for an early site permit.
Section 52.16 Contents of Applications; General Information
This section sets forth the type of general information that is
required to be included in an early site permit application, namely,
the information required by 10 CFR 50.33(a) through (d) and (j).
Section 50.33 requires that the application include information such as
the name and address of the applicant, a description of the business or
occupation of the applicant, and citizenship information of the
applicant. Section 50.33 also provides requirements for the handling of
Restricted Data or other defense information in an application.
Section 52.17 Contents of Applications; Technical Information
The purpose of this section is to set forth the type of technical
information to be included in an application for an early site permit.
Paragraph (a)(1) identifies the information needed for the site safety
review, excluding emergency planning information. The site safety
information is a subset of the information required of applicants for
construction permits. Although an ESP applicant does not need to
specify a particular nuclear plant design, as in construction permit
applications, it does need to provide sufficient surrogate design
information (developed to bound the nuclear plant design(s) that are
being considered by the applicant) so that the NRC can make a
determination on the acceptability of the site and the environmental
impacts, and determine whether designs bounded by the surrogate design
information provided by the applicant can be qualified for the proposed
site. The application must contain, among other things, the specific
number, type (e.g., pressurized-water reactor), and thermal power level
of the facilities, or range of possible facilities, for which the site
may be used; the anticipated maximum levels of radiological and thermal
effluents each facility will produce; the type of cooling systems,
intakes, and outflows that may be associated with each facility; the
boundaries of the site; and the proposed general location of each
facility on the site. As part of the description of the proposed
general location of each facility on the site (Sec. 52.17(a)(1)(v)),
the applicant should describe the foot print for all structures and
external safety-related design features proposed for the site.
The application must also include the seismic, meteorological,
hydrologic, and geologic characteristics of the proposed site with
appropriate consideration of the most severe of the natural phenomena
that have been historically reported for the site and surrounding area
and with sufficient margin for the limited accuracy, quantity, and
period of time in which the historical data have been accumulated. This
information is to ensure that future plants built at the site would be
in compliance with General Design Criterion 2 from appendix A to part
50, which requires that structures, systems, and components important
to safety be designed to withstand the effects of natural phenomena
such as earthquakes, tornadoes, hurricanes, floods, tsunami, and
seiches without loss of capability to perform their safety functions.
The application must also include the location and description of
any nearby industrial, military, or transportation facilities and
routes, and the existing and projected future population profile of the
area surrounding the site. The application must contain an analysis and
evaluation of the major structures, systems, and components of the
facility that bear significantly on the acceptability of the site from
a radiological safety standpoint. In addition, the application must
demonstrate that adequate security plans and measures can be developed
for the site and must provide a description of the quality assurance
program applied to site-related activities.
Paragraph (a)(2) identifies that the application must include an
environmental report that meets the requirements of Sec. 51.50(b).
Environmental reports must focus on the environmental effects of
construction and operation of a nuclear reactor, or reactors, which
have characteristics that fall within the design parameters postulated
in the early site permit. Environmental reports must also include an
evaluation of alternative sites to determine whether there is any
obviously superior alternative to the site proposed. Environmental
reports submitted in an early site permit application are not required
to but may include an assessment of the economic, technical, and other
benefits and costs of the proposed action or an analysis of other
energy alternatives.
Paragraph (b) identifies the emergency planning information to be
included in the application. All ESP applicants are required to
identify in the site safety analysis report (SSAR) physical
characteristics unique to the proposed site that could pose a
significant impediment to the development of emergency plans, e.g., a
physical characteristic or combination of physical characteristics that
could pose major difficulties for evacuation or the taking of other
protective actions. In addition, if the applicant identifies such
physical characteristics, the application must identify measures that
would, when implemented, mitigate or eliminate the significant
impediment. After meeting this mandatory requirement, paragraph (b)
allows applicants the option of either submitting major features of
emergency plans or complete and integrated emergency plans for approval
by the NRC, in consultation with the Department of Homeland Security
(DHS). For complete and integrated emergency plans, the applicant must
include the proposed inspections, tests, and analyses that the holder
of a combined license referencing the early site permit shall perform,
and the acceptance criteria that are necessary and sufficient to
provide reasonable assurance that, if the inspections, tests, and
analyses are performed and the acceptance criteria met, the facility
has been constructed and will operate in conformity with the license,
the provisions of the Atomic Energy Act,
[[Page 49440]]
and the NRC's regulations. The inclusion of such inspections, tests,
analyses, and acceptance criteria (ITAAC) is necessary to allow the NRC
to make the finding that the plans submitted by the applicant provide
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency. Paragraph (b) also
allows applicants proposing major features of emergency plans to
include proposed ITAAC. Where the applicant is submitting a complete
and integrated emergency plan, a utility plan must be submitted if any
offsite agencies elect not to participate in the development of
emergency planning information.
If the applicant plans to perform the preparations for construction
activities identified in 10 CFR 50.10(e)(1), then paragraph 52.17(c)
requires the applicant to describe the activities it is requesting to
perform and propose a redress plan that, if carried out, would achieve
a ``self-maintaining, environmentally stable, and aesthetically
acceptable site'' that conforms to local zoning laws. Redress plans are
expected to be modeled on the redress requirements imposed on the
Clinch River Breeder Reactor project (see In the Matter of the U.S.
Department of Energy, et al., LBP-85-7, 21 NRC 507 (1985)). By
containing a redress plan, the ESP will constitute assurance that, if
site preparation activities are conducted but the site is never used
for a nuclear power plant, the site will be returned to an acceptable
and stable condition.
Section 52.18 Standards for Review of Applications
This section identifies the regulations that the NRC staff will use
in performing its review of an application for an early site permit,
including the standards that the NRC staff will use in performing its
assessment of emergency preparedness information provided in the ESP
application.
Section 52.21 Administrative Review of Applications; Hearings
This section identifies the procedural requirements that apply to
the mandatory hearing for the early site permit licensing process. This
section also clarifies that the applicant's environmental report is not
required to but may include an assessment of the benefits of
construction and operation of the reactor or reactors, or an analysis
of alternative energy sources. In addition, the presiding officer in an
ESP hearing is prohibited from admitting contentions on these matters
if those issues were not addressed in the early site permit
application.
Section 52.23 Referral to the Advisory Committee on Reactor Safeguards
(ACRS)
This section states that the ACRS will report on those portions of
the application which concern safety which is the same role the ACRS
had with respect to construction permits in the past.
Section 52.24 Issuance of Early Site Permit
The purpose of this section is to set forth the timing of issuance
of an ESP and the findings that the Commission must make to issue the
ESP, including that issuance of the permit will not be inimical to the
common defense and security or to the health and safety of the public,
that the applicant is technically qualified to engage in activities
necessary to prepare the ESP application and any site preparation
activities that the applicant is seeking approval to perform, and that
the findings required by subpart A of 10 CFR part 51 regarding the NRC
staff's assessment of the environmental impact have been made.
This section also requires that the early site permit specify the
site characteristics, design parameters, and terms and conditions of
the early site. Before issuance of either a construction permit or a
combined license referencing an early site permit, the Commission must
find that any relevant terms and conditions of the early site permit
have been met. Any terms or conditions that could not be met by the
time of issuance of the construction permit or combined license must be
set forth as terms or conditions of the construction permit or combined
license. Finally, this section requires that the early site permit
specify the site preparation activities under Sec. 52.17(c) that the
permit holder is authorized to perform.
Section 52.25 Extent of Activities Permitted
This section specifies that, if the construction preparation
activities authorized by Sec. 52.24(c) are performed and the site is
not referenced in a application for a construction permit or a combined
license while the permit remains valid, then the early site permit
remains in effect for the purpose of site redress with the goal of
achieving an environmentally stable and aesthetically acceptable site.
Section 52.27 Duration of Permit
The purpose of paragraph (a) of this section is to specify the
duration of an early site permit. The applicant can request a duration
of up to 20 years. Paragraph (b) describes the conditions under which
an ESP can continue to be valid beyond its expiration date. Paragraph
(c) allows an applicant for a construction permit or combined license,
at its own risk, to reference an ESP that is under review by the NRC
but not yet granted. Paragraph (d) explains that, upon issuance of a
construction permit or combined license, a referenced early site permit
is subsumed, to the extent referenced, into the construction permit or
combined license. By ``subsumed'' the NRC means that the information
that was contained in the early site permit SSAR becomes part of the
referencing combined license FSAR upon issuance of the combined
licenses in the same manner as if the combined license applicant had
not referenced an early site permit. The NRC is including the phrase
``to the extent referenced,'' to indicate that it is not all of the
information submitted in the early site permit application that is
subsumed into the combined license, but, rather, only that information
that is contained in the SSAR and identified by the applicant as being
referenced in the combined license application. This subsumption of the
early site permit into the referencing license affects the way changes
to the early site permit information will be handled because it breaks
the tie to the finality provisions in Sec. 52.39. After issuance of
the construction permit or combined license, Sec. 52.39 no longer
applies to the early site permit information and such information will
be covered by the same finality provisions as the rest of the
information in the FSAR (with the exception of any referenced design
certification information), as outlined in Sec. 52.98 (e.g., in
accordance with Sec. Sec. 50.54, 50.59, etc.).
Section 52.28 Transfer of Early Site Permit
This section specifies the requirements to be followed if a holder
of an early site permit wants to transfer the ESP to another person or
company.
Section 52.29 Application for Renewal
Paragraph (a) of this section explains the contents and timing of
an application for renewal of an early site permit. Paragraph (b) sets
forth the procedure for requesting a hearing on the application for
renewal. Paragraph (c) explains that an ESP may remain in effect beyond
its expiration under
[[Page 49441]]
certain circumstances. Specifically, an ESP for which a timely
application for renewal has been filed remains in effect until the
Commission has determined whether to renew the permit. If an ESP is not
renewed, it continues to be valid in any proceeding on an application
for a construction permit or a combined license which references the
ESP and was docketed prior to the expiration of the ESP. Finally,
paragraph (d) identifies the responsibilities of the ACRS on an ESP
renewal application.
Section 52.31 Criteria for Renewal
Paragraph (a) of this section sets forth the criteria for granting
a renewal of an early site permit and provides that, if the NRC wants
to impose new requirements, it must demonstrate that the new
requirements meet the backfit standard from Sec. 50.109. Paragraph (b)
explains that even if an application for renewal of an ESP is denied by
the NRC, the applicant can submit a new application for an ESP that
corrects the problems with the application for renewal.
Section 52.33 Duration of Renewal
This section specifies the duration of a renewed early site permit.
An ESP may, upon application, be extended for periods of up to 20 years
beyond the previously approved duration, provided the criteria in Sec.
52.31 are met.
Section 52.35 Use of Site for Other Purposes
The purpose of this section is to explain how the holder of an
early site permit could use the site for other activities. An approved
site may be used for purposes not related to the construction of a
nuclear power facility, e.g., a fossil-fueled station or a park,
provided that the Commission is informed of all significant non-nuclear
uses prior to actual construction or site modification activities. A
permit may be revoked if a non-nuclear use would interfere with a
nuclear use, or would so alter the site that important assumptions
underlying the issuance of the permit were called into question.
Section 52.39 Finality of Early Site Permit Determinations
This section specifies the special backfit requirements that apply
to an early site permit. Paragraph (a) provides requirements regarding
finality of ESP issues as they relate to the Commission. Paragraph
(a)(1) states that, notwithstanding any provision in 10 CFR 50.109
(Backfitting), while an early site permit or renewed early site permit
is in effect, the Commission may not change or impose new site
characteristics, design parameters, or terms and conditions, including
emergency planning requirements, on the early site permit unless the
Commission meets one of four conditions. Those conditions are that the
Commission either determines that a modification is necessary to bring
the permit or the site into compliance with the Commission's
regulations and orders applicable and in effect at the time the permit
was issued; determines that a modification is necessary to assure
adequate protection of the public health and safety or the common
defense and security; determines that a modification is necessary based
on an update under Sec. 52.39(b); or issues a variance requested under
Sec. 52.39(d).
Paragraph (a)(2) addresses the finality of an early site permit for
a license that references the early site permit and requires that the
Commission treat as resolved those matters resolved in the proceeding
on the application for issuance or renewal of the early site permit,
except as provided for in Sec. Sec. 52.39(b), (c), and (d). This
paragraph also addresses finality of changes to an early site permit
approved emergency plan (or major features thereof).
Paragraph (b) requires a license applicant that references an ESP
to update and correct the emergency preparedness information that was
provided in the ESP and to discuss whether the new information
materially changes the bases for compliance with the applicable NRC
requirements. New information which materially changes the bases for
compliance includes: (1) Information which substantially alters the
bases for a previous NRC conclusion with respect to the acceptability
of a material aspect of emergency preparedness or an emergency
preparedness plan, and (2) information which would constitute a
sufficient basis for the Commission to modify or impose new terms and
conditions related to emergency preparedness, in accordance with Sec.
52.39(a)(1). New information which materially changes the Commission's
determination of the matters in Sec. 52.17(b), or results in
modifications of existing terms and conditions by the NRC under Sec.
52.39(a)(1) would be subject to litigation during the licensing
proceedings in accordance with Sec. 52.39(c).
Section 52.39(c) provides requirements for the submittal of
contentions in a proceeding for the issuance of a license referencing
an early site permit and for the filing of petitions requesting that an
early site permit be modified, suspended, or revoked. Paragraph (c)(1)
states that contentions on several matters may be litigated in the
proceeding on a combined license that references an early site permit.
Matters that may be litigated include contentions related to the
following: (1) The nuclear power reactor proposed to be built does not
fit within one or more of the site characteristics or design parameters
included in the early site permit; (2) one or more of the terms and
conditions of the early site permit have not been met; (3) a variance
requested under Sec. 52.39(d) is unwarranted or should be modified;
(4) new or additional information is provided in the application that
substantially alters the bases for a previous NRC conclusion or
constitutes a sufficient basis for Commission to modify or impose new
terms and conditions related to emergency preparedness; or (5) any
significant environmental issue that was not resolved in the early site
permit proceeding, or any issue involving the impacts of construction
and operation of the facility that was resolved in the early site
permit proceeding for which significant new information has been
identified. An issue related to the impacts of construction and
operation of the facility resolved in the early site permit proceeding
is afforded finality at the combined license stage provided that there
is no ``new and significant'' information on the issue. If an
environmental issue was not resolved at the early site permit stage,
either because information was not sufficient to resolve it or because
the early site permit applicant was permitted to defer it (e.g., need
for power analysis), then the combined license applicant would need to
address the issue in its combined license application. The NRC, in the
context of a combined license application that references an early site
permit, has defined the term ``new'' in the phrase ``new and
significant information'' as any information that was both (1) not
considered in preparing the ESP environmental report or EIS (as may be
evidenced by references in these documents, applicant responses to NRC
requests for additional information, comment letters, etc.) and (2) not
generally known or publicly available during the preparation of the EIS
(such as information in reports, studies, and treatises). This new
information may or may not be significant. For an issue to be
significant, it must be material to the issue being considered, i.e.,
it must have the potential to affect the NRC staff's evaluation of the
issue. The COL applicant need only provide information about a
previously resolved
[[Page 49442]]
environmental issue if it is both new and significant.
Paragraph (c)(2) allows any person to file a petition requesting
that the site characteristics, design parameters, or terms and
conditions of the early site permit be modified, or that the permit be
suspended or revoked. The petition will be considered in accordance
with Sec. 2.206. Section 2.206 provides that any person may file a
request to institute a proceeding to modify, suspend, or revoke a
license, or for any other action as may be proper. Section 52.39(c)(2)
addresses the Commission's required action on such a petition and
states that construction under the construction permit or combined
license will not be affected by the granting of the petition unless the
Commission makes the order immediately effective.
Paragraph (d) provides that an applicant for a license or an
amendment to such a license who has filed an application referencing an
early site permit may request a variance from one or more site
characteristics, design parameters, or terms and conditions of the
early site permit, or from the SSAR. This paragraph also states that,
once a construction permit or combined license referencing an early
site permit is issued, a variance from the early site permit will not
be granted for that construction permit or combined license. At that
point, the early site permit is subsumed into the combined license and
any request for a change to the terms or conditions of the combined
license is a request for a license amendment that must be filed under
the provisions of Sec. 50.90.
The NRC is adding new paragraph (e) in the final rule in response
to public comments expressing support for adding provisions to provide
an early site permit holder with the option of requesting an amendment
to the early site permit in order to resolve issues that were not
addressed in the original early site permit review or to achieve
finality on updated early site permit information. Paragraph (e) states
that the holder of an early site permit may not make changes to the
early site permit, including the SSAR, without prior Commission
approval. The request for a change to the early site permit must be in
the form of an application for a license amendment, and must meet the
requirements of 10 CFR 50.90 and 50.92. The NRC considers an early site
permit SSAR to be equivalent to a combined license FSAR; therefore,
when an early site permit is amended, the SSAR must be revised
consistent with the ESP amendments. In addition, the SSAR retains
continuing viability for early site permits that are for multiple units
after it is referenced in the first combined license. However, unlike
an FSAR, there is no change process for the SSAR that does not require
NRC review and approval.
Finally, the Commission is adding a new paragraph (f) (proposed
paragraph (e)) to the ``finality'' section in each subpart of part 52,
including Sec. 52.39, entitled ``Information requests,'' which
delineates the restrictions on the NRC for information requests to the
holder of the early site permit. This provision is analogous to the
former provision on information requests in paragraph 8 of appendix O
to parts 50 and 52, and is based upon the language of Sec. 50.54(f).
For early site permits, this provision is contained in Sec. 52.39(f),
and requires the NRC to evaluate each information request on the holder
of an early site permit to determine that the burden imposed by the
information request is justified in light of the potential safety
significance of the issue to be addressed in the information request.
The only exceptions would be for information requests seeking to verify
compliance with the current licensing basis of the early site permit.
If the request is from the NRC staff, the request would first have to
be approved by the Executive Director for Operations (EDO) or his or
her designee.
Subpart B--Standard Design Certifications
Section 52.41 Scope of Subpart
This section describes the scope of this licensing process for
certification of standard nuclear power plant designs. Under this
subpart, an applicant may request pre-approval of either an
evolutionary light-water or advanced nuclear power plant design,
separate from a site review or other licensing action, and subsequently
reference that certified design in an application to build a nuclear
power plant. The requirements for the type of plant to be certified
were moved from Sec. 52.45 to this section. The scope of the standard
plant design must be essentially complete as described in Sec.
52.47(c).
Section 52.43 Relationship to Other Subparts
The purpose of this section is to explain the relationship of the
design certification process to the processes set forth in subparts C,
E, and F of 10 CFR part 52, which provide for combined licenses,
standard design approvals, and manufacturing licenses. The requirement
to hold a final design approval under former appendix O to part 52 as a
prerequisite to design certification was deleted from Sec. 52.45.
However, applicants for design certification have the option of also
applying for a standard design approval under subpart E. Also,
applicants for a manufacturing license may reference a certified
design.
Section 52.45 Filing of Applications
This revised section is similar to the ``filing of applications''
sections in subparts A and C of this part. This section explains how to
file an application for design certification and how the fees for NRC's
review of the application will be assessed. Because design
certification is a rule and not a license, the applicant for design
certification does not need to be a U.S. citizen or company (AEA,
Section 103).
Section 52.46 Contents of Applications; General Information
This is a new section and it is similar to the ``general
information'' sections in subparts A and C of this part. It identifies
the general information that must be included in all applications.
Section 52.47 Contents of Applications; Technical Information
The purpose of this section is to identify the technical
information that must be included in an application for design
certification. This section was revised to provide a comprehensive list
of requirements for a design certification application. Paragraphs (a)
and (c) describe the information that must be included in the FSAR,
which is included in the application, and paragraph (b) describes the
information that must also be included in the application but does not
need to be included in the FSAR. Paragraph (c) describes additional
requirements for particular types of applications. This section also
specifies the level of detail for the design information that must be
provided in an application.
Many of the requirements in this section were taken from 10 CFR
50.34 or are pointers to technical requirements in parts 20, 50, 51,
and 73 that must be addressed in the application. The requirements
taken from Sec. 50.34 are a subset of the information required of
applicants for construction permits and operating licenses. Other
requirements came from the original version of 10 CFR 52.47 or were
developed by the Commission during the initial design certification
reviews (e.g., SECY-93-087, ML003708021).
Although an applicant for design certification does not need to
specify a particular site for the nuclear power plant, as in a combined
license application, it does need to identify the site parameters,
under paragraph (a)(1),
[[Page 49443]]
that the standard nuclear power plant is designed to meet, e.g.,
postulated values for the safe-shutdown earthquake response spectra and
maximum tornado wind speed. These parameters are usually selected to
envelop a large portion of existing nuclear plant sites in the United
States. Once the design is certified by the NRC, conformance of the
actual site with the established site parameters must be demonstrated
by the applicant for a combined license and verified by the NRC when
the application is submitted.
Paragraph (a)(7) requires the applicant for design certification to
describe its qualifications to design and analyze a standard nuclear
power plant, which may become part of the bases for a future license.
Paragraph (a)(13) requires the applicant to provide the electric
equipment list required by Sec. 50.49(d). The NRC understands that the
applicant may not be able to establish qualification files for all
applicable components.
In its staff requirements memorandum (SRM) on SECY-90-377,
``Requirements for Design Certification under 10 CFR part 52,'' dated
February 15, 1991, the Commission directed the staff to ensure that the
design certification process preserves operating experience insights in
the certified design. Therefore, for plant designs that are based on or
are evolutions of nuclear plants that have operated in the United
States, paragraph (a)(22) requires the applicant to demonstrate how
relevant operating experience insights, from NRC's generic letters and
bulletins issued after the most recent revision of the applicable SRP
and 6 months before the docket date of the application, have been
incorporated into the plant design. Operating experience includes
consideration of operating events and the reliability and performance
of structures, systems, and components. If the application is for a
design that is not based on or is not an evolution of a nuclear plant
that operated in the United States, the applicant must demonstrate how
insights from any relevant international operating experience have been
incorporated into that plant design.
In its SRMs, dated June 26, 1990, and July 21, 1993, on SECY-90-16,
``Evolutionary Light-Water Reactor Certification Issues and their
Relationship to Current Regulatory Requirements,'' and SECY-93-087,
``Policy, Technical, and Licensing Issues Pertaining to Evolutionary
and Advanced Light-Water Reactor Designs,'' respectively, the
Commission approved NRC staff recommendations for selected preventative
and mitigative design features for future light-water reactor designs.
Paragraph (a)(23) requires the applicant to provide a description and
analysis of those design features discussed in SECY-90-16 and SECY-93-
087. Postulated severe accidents are not design-basis accidents (DBAs)
and the severe accident design features do not have to meet the
requirements for DBAs. However, the severe accident design features are
part of a plant's design bases information.
Paragraph (a)(24) requires the applicant to provide a conceptual
design for those design features that are outside the scope of the
certified design, e.g., service water intake structure or ultimate heat
sink.
Paragraph (a)(25) requires the applicant to describe the interface
requirements for those design features that are outside the scope of
the certified design, e.g., service water intake structure or ultimate
heat sink. Paragraph (a)(26) requires justification that the interface
requirements can be verified with the ITAAC for the plant.
Paragraph (a)(27) requires the applicant to provide a description
of the design-specific PRA and its results. Guidance on how to meet the
PRA information requirement will be provided in separate regulatory
guidance documents.
Paragraph (b)(1) requires the applicant to provide the ITAAC that
are necessary and sufficient to demonstrate that a facility that
references the design certification has been constructed and will be
operated in conformity with the design certification, the Atomic Energy
Act of 1954, as amended, and the Commission's rules and regulations.
These ITAAC will be a part of the Commission's verification program and
must cover all of the design information that is within the scope of
the certified design. ITAAC for the remaining design features that are
outside of the scope of the certified design will be provided in a
combined license application that references the design certification
rule.
In its SRM on SECY-91-229, ``Severe Accident Mitigation Design
Alternatives for Certified Standard Designs,'' dated October 25, 1991,
the Commission approved the staff's recommendation that design
certification applicants assess SAMDAs for their standard plant
designs. The Commission required SAMDA evaluations in order to achieve
greater finality for the design features that are resolved in design
certification rulemakings. For further explanation, see discussion in
SECY-93-087, dated April 2, 1993. In order to implement this
requirement, paragraph (b)(2) requires the applicant to provide a SAMDA
evaluation for the standard plant design. This assessment is distinct
from, and in addition to, the requirement in paragraph (a)(23) to
provide a description and analysis of severe accident design features.
Paragraph (c)(1) requires an essentially complete scope of design
in applications for evolutionary nuclear power plants. These plants are
improved versions of light-water reactor designs that were in operation
when part 52 was originally codified. Examples of evolutionary designs
include General Electric's U.S. Advanced Boiling Water Reactor and
Westinghouse's SP/90 and System 80+ designs. Evolutionary designs do
not have to meet the design qualification testing requirements set
forth in 10 CFR 50.43(e).
Paragraph (c)(2) requires applications for ``advanced'' nuclear
power plants to provide an essentially complete scope of design and
meet the design qualification testing requirements in 10 CFR 50.43(e).
Advanced designs differ significantly from evolutionary light-water
reactor designs or incorporate, to a greater extent than evolutionary
designs do, simplified, inherent, passive, or other innovative means to
accomplish their safety functions. Examples of advanced nuclear power
plant designs include General Atomic's Modular High Temperature Gas-
Cooled Reactor, General Electric's Simplified Boiling Water Reactor,
and Westinghouse's AP600.
Paragraph (c)(3) requires applications for modular nuclear power
plant designs to describe and analyze the possible operating
configurations of reactor modules. Modular designs are defined in Sec.
52.1. Modular plant designs are not portions of a single nuclear plant,
rather they are separate nuclear power reactors with some shared or
common systems.
Section 52.48 Standards for Review of Applications
This section sets forth the parts of 10 CFR that contain applicable
requirements for the technical review of design certification
applications. The applicability of these requirements to the design
certification process is specified in the identified parts. The
Commission recognizes that new designs may incorporate design features
that are not addressed by the current standards set out in 10 CFR parts
20, 50 and its appendices, 51, 73, or 100, and that new standards may
be required to address these new design features. The Commission will
determine whether additional rulemakings are needed or appropriate to
resolve generic safety
[[Page 49444]]
issues that are applicable to multiple designs. On the other hand, new
design features that are unique to a particular design could be
addressed in the design certification rulemaking for that particular
design.
Section 52.51 Administrative Review of Applications
This section sets forth the procedures for performing a notice and
comment rulemaking for design certification. Paragraph (b) states that
the Commission will determine, at its sole discretion, whether to hold
a legislative hearing on the proposed design certification rule under
the procedures in subpart O of 10 CFR part 2. Paragraph (c) states that
proprietary information contained in an application for design
certification will be given the same treatment that such information
would be given in a proceeding on an application for a construction
permit or an operating license under 10 CFR part 50. This gives the
design certification applicant (vendor) an opportunity to treat
elements of its design as trade secrets.
Section 52.53 Referral to the Advisory Committee on Reactor Safeguards
(ACRS)
This section states that the application for design certification
shall be sent to the ACRS for its review of safety issues.
Section 52.54 Issuance of Standard Design Certification
Paragraph (a) of this section sets forth the findings that the
Commission must make in order to issue a design certification rule.
Paragraph (b) requires that site parameters, design characteristics,
and any additional requirements and restrictions be specified in the
design certification rule. Previous DCRs set forth the additional
requirements and restrictions in Section IV of the rule. Site
parameters and design characteristics are defined in Sec. 52.1 and can
be specified in the design control document. These values will be used
during the review of a combined license application that references the
design certification rule to verify that the standard plant design
conforms with the characteristics of the actual site and the design
parameters used in the early site permit.
Section 52.54 was amended to include a new paragraph (c) which
requires that every DCR contain a provision stating that, after the
Commission has adopted the final DCR, the applicant for that design
certification will not permit any individual to have access to, or any
facility to possess, Restricted Data or classified National Security
Information until the individual and/or facility has been approved for
access under the provisions of 10 CFR parts 25 and/or 95. The NRC
believes that this amendment, along with the changes to parts 25, 95,
and Sec. 50.37, are necessary to ensure that access to classified
information is adequately controlled by all entities applying for NRC
certifications.
Section 52.55 Duration of Certification
The purpose of this section is to specify the duration that a
standard design certification is valid for referencing in a combined
license application.
Section 52.57 Application for Renewal
The purpose of this section is to set forth the process for
applying for renewal of an existing design certification rule.
Paragraph (a) specifies the time period for submitting an application
for renewal and states that any person can apply for renewal. However,
if the applicant for renewal is not the same person or entity that
applied for the existing design certification, as identified in Section
I of the DCR, then the new applicant is required to demonstrate that
they have the capability to provide the detailed design for that
certified nuclear power plant under Sec. 52.63(c) or Sec. 52.73(b).
Section 52.59 Criteria for Renewal
The purpose of this section is to identify the regulations that
will be used to determine if an existing design certification should be
renewed. Paragraph (a) states that the Commission will grant a request
for renewal if the design complies with the regulations in effect at
the time the certification was originally issued (see Section V of an
existing design certification rule) and imposition of any new safety
requirements on the design during a renewal proceeding will be governed
by the backfit standards in paragraph (b).
Under paragraph (c), the applicant for renewal may request an
amendment to the existing certified design to make some design changes
provided that the new design meets the regulations in effect at the
time that the amended, renewed design certification rule is issued and
the changes do not require a major review or reanalysis of the new
design. If the changes to the original design certification are so
extensive that the NRC concludes an essentially new standard design is
being proposed, then the applicant must submit an application for a new
design certification under Sec. 52.45.
Under paragraph (d), denial by the NRC of a request for renewal of
a design certification does not prevent an applicant from submitting a
new application for certification under Sec. 52.45.
Section 52.61 Duration of Renewal
This section specifies the duration that a renewed design
certification is valid for referencing in a combined license
application.
Section 52.63 Finality of Standard Design Certifications
The purpose of this section is to set forth the process for
amending or backfitting existing design certification rules (DCRs) or
issuing orders to nuclear plants that referenced a DCR. This section
also describes the finality of issue resolution under a design
certification and the process for plant-specific departures from a
certified design. This amendment process places a nuclear plant
designer on the same footing as the Commission or any other member of
the public (see 54 FR 15377, first column, April 18, 1989). Therefore,
it cannot be said that this section makes it easier for a designer to
amend design certification information than for the NRC to backfit the
certified design. The amendment and backfitting process uses the phrase
``certification information'' in order to distinguish the rule language
in the DCRs from the design certification information (e.g., Tier 1 and
Tier 2) that is incorporated by reference in the DCRs.
No matter who proposes it, a generic change under Sec. 52.63(a)(1)
will not be made to a DCR while it is in effect unless the change: (1)
is necessary for compliance with Commission regulations applicable and
in effect at the time the certification was issued; (2) is necessary to
provide adequate protection of the public health and safety or common
defense and security; (3) reduces unnecessary regulatory burden and
maintains protection to public health and safety and common defense and
security; (4) provides the detailed design information necessary to
resolve selected design acceptance criteria; (5) corrects material
errors in the certification information; (6) substantially increases
overall safety, reliability, or security of a facility and the costs of
the change are justified; or (7) contributes to increased
standardization of the certification information.
Paragraphs (a)(1)(i) and (a)(1)(ii) did not change in the final
rule. Paragraph (a)(1)(i) provides the compliance exception to the
NRC's backfit process. Paragraph (a)(1)(ii) sets forth the special
[[Page 49445]]
backfit criteria, which uses the adequate protection standard rather
than the backfit standard in 10 CFR 50.109. The remaining paragraphs
permit amendments of design certification information without meeting
the special backfit requirement in Sec. 52.63(a)(1)(ii).
Paragraph (a)(1)(iii) allows the Commission to change the design
certification rule language to reduce unnecessary regulatory burdens,
i.e., incorporate the revised Sec. 50.59 change criteria, or change
the certification information if the change provides a reduction in
regulatory burden and maintains protection to public health and safety
and common defense and security. Maintaining protection generally
embodies the same safety principles used by the NRC in applying risk-
informed decision-making, i.e., ensuring that adequate protection is
provided, applicable regulations are met, sufficient safety margins are
maintained, defense-in-depth is maintained, and that any changes in
risk are small and consistent with the Commission's Safety Goal Policy
Statement (refer to NRC's RG 1.174).
Paragraph (a)(1)(iv) allows for generic resolutions of design
acceptance criteria (DAC) by amending DCRs. The DAC are a special type
of ITAAC that are used to verify the resolution of design issues where
sufficient design information was not provided in the design
certification application. By generically resolving DAC with the
amendment process, the Commission achieves resolution of additional
design issues, achieves finality for those issue resolutions, and
avoids repetitive consideration of those design issues in individual
combined license proceedings. Also, the amendments will enhance
standardization by further completing the certification information.
The NRC staff will review the amendment application to ensure that the
DAC are met and that the new design information conforms with the
applicable regulations.
Paragraph (a)(1)(v) allows for generic resolutions of material
errors in the certification information. This provision is only to be
used to correct a material error, which is an error that significantly
and adversely affects a design function or analysis conclusion
described in the design control document (certification information).
The Commission wants to correct material errors so that these errors
will not have to be addressed in individual licensing proceedings.
Paragraph (a)(1)(vi) allows for generic amendments of certification
information that will substantially increase the overall safety,
reliability, or security of facility design, construction, or operation
provided that the direct and indirect costs of implementation of the
amendment are justified in view of this increased safety, reliability,
or security. This amendment process will function similar to the
backfitting process in 10 CFR 50.109.
Finally, paragraph (a)(1)(vii) allows for generic amendments that
would increase the standardization of certification information in
referencing applications. The Commission is still committed to
achieving and maintaining the benefits of standardization. Therefore,
the final rule allows for generic amendments of certification
information through this additional process, provided that the
amendment is applied to all plants that reference the DCR. This
paragraph will allow applicants and licensees to request corrections or
changes to certification information through a generic process rather
than through individual licensing actions. In determining whether to
codify a proposed amendment under this paragraph, the Commission will
give special consideration to comments from applicants or licensees who
referenced the DCR regarding whether they want to backfit their plants
with these additional changes.
The process for amending DCRs will be a rulemaking with opportunity
for public comment under paragraph (a)(2). As part of the rulemaking
under Sec. 52.63(a)(1), except for Sec. 52.63(a)(1)(ii), the
Commission will give consideration to whether the benefits justify the
costs for plants that are already licensed or for which an application
for a permit or license is under consideration. The duration of the
amended DCR will be for the same period of time as the original DCR and
have the same expiration date.
Once a DCR is amended by rulemaking, under paragraph (a)(3) the
changes will apply to all future applications referencing the DCR as
well as all current plants referencing the design certification, unless
the change has been rendered ``technically irrelevant'' through other
action taken under paragraphs (a)(4) or (b)(1) of this section. Thus,
standardization is maintained by ensuring that any amendment to a DCR
is imposed upon all nuclear power plants referencing the design
certification rule.
Paragraph (a)(4) sets forth the criteria that must be met before
the Commission can impose new requirements by plant-specific order on a
nuclear plant that references a DCR. Under this paragraph, the
Commission must meet either the compliance or adequate protection
backfit criteria and cite one or more special circumstances as defined
in Sec. 52.7. In addition, the Commission shall consider whether the
special circumstances that justify the plant-specific order outweigh
any decrease in safety that may result from the reduction in
standardization caused by the plant-specific order. This additional
requirement was added to ensure that the benefits of standardization
will be preserved.
Paragraph (a)(5) sets forth the finality of matters that are
resolved as part of a design certification rulemaking. Each of the DCRs
have detailed provisions on the issues that were resolved for that
plant design and detailed processes for changes to and departures from
certification information (refer to Sections VI and VIII of appendices
A, B, C, or D to part 52).
Paragraphs (b)(1) and (b)(2) provide processes for requesting
exemptions and departures from certification information. As part of
its adoption of a two-tiered rule structure (refer to SRM on SECY-90-
377, dated February 15, 1991), the Commission codified detailed
processes for changes to and departures from certification information
in each of the design certification rules (refer to Section VIII of
appendices A, B, C, or D to part 52). The processes for a specific
certified design must be used when requesting exemptions and departures
from certification information.
Paragraph (c) identifies the detailed design information that an
applicant for a combined license must have completed and available for
audit by the NRC. The NRC expects that design certification applicants
(vendors) will have this information available during the review of a
combined license application that references the certified design.
Because a rule certifying a standard plant design does not belong to
the designer (vendor), an applicant for a combined license that
references the DCR could use a vendor other than the applicant that
achieved the design certification. In that situation, the combined
license applicant must acquire the detailed design information
identified in paragraph (c) in order to demonstrate that the new vendor
has the ability to provide the certified design and that the combined
license applicant's design information is consistent with the design
information for the DCR.
Subpart C--Combined Licenses
Section 52.71 Scope of Subpart
This section describes the scope of the requirements in this
subpart. Under this subpart an applicant can request a combined
construction permit and operating license with conditions (combined
license) for a nuclear power
[[Page 49446]]
facility. The combined license is essentially a combination of a
construction permit, which requires consideration and resolution of
many of the issues currently considered at the operating license stage,
and a conditional operating license. Operation is allowed only after
the Commission has made the finding that all acceptance criteria in
ITAAC have been met.
The combined license application could describe a site and a custom
design, or it could reference an early site permit (subpart A of part
52), a standard design certification (subpart B of part 52), a standard
design approval (subpart E of part 52), or a reactor manufactured under
a manufacturing licenses (subpart F of part 52) or a combination
thereof. Although a pre-approved site and certified standard design
need not be referenced for the combined license, maximum efficiency
will result if site-related issues, as well as design-related issues,
have been resolved before commencement of the combined license
proceeding.
Section 52.73 Relationship to Other Subparts
The purpose of this section is to explain the relationship of the
combined license process to the licensing processes in subparts A, B,
E, and F of 10 CFR part 52.
Section 52.75 Filing of Applications
This section explains who can file, how to file, and the fees for
NRC review of an application for a combined license.
Section 52.77 Contents of Applications; General Information
This section sets forth the type of general information that is
required to be included in an combined license application, namely, the
information required by 10 CFR 50.33. Section 50.33 requires that the
application include information such as the name and address of the
applicant, a description of the business or occupation of the
applicant, citizenship information of the applicant, the class of
license applied for, the use to which the facility will be put, the
time for which the license is sought, financial qualification
information, State and local emergency response plans, the earliest and
latest dates for the completion of construction, and information about
decommissioning funding. Section 50.33 also provides requirements for
the handling of Restricted Data or other defense information in an
application.
Section 52.79 Contents of Applications; Technical Information in Final
Safety Analysis Report
The purpose of this section is to identify specific technical
information to be included in the final safety analysis report as part
of an application for a combined license. This generally includes the
same information required of applicants for construction permits and
operating licenses under 10 CFR part 50.
This section specifies the complete set of FSAR information needed
for a combined license that is a stand-alone application, but also
takes into account that certain information may already have been
submitted and reviewed in those instances where the application
references an early site permit (subpart A), a certified design
(subpart B), a standard design approval (subpart E), a manufacturing
license (subpart F), or some combination. The required FSAR information
also includes requirements for descriptions of operational programs
that need to be included in the FSAR to allow a reasonable assurance
finding of acceptability. These additional requirements are in support
of the Commission's direction to the staff in SRM-SECY-02-0067 dated
September 11, 2002, ``Inspections, Tests, Analyses, and Acceptance
Criteria for Operational Programs (Programmatic ITAAC),'' that a
combined license applicant was not required to have ITAAC for
operational programs if the applicant fully described the operational
program and its implementation in the combined license application. In
this SRM, the Commission stated:
[a]n ITAAC for a program should not be necessary if the program and
its implementation are fully described in the application and found
to be acceptable by the NRC at the COL stage. The burden is on the
applicant to provide the necessary and sufficient programmatic
information for approval of the COL without ITAAC.
The Commission clarified its definition of fully described in SRM-
SECY-04-0032, ``Programmatic Information Needed for Approval of a
Combined License Application Without Inspections, Tests, Analyses, and
Acceptance Criteria,'' dated May 14, 2004, as follows:
In this context, fully described should be understood to mean
that the program is clearly and sufficiently described in terms of
the scope and level of detail to allow a reasonable assurance
finding of acceptability. Required programs should always be
described at a functional level and at an increased level of detail
where implementation choices could materially and negatively affect
the program effectiveness and acceptability.
Accordingly, this section contains requirements for descriptions of
operational programs and their implementation.
Paragraph (b) describes the information that is needed if the
application references an early site permit. Although a combined
license applicant referencing a certified design need not resubmit
information or analyses submitted in connection with the early site
permit, the combined license application FSARs must either include or
incorporate by reference the SSAR for the early site permit. The SSAR
must be included or incorporated into the combined license FSAR to
ensure that matters addressed in the SSAR legally become part of the
FSAR upon issuance of the combined license. This will also ensure that
the information in the SSAR is subject to control under Sec. 50.59
after issuance of the combined license. This provision is meant to
convey that the combined license applicant referencing the early site
permit does not need to resubmit, for NRC review, information or
analyses that were already reviewed and resolved in the early site
permit proceeding (such as information provided in responses to NRC
requests for additional information). At the same time, this provision
provides combined license applicants guidance as to what the combined
license application must contain to be considered complete, including a
requirement that it contain or incorporate the early site permit SSAR.
Because an early site permit applicant need not specify a
particular nuclear plant design, the combined license application must
demonstrate that the design of the facility falls within the site
characteristics and postulated design parameters specified in the early
site permit. If the application does not demonstrate that design of the
facility falls within the site characteristics and design parameters of
the early site permit, then, the applicant must request for a variance
from the early site permit. Paragraph (b) requires that the application
demonstrate that all terms and conditions in the early site permit,
excluding terms and conditions imposed under Sec. 50.36b, be satisfied
by the date of issuance of the combined license. Any terms or
conditions of the early site permit that could not be met by the time
of issuance of the combined license must be set forth as terms or
conditions of the combined license. Early site permit conditions
imposed under Sec. 50.36b are to be addressed in the environmental
report and not in the FSAR.
[[Page 49447]]
Paragraph (b) also addresses emergency planning information
submitted in a referenced early site permit and requires that the
combined license application include any new or additional information
to update or correct information provided with the early site permit
and to discuss whether the new information may materially change the
bases for compliance with the applicable NRC requirements. New
information which materially changes the bases for compliance includes:
(1) information which substantially alters the bases for a previous NRC
conclusion with respect to the acceptability of a material aspect of
emergency preparedness or an emergency preparedness plan, as well as
(2) information which would constitute a sufficient basis for the
Commission to modify or impose new terms and conditions related to
emergency preparedness in accordance with Sec. 52.39(a)(1). New
information that substantially alters the bases for a previous NRC
conclusion or constitutes a sufficient basis for Commission to modify
or impose new terms and conditions related to emergency preparedness
would be subject to litigation during the combined license proceeding
in accordance with Sec. 52.39(c). This paragraph also addresses
referenced early site permit emergency plans that incorporate existing
emergency plans and requires the combined license application to
identify changes to the emergency plans that constitute a decrease in
effectiveness under 10 CFR 50.54(q). This requirement ensures that the
NRC can review such changes to assess their impact on the emergency
plans for the proposed combined license facility.
Paragraph (c) and (d) provide application requirements for a
combined license that is referencing a standard design approval or a
standard design certification, respectively. Similar to a combined
license application referencing an early site permit, a combined
license application referencing a design approval or design
certification must either include or incorporate by reference the
design approval or design certification FSAR. Because a design approval
or design certification applicant need not specify a particular site,
the combined license application must demonstrate that characteristics
of the site fall within the site parameters specified in the design
approval or design certification. In addition, the plant-specific PRA
information must use the PRA information for the design certification
and must be updated to account for site-specific design information and
any design changes or departures. An applicant referencing a design
certification must demonstrate that the interface requirements
established for the design have been met. Applicants referencing either
a design approval or a design certification must demonstrate that any
terms and conditions in the design approval or requirements and
restrictions in the referenced design certification rule will be
satisfied by the date that the combined license is issued. Any terms or
conditions of the design approval that cannot be met or satisfied by
the time of issuance of the combined license must be set forth as terms
or conditions of the combined license. Likewise, any requirements or
restrictions of the design certification that cannot be met or
satisfied by the time of issuance of the combined license must be set
forth as terms or conditions of the combined license.
Paragraph (e) describes the information that is needed if the
combined license application references one or more manufactured
reactors. Similar to a combined license application referencing an
early site permit, design approval, or design certification, a combined
license application referencing one or more manufactured nuclear power
reactors under subpart F or part 52 must either include or incorporate
by reference the manufacturing license FSAR. Because a manufacturing
license applicant need not specify a particular site for the
installation of a manufactured reactor, the combined license
application must demonstrate that the site parameters for the
manufactured reactor are bounded by the site where the manufactured
reactor is to be installed and used. In addition, the plant-specific
PRA information must use the PRA information for the manufactured
reactor and must be updated to account for site-specific design
information and any design changes or departures. The combined license
application must also demonstrate that the interface requirements
established for the design have been met and that any terms and
conditions in the manufacturing license will be satisfied by the date
that the combined license is issued. Any terms or conditions of the
manufacturing license that could not be met by the time of issuance of
the combined license must be set forth as terms or conditions of the
combined license.
Section 52.80 Contents of Applications; Additional Technical
Information
This section covers the required technical contents of a combined
license application that are not contained in the FSAR. These
application contents include the proposed ITAAC, the environmental
report, and information to address an applicant's request to perform
activities at the site allowed by 10 CFR 50.10(e) before issuance of
the combined license.
Paragraph (a) requires the application to include the proposed
ITAAC and, if the application references an early site permit with
ITAAC or a design certification, requires the applicant to use the
ITAAC contained in the early site permit or design certification for
the applicable portion of the combined license application. ITAAC that
must be included are those that are necessary and sufficient to
demonstrate that the facility has been constructed and will be operated
in conformity with the combined license, the provisions of the Atomic
Energy Act of 1954 and the Commission's rules and regulations. In
addition, under Section 52.103(g), the Commission must find that all
acceptance criteria specified in the license are met before facility
operation. Because ITAAC are the sole source of acceptance criteria for
subsequent resolution of items which cannot be fully evaluated prior to
issuance of a combined license, it is essential that the combined
license ITAAC include all significant issues that require satisfactory
resolution before fuel loading.
This paragraph also provides an applicant for a combined license
with a process for resolving certain acceptance criteria in one or more
of the ITAAC before issuance of the combined license. This provision is
included mainly to allow for completion of DAC at the combined license
application stage because applicants might want to complete certain DAC
before construction. DAC are special design certification rule ITAAC.
DAC set forth processes and criteria for completing certain design
information, such as information about the digital instrumentation and
control system. Many DAC were originally written to be verified as part
of the normal, post-combined license, ITAAC verification process.
Completion of the design matters covered by DAC before the issuance of
a combined license is consistent with the Commission's original concept
for design certification and issuance of a combined license. When it
adopted 10 CFR part 52, the Commission intended that a design
certification contain final and complete design information. Allowing a
finding of acceptable completion of DAC before issuance of a combined
license is, therefore, consistent with the
[[Page 49448]]
Commission's original intent. Second, completion of DAC before issuance
of the combined license is consistent with the Commission's goal of
resolving issues before construction. Determining whether DAC have been
successfully completed before issuance of the combined license avoids
the possibility that improperly completed DAC will result in the
construction of improperly designed structures, systems, and
components. Accordingly, a finding of successful completion of DAC may
be made when a combined license is issued, if the combined license
applicant demonstrates that the DAC have been successfully completed.
This process would also allow findings on successful completion of
inspections or tests of components procured before the issuance of the
combined license.
Paragraph (b) requires a complete environmental report in
accordance with 10 CFR 51.50(c).
Paragraph (c) requires that, if the applicant is requesting to
perform any activities at the site allowed by 10 CFR 50.10(e), then the
applicant must identify and describe the activities and propose a plan
for redress of the site in the event that the activities are performed
and either construction is abandoned or the combined license is
revoked. This paragraph also requires the applicant to demonstrate that
there is reasonable assurance that redress carried out under the plan
will achieve an environmentally stable and aesthetically acceptable
site suitable for whatever non-nuclear use may conform with local
zoning laws. These requirements attempt to limit, to the extent
practicable, the environmental impact of any site work done in the case
where construction of the nuclear power facility is not completed.
Section 52.81 Standards for Review of Applications
This section identifies the regulations that the NRC staff will use
in performing its review of an application for a combined license.
Section 52.83 Finality of Referenced NRC Approvals; Partial Initial
Decision of Site Suitability
This section describes the finality of regulatory products that may
be referenced in a combined license application. Specifically,
paragraph (a) states that the finality of matters resolved in a
referenced early site permit, design certification, design approval, or
manufacturing license are governed by the finality provisions in the
respective subparts that address each of these regulatory processes.
Paragraph (b) states that, while a partial decision on site suitability
is in effect under 10 CFR 2.617(b)(2), the finality provisions in 10
CFR 2.629 govern the scope and nature of matters resolved in the
proceeding.
Section 52.85 Administrative Review of Applications; Hearings
This section identifies the procedural requirements that apply to
the mandatory combined license hearing. This section also identifies
that, if an applicant requests a Commission finding on certain ITAAC
with the issuance of the combined license, then those ITAAC will be
identified in the notice of hearing.
Section 52.87 Referral to the Advisory Committee on Reactor Safeguards
(ACRS)
This section states that the ACRS will report on those portions of
the application which concern safety.
Section 52.91 Authorization To Conduct Site Activities
The purpose of this section is to outline the activities that can
be performed at the site by a combined license applicant. Paragraph (a)
of this section discusses the authorization a combined license
applicant needs to obtain in order to perform limited work activities
at the site while the NRC is considering the combined license
application in the case where a combined license applicant does not
reference an early site permit that contains a redress plan. The
requirements contained in paragraph (a) discuss work commonly referred
to as a limited work authorization 1 (LWA-1) that is allowed in
accordance with the requirements contained in 10 CFR 50.10(e)(1). These
requirements do not allow the applicant to perform LWA-1 activities
without first submitting a redress plan and obtaining the separate
authorization required by 10 CFR 50.10(e)(1). Plans are expected to be
modeled on the Midland Site Stabilization Report that was submitted on
October 2, 1986 (ML061710504).
Paragraph (a) recognizes this possibility and notes that
authorization may be granted only after the presiding officer in the
proceeding on the application has made the findings and determination
required by 10 CFR 50.10(e)(2) and has determined that redress carried
out under the site redress plan will return the site to an
aesthetically acceptable and environmentally stable condition.
Paragraph (b) contains requirements for work commonly referred to
as an LWA-2. An LWA-2 allows structural work for structures, systems,
and components which prevent or mitigate the consequences of postulated
accidents that could cause undue risk to the health and safety of the
public. Because the design must be known to obtain authorization for
LWA-2 activities, an LWA-2 is an option for a combined license
applicant but not an option for an early site permit holder. A combined
license applicant may request LWA-2 authority prior to the combined
license being granted. Paragraph (b) recognizes this possibility and
notes that authorization may be granted only after the presiding
officer in the combined license makes the additional finding required
by 10 CFR 50.10(e)(3)(ii), namely, that there are no unresolved safety
issues relating to the LWA-2 activities.
Paragraph (c) of this section clarifies that, if work is performed
either under an LWA-1, or LWA-2 or both, and the combined license
application is subsequently withdrawn by the applicant or denied by the
NRC, then the combined license applicant must redress the site in
accordance with the terms of the site redress plan. Paragraph (c) of
this section also provides the combined license applicant with the
ability to redress the site for an alternate use that was not
considered at the time that the original redress plan was prepared.
Section 52.93 Exemptions and Variances
The purpose of this section is to describe the process for combined
license applicants to obtain exemptions and variances. If the request
is for an exemption from any part of a referenced design certification
rule, the Commission can grant the request only if it determines that
the exemption complies with any exemption provisions in the referenced
design certification rule, or with Sec. 52.63 if there are no
applicable exemption provisions in the referenced design certification
rule. A request for an exemption that is outside the scope of a design
certification rule must be processed in accordance with the
requirements contained in Sec. 52.7.
For the General Electric ABWR, Westinghouse System 80+,
Westinghouse AP600, and Westinghouse AP1000 designs, these requirements
are contained in Section VIII, ``Processes for Changes and
Departures,'' of appendices A, B, C, and D respectively, of 10 CFR part
52. Section VIII of these appendices discusses the process for
exemptions from different portions of the design certification rule.
The section-by-section analysis for these respective rules
[[Page 49449]]
discuss requirements regarding processing of exemptions that are
expected to be carried forward to future design certification
rulemakings. Therefore, if applicable, the applicant should refer to
the respective section-by-section analysis in the portion of the design
certification rule that discusses exemptions for additional
information. Exemptions requested in accordance with this section are
subject to litigation in the same manner as other issues in the
licensee hearing.
Paragraph (b) of this section sets forth the process for requesting
variances from an early site permit if one is referenced in the
combined license. Paragraph (c) sets forth the process for requesting
variances from one or more design characteristics, site parameters,
terms and conditions, or approved design of a manufactured reactor.
Issuance of a variance is subject to litigation during the combined
license proceeding in the same manner as other issues material to that
proceeding.
Section 52.97 Issuance of Combined Licenses
The purpose of this section is to set forth the process for issuing
a combined license. Paragraph (a)(1) of this section sets forth the
requirements relative to the Commission findings that must be made for
granting of a combined license.
Paragraph (a)(2) of this section allows for completion of certain
acceptance criteria in one or more of the ITAAC in a combined license
being met prior to granting of the combined license. This paragraph
could apply to DAC found in the applicable design certification rules.
DAC set forth processes and criteria for completing certain design
information, such as information about the digital instrumentation and
control system. Paragraph (a)(2) would allow the Commission to make a
finding of successful completion of DAC when a combined license is
issued, if the combined license applicant demonstrates that the DAC
have been successfully completed. This process would also allow
findings on successful completion of inspections or tests of components
procured before the issuance of a combined license. Paragraph (a)(2)
notes that such a finding will preclude any required finding under
Sec. 52.103(g) with respect to that ITAAC.
Paragraph (b) requires the Commission to identify the ITAAC within
the combined license that the licensee shall perform, and the
acceptance criteria that, if met, are necessary and sufficient to
provide reasonable assurance that the facility has been constructed and
will be operated in conformity with the license, the provisions of the
Act, and the Commission's rules and regulations. This definition of
what ITAAC are intended to accomplish is consistent with that contained
in Sec. 52.17 regarding early site permits, Sec. 52.47 regarding
design certifications and Sec. 52.80, which are discussed above. If
the combined license application references an early site permit with
ITAAC related to emergency planning information, then the applicant
must use these ITAAC in the emergency planning information submitted
with the combined license application. If a combined license applicant
references a design certification rule, the ITAAC contained in the
license would be those contained in the design certification rule plus
any additional ITAAC that were identified during the combined license
review that were outside the scope of the certified design. If the
Commission wishes to identify additional ITAAC that fall within the
scope of the review of the referenced certified design it needs to meet
the requirements contained in the design certification rule itself (see
Section VIII.A.3 of appendix A, B, C, and D for the ABWR, System 80+,
AP600, and AP1000) and the requirements contained in Sec. 52.63. If a
combined license applicant does not reference an early site permit or a
certified design, then the ITAAC that are identified by the Commission
for paragraph (b) of this section are those that were identified during
the combined license review.
Section 52.98 Finality of Combined Licenses; Information Requests
This section covers the finality of combined license provisions and
sets forth the requirements to modify the combined license after it has
been issued. After issuance of a combined license, the Commission may
not modify, add, or delete any term or condition of the combined
license, the design of the facility, the inspections, tests, analyses,
and acceptance criteria contained in the license which are not derived
from a referenced standard design certification or manufacturing
license, except in accordance with the backfit provisions of Sec. Sec.
52.103 or 50.109, as applicable.
Paragraphs (b), (c), and (d) outline the applicability of the
change processes in 10 CFR part 50, Section VIII of the design
certification rules, and subpart F of 10 CFR part 52 to a combined
license. The change processes in 10 CFR part 50 apply to a combined
license that does not reference a design certification rule or a
reactor manufactured under a manufacturing license. Section 52.98(c)
states that the change processes in Section VIII of the design
certification rules apply to changes within the scope of the referenced
certified design. However, if the proposed change affects the design
information that is outside of the scope of the design certification
rule, the part 50 change processes apply unless the change also affects
the design certification information. For that situation, both change
processes may apply. If the combined license references a reactor
manufactured under a subpart F manufacturing license, then changes to
or variances from information within the scope of the manufactured
reactor's design are subject to the change processes in Sec. 52.171.
Paragraph (e) was added in 1992, and discussed in the section-by-
section analysis (57 FR 60976; December 23, 1992), as following:
This section has been amended with regard to making amendments
to a combined license immediately effective under the so-called
``Sholly Amendment.'' Under the Energy Policy Act, an amendment to a
combined license can be made immediately effective if the Commission
determines there are no significant hazards considerations. This
section of the rule has been revised to incorporate the statutory
provisions and previously issued Commission regulations implementing
the ``Sholly'' amendment. The Commission, however, stresses that it
will not look with favor upon license amendments to a combined
license filed shortly before planned operation that could have the
effect of undermining standardization or changing the scope of
imminent or pending hearings on conformance issues.
Paragraph (f) states that any modification to a combined license is
an amendment to the license and that there must be an opportunity for
hearing on these amendments. Such amendments would be processed in
accordance with the requirements contained in 10 CFR 50.90 and 50.91.
In addition, if the applicant has referenced a certified design, or a
reactor manufactured under a manufacturing license, additional
requirements may apply. For example, a combined license that references
an ABWR certified design may request an exemption from Tier 1 material
in accordance with the provisions contained in Section VIII.A.4 of
appendix A of 10 CFR part 52. In such a case, the licensee would have
to process an exemption in accordance with the requirements contained
in appendix A to part 52 and 10 CFR 52.63(b)(1) and a license amendment
in accordance with paragraph (f) of this section.
Paragraph (g) which is analogous to Sec. Sec. 52.39(f), 52.145(c),
and 52.171(c),
[[Page 49450]]
provides that NRC information requests must be evaluated before
issuance to ensure that the burden to be imposed by the information
request is justified in view of the potential safety significance of
the issue to be addressed, except when the information requests seeks
to verify compliance with the current licensing basis of the combined
license. Information requests may be in the form of a new rule
requiring submission of information (i.e., a new information collection
and reporting requirement), or in the form of a NRC staff request for
information. Information requests by the staff must be in accordance
with 10 CFR 50.54(f) and must be approved by the EDO or his or her
designee before the request may be issued.
Section 52.99 Inspection During Construction
The purpose of this section is to set forth the requirements to
support the NRC's inspections during construction. A new Sec. 52.99(a)
has been added to require that the licensee submit to the NRC, no later
than 1 year after issuance of the combined license or at the start of
construction as defined in 10 CFR 50.10, whichever is later, its
schedule for completing the inspections, tests, or analyses in the
ITAAC. This provision also requires the licensee to submit updates to
the ITAAC schedule every 6 months thereafter and, within 1 year of its
scheduled date for initial loading of fuel, licensees must submit
updates to the ITAAC schedule every 30 days until the final
notification is provided to the NRC under Sec. 52.99(c). The
information provided by the licensee will be used by NRC in developing
the NRC's inspection activities and activities necessary to support the
Commission's finding whether all of the ITAAC have been met prior to
the licensee's scheduled date for fuel load. Even in the case where
there were no changes to a licensee's ITAAC schedule during an update
cycle, the NRC expect the licensee to notify the NRC that there have
been no changes to the schedule.
Section 52.99 has also been amended to incorporate rule language
from the design certification rules in 10 CFR part 52 regarding the
completion of ITAAC (see paragraphs IX.A and IX.B.3 of appendix A to
part 52). During the preparation of the design certification rules for
the ABWR and System 80+ designs, the NRC staff and nuclear industry
representatives agreed on certain requirements for the performance and
completion of the inspections, tests, or analyses in ITAAC. In the
design certification rulemakings, the Commission codified these ITAAC
requirements into Section IX of the regulations. The purpose of the
requirement in Sec. 52.99(b) is to clarify that an applicant may
proceed at its own risk with design and procurement activities subject
to ITAAC, and that a licensee may proceed at its own risk with design,
procurement, construction, and preoperational testing activities
subject to an ITAAC, even though the NRC may not have found that any
particular ITAAC has been met.
Section 52.99(c)(1) requires the licensee to notify the NRC that
the prescribed inspections, tests, and analyses have been performed and
that the prescribed acceptance criteria have been met. Section
52.99(c)(1) further requires that the notification contain sufficient
information to demonstrate that the prescribed inspections, tests, and
analyses have been performed and that the prescribed acceptance
criteria have been met.
Section 52.99(c)(2) requires that, if the licensee has not
provided, by the date 225 days before the scheduled date for initial
loading of fuel, the notification required by paragraph (c)(1) of this
section for all ITAAC, then the licensee shall notify the NRC that the
prescribed inspections, tests, or analyses for all uncompleted ITAAC
will be performed and that the prescribed acceptance criteria will be
met prior to operation (consistent with the Section 185.b requirement
that the Commission, ``prior to operation,'' find that the acceptance
criteria in the combined license are met). The notification must be
provided no later than the date 225 days before the scheduled date for
initial loading of fuel, and must provide sufficient information to
demonstrate that the prescribed inspections, tests, or analyses will be
performed and the prescribed acceptance criteria for the uncompleted
ITAAC will be met.
Section 52.99(c) ensures that: (1) The NRC has sufficient
information to complete all of the activities necessary for the
Commission to make a determination as to whether all of the ITAAC have
been or will be met prior to initial operation; and (2) interested
persons will have access to information on both completed and
uncompleted ITAAC at a level of detail sufficient to address the AEA
Section 189.a(1)(B) threshold for requesting a hearing on acceptance
criteria. It is the licensee's burden to demonstrate compliance with
the ITAAC and the NRC expects the information submitted under paragraph
(c)(1) to contain more than just a simple statement that the licensee
believes the ITAAC has been completed and the acceptance criteria met.
The NRC expects the notification to be sufficiently complete and
detailed for a reasonable person to understand the bases for the
licensee's representation that the inspections, tests, and analyses
have been successfully completed and the acceptance criteria have been
met. The term ``sufficient information'' requires, at a minimum, a
summary description of the bases for the licensee's conclusion that the
inspections, tests, or analyses have been performed and that the
prescribed acceptance criteria have been met. Furthermore, with respect
to uncompleted ITAAC, it is the licensee's burden to demonstrate that
it will comply with the ITAAC and the NRC expects the information that
the licensee submits under paragraph (c)(2) to be sufficiently detailed
such that the NRC can determine what activities it will need to
undertake to determine if the acceptance criteria for each of the
uncompleted ITAAC have been met, once the licensee notifies the NRC
that those ITAAC have been successfully completed and their acceptance
criteria met. The term ``sufficient information'' requires, at a
minimum, a summary description of the bases for the licensee's
conclusion that the inspections, tests, or analyses will be performed
and that the prescribed acceptance criteria will be met. In addition,
``sufficient information'' includes, but is not limited to, a
description of the specific procedures and analytical methods to be
used for performing the inspections, tests, and analyses and
determining that the acceptance criteria have been met.
The NRC notes that, even though it did not include a provision
requiring the completion of all ITAAC by a certain time prior to the
licensee's scheduled fuel load date, the NRC staff will require some
period of time to perform its review of the last ITAAC once the
licensee submits its notification that the ITAAC has been successfully
completed and the acceptance criteria met. In addition, the Commission
itself will require some period of time to perform its review of the
staff's conclusions regarding all of the ITAAC and the staff's
recommendations regarding the Commission finding under Sec. 52.103(g).
Therefore, licensees should structure their construction schedules to
take into account these time periods.
A new paragraph (d) states the options that a licensee will have in
the event that it is determined that any of the acceptance criteria in
the ITAAC have not been met. If an activity is subject to an ITAAC
derived from a referenced standard design certification and the
licensee has not demonstrated that the ITAAC has been met, the licensee
may take corrective actions to
[[Page 49451]]
successfully complete that ITAAC or request an exemption from the
standard design certification ITAAC, as applicable. A request for an
exemption must also be accompanied by a request for a license amendment
under Sec. 52.98(f). Also, if an activity that is subject to an ITAAC
is not derived from a referenced standard design certification and the
licensee has not demonstrated that the ITAAC has been met, the licensee
may take corrective actions to successfully complete that ITAAC or
request a license amendment under Sec. 52.98(f).
Paragraph (e)(1) of this section indicates that the NRC is
responsible for ensuring (through its inspection and audit activities)
that the combined license holder performs and documents the completion
of inspections, tests, and analyses in the ITAAC. When part 52 was
first adopted by the Commission in 1989 (April 18, 1989; 54 FR 15372),
the rule provided that the NRC staff shall ensure that the inspections,
tests, and analyses in the ITAAC are performed, and did not refer to
the Commission finding on acceptance criteria being met. The Commission
revised the language in this portion of the rule in 1992 (December 23,
1992; 57 FR 60975) to reflect changes to Section 185 of the AEA made by
Congress in the Energy Policy Act of 1992 (1992 EPA), which states:
Following issuance of the combined license, the Commission shall
ensure that the prescribed inspections, tests, and analyses are
performed and, prior to operation of the facility, shall find that
the prescribed acceptance criteria are met.
Thus, the revisions to this portion of the rule in 1992 simply
reflected the language of the 1992 EPA. However, the Commission does
not believe that Congress, by adopting language in Section 185 stating
that the Commission shall ensure that the ITAAC are performed, intended
to prohibit the Commission's long-standing practice of delegating to
the NRC staff the responsibility for performing the necessary
activities, including audits and inspections, to ensure that ``the
required inspections, tests, and analyses in the ITAAC are performed.''
Accordingly, the language from the 1992 rule change is retained in this
final rule.
Paragraph (e)(1) requires the NRC to publish, at appropriate
intervals until the last date for submission of requests for hearing
under Sec. 52.103(a), notices in the Federal Register of the NRC
staff's determination of the successful completion of inspections,
tests, and analyses. Paragraph (e)(2) provides that the NRC shall make
publicly available the licensee notifications under paragraphs (c)(1)
and (c)(2). In general, the NRC expects to make the paragraph (c)(1)
notifications availability shortly after the NRC has received the
notifications and concluded that they are complete and detailed.
Furthermore, by the date of the Federal Register notice of intended
operation and opportunity to request a hearing on whether acceptance
criteria have been or will be met (under Sec. 52.103(a)), the NRC will
make available the notifications under paragraph (c)(2), and the
notifications under paragraph (c)(2) for all ITAAC for which paragraph
(c)(1) notifications have not been provided by the licensee.
Section 52.103 Operation Under a Combined License
The purpose of this section is to set forth the requirements for
operation under a combined license. This section has been previously
discussed in a section-by-section analysis for the 1992 revisions to
part 52 (57 FR 60976; December 23, 1992) which the NRC adopted in
response to the Energy Policy Act of 1992. The 1992 section-by-section
analysis states:
In an effort to adhere as closely as possible to the new
statutory requirements of the Energy Policy Act, the NRC has
replaced most of its old Sec. 52.103 with the text of section 2802
of that Act. Under the revised language, any request for a post-
construction hearing must show, prima facie, both that one or more
of the acceptance criteria are not or will not be met, and those
specific operational consequences of nonconformance that would be
contrary to providing reasonable assurance that the public health
and safety will be adequately protected. The Commission may permit
interim operation of a facility pending a hearing if it determines
that this assurance exists. The Commission has the discretion to
decide if any post-construction hearing will use formal or informal
hearing procedures, and it must state publicly the reasons for
choosing either set of procedures. The Commission must find, prior
to operation of the facility, that the acceptance criteria have been
met.
Paragraph (a) of this section is revised to require licensees to
notify the NRC of its schedule date for initial loading of fuel no
later than 270 days before the scheduled date and to notify the NRC of
updates to its schedule every 30 days thereafter. This information will
be used by the NRC to develop the notice of intended operation in the
Federal Register, which must be published not less than 180 days before
the licensee's initial fuel load date, as required by Section
189.a.(1)(B) of the AEA. In addition, paragraph (a) addresses the
possibility that an applicant for a combined license may choose to
resolve certain acceptance criteria in one or more of the ITAAC
required by Sec. 52.80 before issuance of the combined license. In
such a case, if the Commission makes a finding in accordance with Sec.
52.97 associated with these ITAAC at the time that a combined license
is granted, these ITAAC would not be subjected to a hearing opportunity
again under paragraph (a) of this section. The section-by-section
analysis for Sec. 52.97 discusses this issue in more detail.
Paragraph (b) provides the criteria that must be met for any
request for a hearing on whether the facility complies or will comply
with the acceptance criteria. The petitioner must set forth with
reasonable specificity the facts and arguments which form the basis for
the request. These provisions are designed to accord finality to the
Commission's earlier decisions regarding the facility and to ensure
that any proceeding is focused on significant safety issues.
Paragraph (c) requires the Commission to expeditiously either deny
or grant any request for a hearing under this section. If a request is
granted, the Commission must determine whether to allow interim
operation of the facility based on reasonable assurance of adequate
protection of the public health and safety.
Paragraph (d) provides that the Commission will determine the
appropriate hearing procedures in accordance with 10 CFR part 2 for any
hearing under paragraph (a) of this section. Under Sec. 2.309, as
adopted by the Commission in 2004 (69 FR 2182; January 14, 2004), such
a hearing would ordinarily be conducted under subpart L of part 2.
However, the Commission may direct, in the notice of required by
paragraph (a) or in a subsequent order, that any hearing that may be
conducted in a particular combined license proceeding under paragraph
(a) use other, less formal hearing procedures, consistent with the
requirements of the AEA. Any such Commission direction is consistent
with the Commission's statement in the SOC for the 1989 final part 52
rulemaking (54 FR 15372, 15383; April 18, 1989) that any hearing held
under former Sec. 52.103(b)(2)(i) (Sec. 52.103(b) in this final rule)
will use informal procedures to the maximum extent practical and
permissible under law.
Paragraph (e) states that the Commission will, to the maximum
extent possible, render a decision on issues raised in any hearing
request within 180 days of the publication of the notice or by the
anticipated date for initial fuel load, whichever is later.
[[Page 49452]]
Paragraph (f) provides requirements related to the submittal of
petitions to modify the terms and conditions of a combined license and
states that fuel loading and operation under a combined license will
not be affected by the granting of a petition unless the Commission
makes an order immediately effective.
Paragraph (g) prohibits the licensee from operating the facility
until the Commission makes a finding that the acceptance criteria in
the combined license are met (except for acceptance criteria that the
Commission found were met when the combined license was issued). The
NRC believes that the rule should reflect, as closely as possible, the
statutory requirement in Section 185.b of the AEA. Although the NRC has
historically viewed ``operation'' as including loading of fuel into the
reactor, the NRC believes it is not necessary to change the language of
Sec. 52.103(g) to continue the historical practice.
Paragraph (h) of this section incorporates rule language from the
design certification rules in 10 CFR part 52 regarding the completion
of ITAAC (see paragraphs IX.A and IX.B.3 of appendix A to part 52).
This paragraph states that ITAAC do not, by virtue of their inclusion
in the design certification rule or combined license, constitute
regulatory requirements after the licensee has received authorization
to load fuel or for any renewal of the license. However, subsequent
modifications to the facility or procedures described in the FSAR must
comply with the requirements in Sec. 52.98.
Section 52.104 Duration of Combined License
This section addresses the duration of a combined license which is
a period not to exceed 40 years from the date that the Commission makes
the finding that the acceptance criteria in the license are met, in
accordance with Sec. 52.103(g). Where the Commission has allowed
operation during an interim period under Sec. 52.103(c), the period of
operation is not to exceed 40 years from the date allowing operation
during the interim period. This provision implements Section 621 of the
Energy Policy Act of 2005 which amended Section 103c. of the AEA. The
AEA provided that the 40 year duration started on the date that the
Commission authorized construction of the facility (i.e., the date of
issuance of the combined license).
Section 52.105 Transfer of Combined License
This section states that a combined license may by transferred in
accordance with 10 CFR 50.80, ``Transfer of licenses.'' Section 50.80
provides the requirements regarding application for a license transfer.
All license transfers must be approved by the Commission.
Section 52.107 Application for Renewal
This section states that an application to renew a combined license
must be in accordance with 10 CFR part 54, ``Requirements for Renewal
of Operating Licenses for Nuclear Power Plants.''
Section 52.109 Continuation of Combined License
This section, which is analogous to Sec. 50.51, provides
requirements for a combined license facility that has permanently
ceased operations and states that the license continues in effect
beyond the expiration date until the Commission notifies the licensee
in writing that the license is terminated. During this period, the
licensee is required to decommission and decontaminate the facility;
maintain the facility, including the spent fuel, in a safe condition;
and continue to follow the NRC's regulations and the provisions of the
combined license.
Section 52.110 Termination of License
This section, which is analogous to Sec. 50.82, provides
requirements the termination of a combined license. These provisions
include a requirement to notify the NRC within 30 days when a licensee
has decided to permanently cease operations and to submit a
certification to the NRC once fuel has been permanently removed from
the reactor vessel. This section also requires decommissioning of the
facility within 60 years of permanent cessation of operations and
outlines requirements regarding decommissioning activities.
Subpart E--Standard Design Approvals
Section 52.131 Scope of Subpart
This section describes the scope of this process for design
approvals of standard nuclear power plants or major portions thereof,
i.e., a nuclear steam supply system or balance of plant. Under this
subpart an applicant may request pre-approval of a standard nuclear
power plant design, separate from a site review or other licensing
action, and subsequently have that design approval referenced in an
application to build a nuclear power plant. This licensing process was
first adopted by the Commission in 1975 and has been used many times.
Section 52.133 Relationship to Other Subparts
The purpose of this section is to explain the relationship of the
standard design approval process to the processes set forth in subparts
B, C, and F of 10 CFR part 52, which provide for design certifications,
combined licenses, and manufacturing licenses. The Commission continues
to believe that the best approach for obtaining early resolution of
design issues is through the design certification process in subpart B
of this part. Applicants for a design approval have the option of also
applying for design certification. Applicants for a combined license or
a manufacturing license may reference a design approval.
Section 52.135 Filing of Applications
This section explains how to file an application for a standard
design approval and how the fees for NRC's review of the application
will be assessed. Applications are limited to final design information,
in order to remove the unpredictability of issuing a construction
permit that references only preliminary design information and
initiating construction while the final design information is being
completed. Approval of a final standard design ensures early
consideration and resolution of technical matters by the NRC staff
before there is any substantial commitment of resources, which will
greatly enhance regulatory stability and predictability.
Section 52.136 Contents of Applications; General Information
This section identifies the general information that must be
included in all applications.
Section 52.137 Contents of Applications; Technical Information
The purpose of this section is to identify the technical
information that must be included in an application for a design
approval. Paragraphs (a) and (c) describe information that must be
included in the FSAR, which is included in the application, and
paragraph (b) describes the information that must also be included in
the application but does not need to be included in the FSAR.
Applications for a major portion of the plant design, such as the
nuclear steam supply system, only need to contain the technical
information that is applicable to the major portion of the plant for
which NRC staff approval is requested.
[[Page 49453]]
Many of the requirements in this section were taken from 10 CFR
50.34 or are pointers to technical requirements in parts 20, 50, and 73
that must be addressed in the application. The requirements taken from
Sec. 50.34 are a subset of the information required of applicants for
construction permits and operating licenses. Other requirements came
from appendix O to part 50 or were created by the Commission during its
simultaneous reviews of applications for design approvals and design
certifications.
Although an applicant for design approval does not need to specify
a particular site for the nuclear power plant, which is required in a
combined license application, it does need to identify the site
parameters that the standard nuclear power plant or major portion
thereof is designed to meet, e.g., postulated values for the safe
shutdown earthquake response spectra and maximum tornado wind speed.
These parameters are usually selected to envelop a large portion of
nuclear plant sites in the United States. Once the design is approved
by the NRC, conformance of the actual site characteristics with the
established site parameters must be demonstrated by an applicant
referencing the design approval and verified by the NRC staff at the
time that the referencing application is submitted, i.e., combined
license application.
Paragraph (a)(7) requires the applicant for design approval to
describe its qualifications to design and analyze a standard nuclear
power plant.
In its staff requirements memorandum (SRM) on SECY-90-377,
``Requirements for Design Certification under 10 CFR part 52,'' dated
February 15, 1991, the Commission stated that information submitted in
an application should incorporate the experience from operating events
in current designs which we want to prevent in the future. Therefore,
for plant designs that are based on or are evolutions of nuclear plants
that have operated in the United States, paragraph (a)(22) requires the
applicant to demonstrate how relevant operating experience insights,
from NRC's generic letters and bulletins issued after the most recent
revision of the applicable SRP and 6 months before the docket date of
the application, have been incorporated into the plant design.
Operating experience includes consideration of operating events and the
reliability and performance of structures, systems, and components. If
the application is for a design that is not based on or is not an
evolution of a nuclear plant that operated in the United States, the
applicant must demonstrate how insights from any relevant international
operating experience have been incorporated into that plant design.
In its SRMs, dated June 26, 1990, and July 21, 1993, on SECY-90-16,
``Evolutionary Light-Water Reactor Certification Issues and their
Relationship to Current Regulatory Requirements,'' and SECY-93-087,
``Policy, Technical, and Licensing Issues Pertaining to Evolutionary
and Advanced Light-Water Reactor Designs,'' respectively, the
Commission approved NRC staff recommendations for selected preventative
and mitigative design features for future light-water reactor designs.
Paragraph (a)(23) requires the applicant to provide a description and
analysis of those design features discussed in SECY-90-16 and SECY-93-
87.
Paragraph (a)(U0 ) requires the application to describe the
interfaces for those design features that are outside the scope of the
approved design, e.g., service water intake structure or ultimate heat
sink or, if the application is for approval of a major portion of the
plant design, the interfaces between the nuclear steam supply system
and the balance of plant.
Paragraph (a)(25) requires the applicant to provide a description
of the design-specific PRA and its results. Guidance on meeting the PRA
information requirements will be provided in separate regulatory
guidance documents.
Paragraph (b) requires applications for ``advanced'' nuclear power
plants to meet the design qualification testing requirements in 10 CFR
50.43(e). Advanced designs differ significantly from evolutionary
light-water reactor designs or incorporate, to a greater extent than
evolutionary designs do, simplified, inherent, passive, or other
innovative means to accomplish their safety functions. Examples of
advanced nuclear power plant designs include General Atomic's Modular
High Temperature Gas-Cooled Reactor, General Electric's Simplified
Boiling Water Reactor, and Westinghouse's AP600.
Section 52.139 Standards for Review of Applications
This section sets forth the parts of 10 CFR that contain applicable
requirements for the technical review of applications for a design
approval. The applicability of these requirements is specified in the
identified parts. The Commission recognizes that new designs may
incorporate design features that are not addressed by the current
standards in 10 CFR parts 20, 50 and its appendices, 73, or 100 and
that new standards may be required to address these new design
features. The Commission will determine whether rulemakings are needed
or appropriate to resolve generic safety issues that are applicable to
multiple designs.
Section 52.141 Referral to the Advisory Committee on Reactor Safeguards
(ACRS)
This section states that the application for design approval shall
be sent to the ACRS for its review of safety issues.
Section 52.143 Staff Approval of Design
This section states that upon completion of the NRC staff's review
of the standard design and receipt of a letter report from the ACRS,
the staff shall issue a final safety evaluation report (FSER) and make
that report available on the NRC's Web site. Also, if the FSER
demonstrates that the standard design is acceptable, the Director of
the Office of New Reactors or the Office of Nuclear Reactor Regulation
may issue a final design approval with appropriate terms and
conditions. The NRC's approval of a standard design is commonly
referred to as an FDA because it is an approval of final design
information.
Section 52.145 Finality of Standard Design Approvals; Information
Requests
This section states that a valid FDA must be relied upon by the
ACRS and NRR in any review of a license application that references the
FDA unless significant new information substantially affects the
staff's FSER. The Commission, Atomic Safety Licensing Board Panel, or
presiding officers are not bound by NRC staff determinations in the FDA
or FSER for the standard plant design. Therefore, there is no issue
preclusion in the mandatory hearing for a combined license that
references an FDA. Generic changes to the standard design can be made
as a compliance backfit or under the backfit process in 10 CFR 50.109.
Under paragraph (c), the justification for requests for information to
FDA holders must be approved by the EDO or his or her designee, in
accordance with the process set forth in 10 CFR 50.54(f).
Section 52.147 Section Duration of Design Approval
The purpose of this section is to specify the time period that an
FDA can be referenced in a construction permit, operating license,
combined license, or manufacturing license application.
[[Page 49454]]
Subpart F--Manufacturing Licenses
Section 52.151 Scope of Subpart
This new section is analogous to the ``scope of subpart'' sections
in subparts A through C of part 52 (e.g., Sec. Sec. 52.13, 52.41,
52.71). Section 52.151 describes the general subject matter of subpart
F as the requirements and procedures applicable to NRC issuance of
licenses authorizing the manufacture of nuclear power reactors to be
installed at sites not identified in the manufacturing license
application. This subpart does not cover the manufacture of
subcomponents (e.g., a pump or a reactor pressure vessel) or major
subassemblies (e.g., an integrated module consisting of a pump, piping
and instrumentation and control) for installation in a nuclear power
plant, either on a specific site, or being delivered for integration
into a nuclear power plant under a manufacturing license issued under
this subpart. For purposes of this subpart, a manufactured ``nuclear
power reactor'' would not include site-specific SSCs such as the site
foundation or SSCs related to the ultimate heat sink.
Section 52.153 Relationship to Other Subparts
This new section is analogous to the ``relationship to other
subpart'' sections in subparts A through C of part 52 (e.g., Sec. Sec.
52.13, 52.43, 52.73). Section 52.153 explains how this subpart relates
to other licensing processes in parts 50 and 52, as well as to the
regulatory approvals in part 52.
A manufactured reactor may only be transported to and installed at
a site for which either a construction permit under part 50 or a
combined license under part 52 has been issued to a licensee, as stated
in paragraph (a). However, the licensing requirements associated with
transport of a manufactured reactor from its place of manufacture to
the site where it is to be installed and operated are not addressed in
this rulemaking.
The NRC will issue a manufacturing license only if it approves the
final design of the reactor to be manufactured. Paragraph (b) provides
that the manufacturing license applicant may reference either a
standard design certification rule or a standard design approval, in
order to speed the NRC's review of the manufacturing license
application. The language of paragraph (b) has been corrected in the
final rule by deleting the reference to ``preliminary or final'' design
approvals, inasmuch as the final part 52 rule does not provide for
preliminary design approvals.
Section 52.155 Filing of Applications
This new section is analogous to the ``filing of applications''
sections in subparts A through C of part 52 (e.g., Sec. Sec. 52.15,
52.45, 52.75). Section 52.155 addresses who may file an application for
a manufacturing license, the administrative requirements with respect
to filing (referring to Sec. Sec. 52.3 and 50.30), and the fees for
filing and review of the application (referring to 10 CFR part 170).
With respect to these matters, a manufacturing license application is
no different than any other license application under parts 50 or 52,
and the applicant shall comply with all of these administrative
requirements (which have been revised as part of the final rule to
refer, as necessary, to manufacturing licenses).
Section 52.156 Contents of Applications; General Information
This new section is analogous to the ``contents of application;
general information'' sections in subparts A through C of part 52
(e.g., Sec. Sec. 52.16, 52.46, 52.77). Section 52.156 requires that
the applicant include the information set forth in Sec. 50.33(a)
through (d) and (j), which are the same information required to be
supplied by applicants of construction permits, early site permits,
operating licenses, and combined licenses. Paragraphs (a) through (d)
of Sec. 50.33 require an application to include information
identifying the applicant, including its name, address, business or
occupation, and certain corporate information, including whether it is
owned, controlled, or dominated by an alien, foreign corporation, or
foreign government. Paragraph (j) of Sec. 50.33 requires the applicant
to segregate and protect any Restricted Data or other defense
information from unclassified information. Manufacturing license
applicants should note that there are other NRC requirements governing
Restricted Data or National Security Information in other parts of 10
CFR Chapter I, including 10 CFR parts 10, 50, and 95.
Section 52.157 Contents of Applications; Technical Information in Final
Safety Analysis Report
This new section is analogous to the ``contents of application;
technical information'' sections in subparts A through C of part 52
(e.g., Sec. Sec. 52.17, 52.47, 52.79). Section 52.157 identifies the
technical information that must be included in an application for a
manufacturing license. These requirements were modeled on those
subparts, in particular subpart B's provisions dealing with standard
design certifications, because of the commonality with respect to the
nature and scope of NRC approval of the design in both regulatory
processes. As with the existing part 50 licensing process, and part
52's combined license and standard design certification processes, the
manufacturing license application must include an FSAR. The FSAR
contains the information necessary for the NRC to determine the safety
of the reactor design to be manufactured and the adequacy of the
applicant's proposed means of assuring that the manufacturing conforms
to the design. The FSAR must contain a level of detail sufficient to
permit preparation of construction and installation specifications by
an applicant who seeks to use the manufactured reactor, and for the NRC
to prepare acceptance and inspection requirements.
The information required to be included in the manufacturing
license FSAR is largely the same as what is required for a design
certification or combined license, but the requirements have been
modified as necessary to reflect the fact that the design and
manufacture of a reactor is being approved by license, but that the
reactor must be transported to a site and integrated into site specific
plant elements in order to operate. In addition, unlike the case with a
design certification, the NRC is not distinguishing between
evolutionary plants versus more advanced plants with respect to the
level of detail required to be developed to support the license
application. The NRC expects that the designs of all manufactured
plants will be completed at a level of detail sufficient for: (1) The
holder of the manufacturing license to develop procurement,
construction and installation specifications; and (2) the NRC to
develop acceptance and inspection requirements.
Paragraph (a) requires that the FSAR contain the principal design
criteria for the reactor to be manufactured, and references appendix A
to 10 CFR part 50 as establishing minimum requirements for the
principal design criteria for water-cooled nuclear power plants. The
NRC expects to develop technology-neutral design criteria for non-light
water cooled reactor designs in the future. This requirement was drawn
from Sec. 50.34(a)(3)(i).
Paragraph (b) requires that the FSAR describe the design bases and
the relation of the design bases to the principal design criteria that
are identified in accordance with paragraph (a). This requirement was
drawn from Sec. 50.34(a)(3)(ii).
[[Page 49455]]
Paragraph (c) requires that the FSAR describe and analyze the
structures, systems, and components of the reactor to be manufactured,
with the objective of demonstrating that the necessary safety functions
will be accomplished. This requirement was drawn from Sec. 50.34(a)(1)
and (b)(2), but modified to reflect the fact that a manufacturing
license represents approval of a final reactor design.
Paragraph (d) requires that the FSAR describe the safety features
that are engineered into the reactor. This requirement was drawn from
Sec. 50.34(a)(1)(ii)(D), but modified to reflect the fact that a
manufacturing license represents approval of a final reactor design.
Paragraph (e) requires the FSAR to describe the kinds and
quantities of radioactive materials expected to be produced in the
operation and the means for controlling and limiting radioactive
effluents and radiation exposures within the limits set forth in part
20.
Paragraph (f) requires that the FSAR include that information
necessary to establish that the design of the reactor to be
manufactured complies with 18 delineated technical requirements in 10
CFR part 50. Applicants and licensees should note that the part 50
requirements listed in paragraph (f) do not constitute the sum total of
requirements in part 50 for which either an applicant for or holder of
a manufacturing license must comply with in its application and
throughout the life of its license. Rather, the listed requirements in
paragraph (f) simply represents the minimum necessary content of the
FSAR for a manufacturing license. The part 50 requirements listed in
paragraph (e) are mainly applicable to LWRs. Potential applicants and
licensees should also note that the NRC may, in the future, adopt
additional technical requirements in part 50 applicable to LWRs. If the
NRC believes that future manufacturing license holder's compliance with
that new requirement must be documented and controlled through the
FSAR, the NRC will make a conforming change in Sec. 52.157 to refer to
the new part 50 requirement. A similar course would also be followed if
the NRC backfits, in accordance with the finality provisions in Sec.
52.171, the new requirement on existing manufacturing licenses.
Paragraph (f)(19) requires that the FSAR include the site
parameters postulated for the design of the manufactured reactor.
Although an applicant for a manufacturing license does not need to
specify a particular site where the manufactured reactor will be
integrated into a nuclear power plant, as in a combined license
application, it does need to identify the site parameters, under
paragraph (f)(20), that the manufactured reactor is designed to meet,
e.g., postulated values for the safe-shutdown earthquake response
spectra and maximum tornado wind speed. These parameters are usually
selected to envelop a large portion of nuclear plant sites in the
United States. Once the manufacturing license is issued by the NRC,
conformance of the actual site with the established site parameters
must be demonstrated by the applicant referencing the use of the
manufactured reactor.
Paragraph (f)(20) requires the FSAR to describe the interface
requirements for those design features that are outside the scope of
the design of the manufactured reactor, e.g., service water intake
structure or ultimate heat sink, and paragraph (f)(21) requires
justification that compliance with the interface requirements in
paragraph (g) can be verified through inspections or tests (which may
be conducted at the plant where the manufactured reactor is utilized,
or elsewhere, e.g., the place of manufacture) or analysis. This
paragraph does not require, however, that the FSAR contain ``acceptance
criteria'' for determining whether the interface requirements have been
met.
Paragraph (f)(22) requires the FSAR to include a representative
conceptual design for the nuclear power facility using the manufactured
reactor. This will be used by the NRC in its review of the FSAR, to
assess the adequacy of the interface requirements in paragraph (g) of
this section, and to help the Commission in determining the adequacy of
the site parameters and design characteristics to be included in the
manufacturing license. The conceptual design will not, however, be
approved as part of the manufacturing license and the Commission does
not anticipate directly requiring a nuclear power plant utilizing the
manufactured reactor to use the conceptual design. Instead, the
Commission intends to use site parameters, design characteristics,
ITAAC, and interface requirements to ensure that the manufactured
reactor will be utilized safely at a specific nuclear power plant.
Paragraph (f)(23) requires the applicant to provide a description
and analysis of design features to address prevention and mitigation of
severe accidents, consistent with the Commission's SRM on SECY-91-229,
``Severe Accident Mitigation Design Alternatives for Certified Standard
Designs,'' dated October 25, 1991.
Paragraph (f)(U0 ) is reserved to accommodate any new requirement
for the contents of an FSAR submitted as part of an application for a
manufacturing license which the Commission may adopt in the future.
Paragraph (f)(25) requires FSARs for modular nuclear power plant
designs to describe and analyze the various options for the
configuration of the multi-reactor nuclear power plant. Modular nuclear
power plant designs are defined in Sec. 52.1. Modular designs are not
portions of a single nuclear plant, rather they are separate nuclear
reactors with some shared or common systems.
Paragraphs (f)(26)(i), (ii), (iii), and (v) focus on FSAR
information necessary to demonstrate applicants technical, managerial,
and organizational capability and resources to design and manufacture a
nuclear power reactor consistent with the approved design, and in
accordance with all applicable requirements.
Paragraph (f)(26)(iv) requires the FSAR to include proposed
procedures for the preparation of the manufactured reactor for
shipping, the conduct of shipping, and for verifying the condition of
the manufactured reactor upon receipt at the site. However, the holder
of the manufacturing license need not be responsible for implementing
the procedures for verifying the condition of the reactor upon receipt
at the site. The NRC will require the licensee whose application
referenced the use of the manufactured reactor to implement the
approved verification procedures (this could be done as a license
condition). With respect to shipping, the holder of the manufacturing
license may use an agent (e.g., a shipping company) to transport the
reactor. To ensure that the shipping requirements in the manufacturing
license are complied with by the third party transporter, the NRC has
included a provision in Sec. 52.167(c)(2) requiring the manufacturing
license holder to include, in any contract governing the transport of a
manufactured reactor from the place of manufacture to any other
location, a provision requiring that the person or entity transporting
the manufactured reactor to comply with all NRC-approved shipping
requirements in the manufacturing license.
For plant designs that are based on or are evolutions of nuclear
plants that have operated in the United States, paragraph (f)(29)
requires the applicant to demonstrate how relevant operating experience
insights, from NRC's generic letters and bulletins issued after the
most recent revision of the applicable SRP and 6 months before the
docket
[[Page 49456]]
date of the application, have been incorporated into the design of the
reactor to be manufactured. Operating experience includes consideration
of operating events and the reliability and performance of structures,
systems, and components. If the application is for a design that is not
based on or is not an evolution of a nuclear plant that operated in the
United States, the applicant must demonstrate how insights from any
relevant international operating experience have been incorporated into
that manufactured reactor design.
Paragraph (f)(31) requires that the FSAR include a description of
the design--specific probabilistic risk assessment and its results.
Section 52.158 Contents of Application; Additional Technical
Information
This new section is analogous, in organizational structure, to
Sec. 52.80, ``Contents of application; additional technical
information'' in subpart C of part 52.
Paragraph (a) requires that the application include inspections,
tests, and analyses that the licensee who will be placing the
manufactured reactor on a site and operating the reactor shall perform
and their associated acceptance criteria. The purpose of these ITAAC
are to ensure that: (1) The reactor has been manufactured in
conformance with applicable requirements; and (2) the manufactured
reactor, as emplaced at the site and integrated into any site-specific
portions of the nuclear power plant, will operate in conformance with
the design characteristics in the manufacturing license, the license
authorizing operation of the manufactured reactor, and applicable
requirements. Paragraph (a)(3), which is analogous to Sec.
52.80(a)(3), provides that if the manufacturing license references a
standard design certification, the manufacturing license application
may include a notification that one or more ITAAC in the referenced
design certification rule has been met. In such a situation, the
Federal Register notice of docketing a hearing required by Sec. 52.163
must specifically indicate that the application includes such a
notification.
Paragraph (b)(1) requires that the application include an
environmental report meeting the requirements in 10 CFR 51.54, which
specifies the environmental information that must be submitted by a
manufacturing license applicant to support the NRC's NEPA review. The
Commission notes that environmental report need not include a
discussion of assessment of the benefits and impacts of constructing
and operating the manufactured reactor or an evaluation of alternative
energy sources, under Sec. 52.163 and Sec. 51.54.
Under Sec. 51.54, the environmental report for a manufacturing
license must address the costs and benefits of SAMDAs that could be
incorporated into the design, and the bases for not including SAMDAs
into the design. The SAMDA information that must be included is
essentially the same information that must be provided to support an
application for a standard design certification. However, if the
application references a standard design certification, Sec. 51.54
provides that the manufacturing license's environmental report need not
include the SAMDA evaluation. In such a case, the SAMDA determination
in the EA for the referenced design certification would have finality
in the manufacturing license proceeding, in accordance with Sec.
52.63.
Section 52.159 Standards for Review of Applications
This new section is analogous to the ``standards for review of
applications'' sections in subparts A through C of part 52 (e.g.,
Sec. Sec. 52.18, 52.48, 52.81). Section 52.159 identifies the
regulations that the NRC will use in reviewing an application for a
manufacturing license. The NRC recognizes that reactors to be
manufactured under a manufacturing license may incorporate design
features which are inconsistent with current requirements in 10 CFR
Chapter I, and may require exemptions from current requirements. Such
exemptions would be granted as part of the NRC's issuance of the
manufacturing license, together with alternative requirements
(analogous to the ``applicable regulations'' provisions in the current
design certifications rules, 10 CFR part 52, appendices A-D, Section
V).
Section 52.161 Reserved
This section is reserved to accommodate any new requirements on the
application process for manufacturing license which the NRC may adopt
in the future.
Section 52.163 Administrative Review of Applications; Hearings
This new section is analogous to the ``administrative review of
applications'' sections in subparts A through C of part 52 (e.g.,
Sec. Sec. 52.21, 52.51, 52.85). Section 52.163 specifies that the
procedural requirements in 10 CFR part 2 apply to the NRC's processing
of an application for a manufacturing license, including docketing of
the initial application.
Section 52.163 reiterates the Sec. 2.105 requirement that the NRC
publish in the Federal Register a notice of proposed action on the
application. Apart from the required Federal Register notice, the
Commission also expects to publish on the NRC's Web site notice of
docketing of the application and the opportunity to intervene in the
proceeding, consistent with the Commission's discussion in the 2004
final part 2 rulemaking (January 14, 2004; 69 FR 2182, 2198-99). The
section makes clear, consistent with Sec. 51.54, that the
environmental report submitted by the manufacturing license applicant
need not contain an assessment of the benefits of constructing and/or
operating the manufactured reactor or an evaluation of alternative
energy sources.
Finally, this section indicates that the hearing on the
manufacturing license application will be governed by the procedures in
part 2, subparts C, G, L, and N. The Commission notes that although
subpart G is listed in this paragraph, it is unlikely that there would
be contentions meeting the criteria in Sec. 2.310 (and reiterated in
Sec. 2.700) for conduct of the hearing under subpart G. This is
because the primary focus of the manufacturing license proceeding is on
the adequacy of the design to be manufactured, and the nature of issues
which are most likely to be raised on the design would not ordinarily
involve issues of material fact relating to either: (1) The occurrence
of a past activity, where the credibility of an eyewitness may
reasonably be expected to be at issue; or (2) issues of motive or
intent of the party or eyewitness which are material to the resolution
of the contested matter.
Section 52.165 Referral to the Advisory Committee on Reactor Safeguards
(ACRS)
This new section is analogous to the ``Referral to the Advisory
Committee on Reactor Safeguards'' sections in subparts A through C of
part 52 (e.g., Sec. Sec. 52.21, 52.53, 52.87). It provides that the
ACRS will have the same role with respect to manufacturing licenses
that it has for other nuclear power plant licenses, in that it will
report on those portions of the application which concern safety.
Section 52.167 Issuance of Manufacturing License
This new section is analogous to the ``issuance'' sections in
subparts A through C of part 52 (e.g., Sec. Sec. 52.24, 52.54, 52.97).
Paragraph (a) sets forth the timing of issuance of a manufacturing
license and the findings that the Commission must make in
[[Page 49457]]
order to issue the manufacturing license. The findings that must be
made are similar to those necessary to issue a construction permit,
inasmuch as construction is analogous to manufacturing. The Commission
notes that it reserves the right to withhold issuance of the
manufacturing license, even if all the rules and regulations of the
Commission have been satisfied, based on public health and safety or
common defense and security information or considerations not
adequately addressed in the Commission's rules and regulations.
Paragraph (b) identifies the specific limitations that the
Commission will include in each manufacturing license. They include
technical specifications for the operation of each manufactured
reactor, site parameters, design characteristics, and interface
requirements, which are to be used by the applicant for and holder of
the license referencing the use of the manufactured reactor(s).
Ordinarily, the limitations to be included in the manufacturing license
would be derived from the manufacturing license application, but the
NRC may modify the proposed limitations based upon the NRC's review.
Paragraph (c) restricts the holder of the manufacturing license
from transporting or allowing to be removed from the place of
manufacture the manufactured reactor except to the site of a licensee
who holds either a construction permit or combined license referencing
the use of that manufactured reactor.
Section 52.169 Reserved
This section is reserved to accommodate any new requirements on
either the issuance of, or activities authorized under a manufacturing
license which the Commission may adopt in the future. Any new
requirements adopted after issuance of a manufacturing license, which
are made applicable to that manufacturing license, would have to
satisfy the finality restrictions in Sec. 52.171.
Section 52.171 Finality of Manufacturing Licenses; Information Requests
This new section is analogous to the variously entitled sections
addressing finality and special backfitting protections which are in
subparts A through C of part 52 (e.g., Sec. Sec. 52.39, 52.63,
52.98),\15\ but is more generally modeled on the finality provision for
standard design certifications. In general, paragraph (a) addresses
backfitting and finality restrictions on the NRC, paragraph (b)
addresses finality and standardization restrictions applicable to the
licensee (i.e., the holder the manufacturing license), and paragraph
(c) establishes restrictions on certain NRC information collections
with respect to the manufacturing license.
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\15\ The finality provision in Sec. 52.83 performs a different
function than the finality sections cited above, in that it points
back to, and thereby re-emphasizes, the primary finality provisions
for each license or regulatory approval mechanism in part 52, e.g.,
the finality provision in Sec. 52.39 for early site permits.
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Paragraph (a)(1) states that the Commission may not modify,
rescind, or impose new requirements on the design of a nuclear power
reactor being manufactured, or new requirements for the manufacture of
the nuclear power reactor, unless the Commission determines that a
modification is necessary to either bring the design or the manufacture
of the reactor into compliance with the Commission's requirements
applicable and in effect at the time the manufacturing license was
issued, or to provide reasonable assurance of adequate protection to
public health and safety or common defense and security. This
restriction on the Commission applies, inter alia, in construction
permit, operating license, and combined license proceedings which
reference the use of the manufactured reactor. It also applies in any
enforcement proceeding initiated by the NRC, or in a rulemaking which
proposes to apply new or changed requirements to reactors which have
already been manufactured, as well as any reactors yet to be
manufactured under the manufacturing license. However, the restrictions
in paragraph (a)(1) do not apply to NRC information requests directed
at either the manufacturing license holder, or to any holder of a
license referencing the use of a manufactured reactor; such information
requests are governed by paragraph (c) of this section.
Paragraph (a)(2) provides that any modification to the design of a
manufactured nuclear power reactor which is imposed by the Commission
under paragraph (a)(1) of this section will be applied to all reactors
manufactured under the license, including those that have already been
manufactured, transported, sited, and are in operation. The only
exception would be for those reactors to which the Commission-ordered
modification had been rendered technically irrelevant by action taken
under paragraph (b) of this section, i.e., either the holder of the
manufacturing license has requested a change to the design approved in
the manufacturing license (which ordinarily would apply only to
reactors manufactured after Commission approval of the change), or the
holder of a license referencing the use of the manufactured reactor has
obtained Commission approval for a change to the design of the specific
manufactured reactor(s) utilized by that licensee.
Paragraph (a)(3) delineates the nature of finality associated with
the referencing of a manufactured reactor in subsequent NRC licensing
proceedings. This paragraph provides that finality is accorded to those
matters resolved in the proceeding on the issuance or renewal of the
manufactured reactor. These matters resolved include the adequacy of
the design of the manufactured reactor and the acceptability and
completeness of the ITAAC required by Sec. 52.158(a)(1) to be
performed by the licensee operating the reactor. The matters resolved
also include the SAMDA evaluation prepared by the Commission in
compliance with its obligations under NEPA. This finality extends to
both the Commission's determinations with respect to specific SAMDA
features included in the design of the manufactured reactor, as well as
the Commission's determinations regarding the lack of need for any
other SAMDA features. Finality is accorded in the following situations:
(1) Issuance of a construction permit, operating license, combined
license; (2) any hearing under Sec. 52.103; and (3) enforcement
hearings other than those proceedings initiated by the Commission under
paragraph (a)(1).
Paragraph (b)(1) requires the holder of a manufacturing license to
seek a prior NRC review and approval for any change to the design of
the nuclear power plant authorized to be manufactured. The holder of
the manufacturing license may not make a change to the approved design
for manufacture through the provisions of Sec. 50.59. A request for a
change to the approved design must be in the form of a license
amendment application, and the application will be processed in
accordance with Sec. Sec. 50.90 through 50.92. The Commission notes,
however, that the procedures for no significant hazards consideration
(NSHC) are not applicable to manufacturing licenses, inasmuch as
Section 189.a.(2) of the AEA, which is the statutory authority for
these procedures, does not apply to manufacturing licenses.
Paragraph (b)(2) requires a holder of a license referencing the use
of a manufactured reactor, who wishes to depart from the design
characteristics, site parameters, terms and conditions,
[[Page 49458]]
or approved design of the manufactured reactor, to seek a departure
from the NRC. The manner in which a departure is granted depends upon
the timing of the request. If a departure is requested as part of the
initial combined license application, the departure would be treated as
part of the application and issued as part of the combined license. By
contrast, if the same departure were sought after the combined license
had been issued, then the licensee must apply for the departure in the
form of a license amendment. The criteria for granting the departure is
the exemption criterion in Sec. 52.7; however, the departure itself is
not considered an exemption (unless, of course, the departure also
involves a non-compliance with an underlying Commission regulatory
requirement in 10 CFR Chapter I). Thus, the Commission will not approve
a departure unless the Commission finds, in addition to the routine
exemption criteria in Sec. 52.7, that special circumstances outweigh
any decrease in safety that may result from the reduction in
standardization caused by the departure. As explained earlier, these
limitations are intended to maintain the standardization of
manufactured reactors in operation to the extent practicable. The
licensee may not depart from the design characteristics, site
parameters, terms and conditions, or approved design of the
manufactured reactor through the provisions of Sec. 50.59.
Paragraph (c), which is analogous to Sec. Sec. 52.39(d), 52.98(g),
and 52.145(c), provides that NRC information requests must be evaluated
before issuance to ensure that the burden to be imposed by the
information request is justified in view of the potential safety
significance of the issue to be addressed, except when the information
requests seeks to verify compliance with the current licensing basis of
either the manufacturing license or the manufactured reactor. This
paragraph applies to information requests directed at either the holder
of the manufacturing license or the holder of a license referencing the
use of a manufactured reactor. Information requests may be in the form
of a new rule requiring submission of information (i.e., a new
information collection and reporting requirement), or in the form of a
NRC staff request for information. Information requests by the staff
must be in accordance with 10 CFR 50.54(f) and must be approved by the
EDO or his or her designee before the request may be issued.
Section 52.173 Duration of Manufacturing License
This new section is analogous to the variously-entitled sections
addressing duration (term) of each regulatory process in subparts A
through C of part 52 (e.g., Sec. Sec. 52.33, 52.61, 52.104). Under
Sec. 52.173, a manufacturing license may be issued for not less than 5
nor more than 15 years. Manufacturing of a new reactor may not commence
less than 3 years before the expiration of the manufacturing license,
even though a timely application for renewal has been filed in
accordance with Sec. 52.177. However, if a timely application for
renewal of the manufacturing license has been docketed, manufacturing
of uncompleted reactors whose manufacture commenced 3 years or more
before the expiration date, may continue past the date of expiration of
the license until the NRC acts upon the renewal application, consistent
with the ``Timely Renewal'' doctrine of the Administrative Procedures
Act. The NRC believes that timely renewal protection should only be
provided to those applications which are of sufficient quality to be
docketed. This is consistent with the requirement in Sec. 2.109(b)
requiring filing of a ``sufficient'' application for renewal of
operating licenses as a prerequisite for the applicability of the
timely renewal protection.
Section 52.175 Transfer of Manufacturing License
This new section is analogous to the variously entitled transfer
sections in subparts A and C of part 52 (e.g., Sec. Sec. 52.28,
52.105).\16\ Section 52.175 provides that a manufacturing license may
be transferred in accordance with Sec. 50.80, which constitutes the
Commission's common procedures and criteria governing transfers of
nuclear power plant licenses. The matters to be addressed in a transfer
are limited to the matters identified in Sec. 50.80(b), and the
transfer would not be an opportunity for the Commission to reconsider
safety and environmental matters previously resolved, or to address new
safety matters other than the narrow scope of matters identified in
Sec. 50.80(b).
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\16\ A standard design certification is a rule, rather than a
license. Accordingly, there is no ``holder'' of a standard design
certification rule and no need for a provision addressing
``transfer'' of a standard design certification rule.
---------------------------------------------------------------------------
Section 52.177 Application for Renewal
This new section is analogous to the ``application for renewal''
sections in subparts A through C of part 52 (e.g., Sec. Sec. 52.29,
52.57, 52.107). Section 52.177 sets forth the content of an application
for renewal, specifies the administrative requirements governing the
application, addresses the effectiveness of a manufacturing license
during the period of NRC's consideration of the renewal application,
summarizes how an interested person may request a hearing on the
renewal, and addresses the referral of the renewal application to the
ACRS and the Commission's expectations with respect to the ACRS report
on the application.
Section 52.179 Criteria for Renewal
This new section is analogous to the ``criteria for renewal''
sections in subparts A and B of part 52 (e.g., Sec. Sec. 52.31,
52.59).\17\ Section 52.179 provides that the Commission may grant
renewal of a manufacturing license if the Commission determines that
the license complies with the relevant provisions of the AEA, the
Commission's regulations applicable and in effect at the time the
manufacturing license was originally issued, and any new requirements
which the Commission imposes which: (1) Are necessary for reasonable
assurance of adequate protection to public health and safety or common
defense and security; (2) are necessary for compliance with
Commission's regulations and orders applicable and in effect at the
time the manufacturing license was originally issued; or (3) represent
a substantial increase in overall protection of the public health and
safety or common defense and security and the direct and indirect costs
of implementation are justified in light of the increased protection.
These ``backfitting'' restrictions are similar to--if somewhat narrower
than--the backfitting restrictions applicable to renewal of standard
design certification rules under subpart B of this part.
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\17\ Subpart C does not contain a ``criteria for renewal''
provision, inasmuch as the renewal would be governed by 10 CFR part
54, see Sec. 52.107. Part 54 contains a provision, Sec. 54.29,
setting forth the standards for issuance of renewed licenses.
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Reasonable assurance of adequate protection to public health and
safety and common defense and security is provided under this
regulatory approach, inasmuch as paragraph (b) allows the Commission to
impose new requirements which are necessary for common defense and
security, or are necessary for compliance with the Commission's
regulations and orders applicable and in effect at the time the
manufacturing license was originally issued.
[[Page 49459]]
Section 52.181 Duration of Renewal
This new section is analogous to the ``duration of renewal ''
sections in subparts A and B of part 52 (e.g., Sec. Sec. 52.33,
52.61).\18\ Section 52.181 specifies the term of a renewed
manufacturing license as not less than 5 nor more than 15 years from
the date of expiration of the prior manufacturing license. Thus, a
holder of a manufacturing license with an original term of 15 years,
who is granted a 15-year renewal of the manufacturing license 4 years
before expiration of the license, will obtain a renewed manufacturing
license of 19 years, representing a 15-year term of the renewed license
plus the 4 years remaining on its original license.
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\18\ Subpart C does not contain a ``duration of renewal''
provision, inasmuch as the renewal would be governed in all respects
by 10 CFR part 54, see Sec. 52.107. Part 54 contains a provision,
Sec. 54.31, governing the duration of renewed licenses.
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Subpart G--Reserved
This subpart is reserved for future use by the Commission.
Subpart H--Enforcement
This subpart contains two provisions, Sec. 52.301 and Sec.
52.303, which are comparable to former Sec. 52.111 and Sec. 52.113,
and are analogous to provisions contained in other parts of 10 CFR
Chapter I imposing requirements on regulated entities.
Section 52.301 reiterates, and provides notice to licensees and
applicants under part 52 of the Commission's authority to obtain
injunctions or other court orders for the violations enumerated in this
paragraph.
Section 52.303 provides notice to all persons and entities subject
to part 52 that they are subject to criminal sanctions for willful
violations, attempted violations, or conspiracy to violate certain
regulations under part 52. The regulations for which criminal penalties
apply are limited to those which establish either a regulatory
obligation or prohibition. Most of the regulations in part 52 are
procedural or administrative in nature, and therefore were listed in
Sec. 52.113 as not being subject to criminal sanctions. The
regulations in part 52 which are subject to criminal sanctions are
Sec. Sec. 52.4 (Deliberate misconduct), 52.5 (Employee protection),
52.6 (Completeness of information), 52.25 (Extent of activities
permitted), 52.35 (Use of site for other purpose), 52.91 (Authorization
to conduct site activities), and 52.110 (Termination of license).
Appendix A--U.S. Advanced Boiling Water Reactor
Refer to the section-by-section discussion in the final rule dated
May 12, 1997 (62 FR 25800).
Appendix B--The System 80+ Design
Refer to the section-by-section discussion in the final rule dated
May 21, 1997 (62 FR 27840).
Appendix C--The AP600 Design
Refer to the section-by-section discussion in the final rule dated
December 23, 1999 (64 FR 72002).
Appendix D--The AP1000 Design
Refer to the section-by-section discussion in the final rule dated
January 27, 2006 (71 FR 4464).
Appendix N--Combined Licenses for Nuclear Power Reactors of Identical
Design
Appendix N of part 52 contains the Commission's procedures which
may be used by one or more applicants for combined licenses under part
52, where the applications seek to construct and operate nuclear power
reactors of identical design to be located at multiple sites. The
comparable procedures governing applications for construction permits
and operating licenses using identical nuclear power reactor designs
remain in appendix N of 10 CFR part 50. Hearings for applications filed
under appendix N in part 52, as well as part 50, are governed by
subpart D of part 2. Thus, appendix N and subpart D of part 2 are
integral to each other.
The regulations in appendix N of part 52 apply in two situations:
(1) Where the same applicant seeks combined licenses at different sites
utilizing the identical reactor design; and (2) where two or more
different applicants each seek combined licenses at different sites
utilizing the identical reactor design. In either situation, there is
an identical reactor design. The Commission has deliberately used the
term, ``nuclear power reactor,'' in appendix N and subpart D of part
2--as distinguished from the term, ``nuclear power plant''--to make
clear that the site-specific elements, such as the service water intake
structure or the ultimate heat sink, need not be identical in order for
appendix N and subpart D to apply.
The Commission has conformed appendix N and subpart D of part 2 to
use the term, ``identical'' nuclear power reactor design, and removed
references to ``duplicate'' and ``essentially identical.'' For purposes
of appendix N and subpart D of part 2, designs for reactors are
``identical,'' even if individual licensees request plant-specific
departures or exemptions from a referenced standard design
certification (or application). However, those plant-specific
departures or exemptions are not part of the ``common design.''
Therefore, the NRC's review of those departures and exemptions, as well
as NRC hearings on those departures and exemptions, would be conducted
separately as part of the safety review of each individual application,
and would not be part of the hearing on the common design which would
be conducted under subpart D of part 2.
Section 1
This is a new section specifying that its provisions apply to
applicants for combined licenses under subpart C of part 52. Appendix N
of part 50 would apply to applicants for construction permits and
operating licenses who use identical reactor designs.
Section 2
This section, which is analogous to and derived from former Sec. 2
of appendix N, specifies that each application submitted under this
appendix must be submitted in accordance with the delineated Commission
filing requirements. In addition, to ensure that the NRC is clearly
informed that the applicants wish to have their application processed
under appendix N and subpart D of part 2, this section requires: (1)
That each application state the applicant's intent that the application
be processed by the NRC under appendix N; and (2) that all of the
applications to be treated together under this appendix be listed in
each application. All of the applications must be filed simultaneously,
which will facilitate NRC's administrative handling and technical
review of the applications, as well as efficient conduct of the hearing
process.
Section 3
This section, which is analogous to and derived from former Sec. 3
of appendix N, specifies that combined license applications submitted
under this appendix must include all of the information required to be
submitted in a combined license application in Sec. Sec. 52.77, 52.79,
and 50.80(a) and (b), but makes clear that each of the applications
must identify the common design. The common design may be (but is not
limited to) a standard design certification under subpart B of part 52,
a standard design approval, a ``common custom design,'' or a
manufactured reactor.
[[Page 49460]]
The FSAR for each application must either incorporate by reference
or include the FSAR for the common design, including, as applicable,
the FSAR for the referenced design certification or manufactured
reactor. ``Include,'' means that the FSAR may not simply reference the
common FSAR; the information from the referenced FSAR must be included
within each application's FSAR.
Section 4
This is a new section specifying that each application must submit
an environmental report which complies with the applicable provisions
of part 51 with respect to the content of environmental reports. As an
alternative, this section provides that one or more of the applicants'
environmental reports may incorporate by reference a single
environmental report describing the environmental impacts of the common
design at each of the sites.
Section 5
This is a new section specifying that, upon a determination that
each application is acceptable for docketing, each application will be
docketed and a notice of docketing will be published in the Federal
Register in accordance with 10 CFR 2.104. The notice of docketing must
state that the application will be processed under the provisions of
appendix N. Separate notices of docketing are contemplated, so that a
problem with acceptance review of one application will not prevent the
docketing and initiation of the NRC's technical review of the other
applications determined to be sufficient and acceptable for docketing.
This could occur, for example, if information, submitted by an
applicant which is unrelated to the common design, is determined by the
NRC to be insufficient. However, if the applications are determined to
be acceptable for docketing, Sec. 5 provides the Commission with the
discretion to publish a single notice of docketing for those
applications.
Section 6
This is a new section which provides that the NRC will prepare a
separate draft and final EIS for each of the applications. Scoping may
be conducted simultaneously but need not be conducted jointly (e.g.,
scoping for an application at site 1 need not be conducted as part of
the same process as the scoping for an application for site 2), at
least with respect to site-specific environmental issues. However, for
environmental issues related to the common design, the NRC has the
discretion to conduct joint scoping. The NRC staff is not, however,
required to prepare a joint environmental impact statement for the
common design.
This section also addresses the content of an EIS when the
applications reference either a standard design certification or the
use of a manufactured reactor of common design. In either case, the NRC
has already prepared and finalized an EA which addresses SAMDAs. This
SAMDA analysis is accorded finality under the provisions of Sec. Sec.
52.63 and 52.171, respectively. Therefore, the EIS for each of the
applications must reference the relevant environmental assessment
containing the SAMDA analysis.
Section 7
This section, which is analogous to and derived from former Sec. 1
of appendix N, provides direction to the ACRS with respect to their
report on each of the combined license applications. The ACRS must
issue a separate report on the safety of the common design, except in
those instances where the applications are referencing either a
standard design certification or manufactured reactor (of common
design). In addition, the ACRS must issue a separate report for each
application. This report must be limited to those matters which are not
relevant to the common design. This will facilitate the NRC's licensing
process by eliminating overlap and ensuring that the ACRS reports are
carefully focused on the relevant safety issues.
Section 8
This is a new section, which provides that the Commission shall
designate a presiding officer to conduct the proceeding with respect to
the health and safety, common defense and security, and environmental
matters (i.e., SAMDAs) relating to the common design. The presiding
officer will conduct the hearing in accordance with subpart D of part
2. The presiding officer is required to issue a separate partial
initial decision on matters relevant to the common design, consistent
with 10 CFR 2.405 in subpart D of part 2. Appeals of the partial
initial decision are governed by 10 CFR 2.341, as provided by 10 CFR
2.405. The NRC also notes that issues on the contested design may not
be relitigated in a different phase of the hearing except on the basis
of significant new information that substantially affects the
conclusion(s) reached at the other phase or other good cause. See 10
CFR 2.406.
VII. Availability of Documents
The NRC is making the documents identified below available to
interested persons through one or more of the following methods as
indicated.
Public Document Room (PDR). The NRC Public Document Room is located
at 11555 Rockville Pike, Rockville, Maryland.
Rulemaking Web site (Web). The NRC's interactive rulemaking Web
site is located at http://ruleforum.llnl.gov. These documents may be
viewed and downloaded electronically via this Web site.
NRC's Public Electronic Reading Room (EPDR). The NRC's electronic
public reading room is located at http://www.nrc.gov/reading-rm.html.
The NRC staff contact. Nanette V. Gilles, Mail Stop O-4D9A,
Washington, DC 20555-0001, 301-415-1180.
----------------------------------------------------------------------------------------------------------------
Document PDR Web EPDR NRC staff
----------------------------------------------------------------------------------------------------------------
Part 52 Rule, Cross-Reference Tables............ .............. X ML062550246 X
Comments received............................... X X X
Comment Summary Report.......................... .............. .............. ML063450216
Regulatory Analysis............................. X X ML071490350 X
Regulatory History Index for the proposed July .............. .............. ML032810026
2003 rule......................................
Regulatory History Index for the March 13, 2006, .............. .............. ML062080575
proposed rule..................................
----------------------------------------------------------------------------------------------------------------
VIII. Agreement State Compatibility
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement State Programs'' which became effective on September 3, 1997
(62 FR 46517), NRC program elements (including regulations) are placed
into compatibility categories A, B, C, D, NRC, or adequacy category,
Health and Safety (H&S). Category A includes program elements that are
basic radiation protection standards or related
[[Page 49461]]
definitions, signs, labels, or terms necessary for a common
understanding of radiation protection principles and should be
essentially identical to those of NRC. Category B includes program
elements that have significant direct transboundary implications and
should be essentially identical to those of the NRC. Compatibility
Category C includes program elements that do not meet the criteria of
Category A or B, but the essential objectives of which an Agreement
State should adopt to avoid conflict, duplication, gaps, or other
conditions that would jeopardize an orderly pattern in the regulation
of agreement material on a nationwide basis. Compatibility Category D
includes those program elements that do not meet any of the criteria of
Category A, B, or C, and do not need to be adopted by Agreement States.
Compatibility Category NRC includes program elements that address areas
reserved to the Commission and cannot be relinquished to Agreement
States pursuant to the Atomic Energy Act or provisions of Title 10 of
the Code of Federal Regulations. An Agreement State may inform its
licensees of certain of these NRC provisions through a mechanism that
is appropriate under the State's administrative procedure laws as long
as the State adopts these provisions solely for the purposes of
notification, and does not exercise any regulatory authority pursuant
to them. Category H&S include program elements that are not required
for compatibility, but have a particular health and safety role in the
regulation of agreement material and the State should adopt the
essential objectives of the NRC program elements. In addition, a State
should not adopt provisions that would preclude, or effectively
preclude, a practice authorized by the Atomic Energy Act, and in the
national interest. The proposed revisions are categorized as follows:
List of Changes 10 CFR Part 52 Final Rulemaking
----------------------------------------------------------------------------------------------------------------
Description new, Compatibility Comments regarding
Sections changes designation compatibility designation
----------------------------------------------------------------------------------------------------------------
10 CFR Part 1..................... Statement of D.................... This provision is designated
Organization and Category D because it does
General Information. not meet any of the criteria
of Category A, B, or C. A
State may adopt similar
provisions to reflect their
organizational structure and
may wish to inform its
licensees of the provisions
of this part through a
mechanism that is appropriate
under the State's
administrative procedure
laws.
10 CFR Part 2--Rules of Practice
for Domestic Licensing
Proceedings and Issuance of
Orders
2.1........................... Scope................ D, except portions of These provisions are
these provisions are designated Compatibility
NRC. Category D because they do
not meet any of the criteria
of Category A, B, or C. A
State may adopt similar
provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. Those
portions of the provision
that address areas reserved
to the NRC, e.g., 10 CFR Part
52 standard design approvals,
are designated as a
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Act, 10 CFR 8.4, 10 CFR
Part 150, and other Federal
laws, regulations, or
provisions.
2.4--Definitions.............. Contested proceeding. D, except portions of This definition is designated
the definition are Category D because it does
NRC. not meet any of the criteria
of Category A, B, or C. A
State may adopt a similar
definition that is compatible
with the orderly pattern of
regulation established by the
Atomic Energy Act, as amended
(Act) and is consistent with
their regulatory authority.
Those portions of the
definition that address areas
reserved to the NRC, e.g., 10
CFR Part 52 activities, are
designated as a Compatibility
Category NRC. A State should
not adopt provisions that
would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
License.............. NRC.................. This definition is designated
Compatibility Category NRC
because it addresses areas
reserved to the NRC. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
For purposes of
compatibility, States should
use the language of the 10
CFR 20.1003 definition,
except those portions of the
definition that reference
areas reserved to the NRC,
e.g., 10 CFR Parts 50, 60,
63, and 72, are designated as
a Compatibility Category NRC.
Licensee............. [D].................. This definition also appears
in 10 CFR 20.1003. For
purposes of compatibility,
the language of the Part 20
definition should be used
where it is assigned to
Compatibility Category D.
[[Page 49462]]
2.100 thru 2.390.............. All of the sections D, except portions of These provisions are
covered by Subparts these provisions are designated Compatibility
A, B, and C. NRC. Category D because they do
not meet any of the criteria
of Category A, B, or C. A
State may adopt similar
provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. Those
portions of the provision
that address areas reserved
to the NRC, e.g., 10 CFR
Parts 50, 51, 52, 53, 54, 55,
60, 63, 72, 73, and 76, are
designated as a Compatibility
Category NRC. A State should
not adopt provisions that
would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Act, 10 CFR 8.4, 10 CFR Part
150, and other Federal laws,
regulations, or provisions.
2.400 thru 2.629.............. All of the sections NRC, for all of the These provisions are
covered by Subparts sections. designated Compatibility
D, E, and F. Category NRC because they
address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
2.800......................... Scope and D, except portions of These provisions are
applicability. these provisions are designated Compatibility
NRC. Category D because they do
not meet any of the criteria
of Category A, B, or C. A
State may adopt similar
provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. Those
portions of the provision
that address areas reserved
to the NRC, e.g., 10 CFR Part
52, are designated as a
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Act, 10 CFR 8.4, 10 CFR
Part 150, and other Federal
laws, regulations, or
provisions.
2.801......................... Initiation of D, except portions of These provisions are
rulemaking. these provisions are designated Compatibility
NRC. Category D because they do
not meet any of the criteria
of Category A, B, or C. A
State may adopt similar
provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. Those
portions of the provision
that address areas reserved
to the NRC, e.g., 10 CFR Part
52, are designated as a
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Act, 10 CFR 8.4, 10 CFR
Part 150, and other Federal
laws, regulations, or
provisions.
2.811......................... Filing of standard NRC.................. This provision is designated
design certification Compatibility Category NRC
application, because it addresses an area
required copies. reserved to the NRC. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
2.813......................... Written NRC.................. This provision is designated
communications. Compatibility Category NRC
because it addresses an area
reserved to the NRC. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
2.815......................... Docketing and NRC.................. This provision is designated
acceptance review. Compatibility Category NRC
because it addresses an area
reserved to the NRC. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
[[Page 49463]]
2.817......................... Withdrawal of NRC.................. This provision is designated a
application. Compatibility Category NRC
because it addresses an area
reserved to the NRC. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
2.819......................... Denial of application NRC.................. This provision is designated
for failure to Compatibility Category NRC
supply information. because it addresses an area
reserved to the NRC. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
2.1202........................ Authority and role of NRC.................. This provision is designated
NRC staff. Compatibility Category NRC
because it addresses an area
reserved to the NRC. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
2.1211--[Removed].............
10 CFR Part 10.................... Criteria and NRC for all sections. These provisions are
procedures for designated Compatibility
determining Category NRC because they
eligibility for address areas reserved to the
access to restricted NRC. A State should not adopt
data or national provisions that would confer
security information regulatory authority to the
or an employment State in an area of exclusive
clearance. NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 19--Notices,
Instructions and Reports to
Workers: Inspection and
Investigations
19.1.......................... Purpose.............. D.................... This provision is designated
Category D because it does
not meet any of the criteria
of Category A, B, or C. A
State may adopt a similar
provision that is compatible
with the orderly pattern of
regulation established by the
Atomic Energy Act, as amended
(Act) and are consistent with
their regulatory authority.
19.2.......................... Scope................ D, except portions of This provision is designated
the provisions in Compatibility Category D
(a)(1), (a)(2), because it does not meet any
(a)(3), and (a)(4) of the criteria of Category
are designated as A, B, or C. A State may adopt
NRC. similar provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. Those
portions of the provision
that address areas reserved
to the NRC, e.g., 10 CFR
Parts 50, 51, 52, 53, 54, 60,
63, 72, and 76, are
designated as a Compatibility
Category NRC. A State should
not adopt provisions that
would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Act, 10 CFR 8.4, 10 CFR Part
150, and other Federal laws,
regulations, or provisions.
19.3--Definitions............. License.............. D, except portions of This definition is designated
the definition are Category D because it does
NRC. not meet any of the criteria
of Category A, B, or C. A
State may adopt a similar
definition that is compatible
with the orderly pattern of
regulation established by the
Atomic Energy Act, as amended
(Act) and is consistent with
their regulatory authority.
Those portions of the
definition that address areas
reserved to the NRC, e.g., 10
CFR Parts 50, 51, 52, 53, 54,
55, 60, 63, 72, 73, and 76,
are designated as a
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
This definition appears in 10
CFR 20.1003. For purposes of
compatibility, States should
use the language of the Part
20 definition, which is
assigned a Compatibility
Category D.
[[Page 49464]]
Regulated activities. D.................... This definition is designated
Category D because it does
not meet any of the criteria
of Category A, B, or C. A
State may adopt a similar
definition that is compatible
with the orderly pattern of
regulation established by the
Atomic Energy Act, as amended
(Act) and is consistent with
their regulatory authority.
Regulated entities... D, except portions of This definition is designated
the definition are Category D because it does
NRC. not meet any of the criteria
of Category A, B, or C. A
State may adopt a similar
definition that is compatible
with the orderly pattern of
regulation established by the
Atomic Energy Act, as amended
(Act) and is consistent with
their regulatory authority.
Those portions of the
definition that address areas
reserved to the NRC are
designated Compatibility
Category NRC. A State should
not adopt provisions that
would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
Worker............... C.................... This definition is designated
Compatibility Category C
because of its role in
effective communication, dose
monitoring, and commerce
(transboundary). A State
should adopt definitions that
are compatible with the
orderly pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. The
essential objectives of this
definition should be adopted.
19.11......................... Posting of Notices to C, except portions of This provision is designated
workers. paragraph (a), and Compatibility Category C
all of paragraphs because it is needed to
(b) and (e) are provide a minimum level of
designated as NRC. information to workers and to
assure that this information
is consistent from one
jurisdiction to another since
workers may work in multiple
jurisdictions. A State should
adopt provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. The
essential objectives of this
definition should be adopted.
Those portions of paragraph
(a) that reference 10 CFR
Part 52 activities, and
paragraphs (b) and (e)
address areas reserved to the
NRC, and are designated
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Act, 10 CFR 8.4, 10 CFR
Part 150, and other Federal
laws, regulations, or
provisions.
19.14......................... Presence of C, except paragraph This provision is designated
representatives of (a) is designated as Compatibility Category C
licensees and NRC. because it is needed to
workers during provide a minimum level of
inspections. consistency from one
jurisdiction to another since
workers may work in multiple
jurisdictions. A State should
adopt provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority.
Paragraph (a) addresses areas
reserved to the NRC, and is
designated Compatibility
Category NRC. A State should
not adopt provisions that
would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Act, 10 CFR 8.4, 10 CFR Part
150, and other Federal laws,
regulations, or provisions.
19.20......................... Employee protection.. D, except portions of This provision is designated
the provision are Compatibility Category D
NRC. because it does not meet any
of the criteria of Category
A, B, or C. A State may adopt
provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. Those
portions of the provision
that address areas reserved
to the NRC, e.g., 10 CFR
Parts 50, 52, 54, 60, 63, 72,
and 76, are designated as a
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Act, 10 CFR 8.4, 10 CFR
Part 150, and other Federal
laws, regulations, or
provisions.
[[Page 49465]]
19.31......................... Application for D.................... This provision is designated
exemptions. Category D because it does
not meet any of the criteria
of Category A, B, or C. A
State may adopt provisions
that are compatible with the
orderly pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority.
19.32......................... Discrimination D.................... This provision is designated
prohibited. Category D because it does
not meet any of the criteria
of Category A, B, or C. A
State may adopt provisions
that are compatible with the
orderly pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority.
10 CFR Part 20--Standards of
Protection
20.1002....................... Scope................ D, except portions of This provision is designated
the provision are Compatibility Category D
designated as NRC. because it does not meet any
of the criteria of Category
A, B, or C. A State may adopt
provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. Those
portions of the provision
that address areas reserved
to the NRC, e.g., 10 CFR
Parts 50, 52, 54, 60, 63, 72,
and 76, are designated as a
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Act, 10 CFR 8.4, 10 CFR
Part 150, and other Federal
laws, regulations, or
provisions.
20.1401....................... General provisions C, except portions of This provision is designated
and scope. the provision are Compatibility Category C
designated as NRC. because it is needed to
provide a minimum level of
consistency regarding
decommissioning activities. A
State should adopt provisions
that are compatible with the
orderly pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. The
essential objectives of these
provisions should be adopted
by States. Those portions of
the provision that address
areas reserved to the NRC,
e.g., 10 CFR Parts 50, 52,
54, 60, 63, and 72, are
designated as a Compatibility
Category NRC. A State should
not adopt provisions that
would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Act, 10 CFR 8.4, 10 CFR Part
150, and other Federal laws,
regulations, or provisions.
20.1406....................... Minimization of C, except portions of This provision is designated
contamination. paragraph (a) and Compatibility Category C
all of paragraph (b) because it is needed to
are designated as provide a minimum level of
NRC. safety regarding
decommissioning activities. A
State should adopt provisions
that are compatible with the
orderly pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. The
essential objectives of these
provisions should be adopted
by States. Those portions of
paragraph (a) that reference
10 CFR Part 52 activities,
and paragraphs (b) address
areas reserved to the NRC,
and are designated
Compatibility Category NRC. A
State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Act, 10 CFR 8.4, 10 CFR
Part 150, and other Federal
laws, regulations, or
provisions.
[[Page 49466]]
20.2203....................... Reports of exposures, C paragraphs (a) and Paragraphs (a) and (b) are
etc., exceeding the (b). designated Compatibility
limits. NRC paragraphs (c) Category C, because they are
and (d). needed to provide a common
understanding in collecting
and reporting information on
the regulation of agreement
material on a nationwide
basis. A State should adopt
provisions that are
compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority. The
essential objectives of these
provisions should be adopted
by States. Paragraphs (c) and
(d) address NRC exclusive
areas of authority are
designated Compatibility
Category NRC, and should not
be adopted by States. A State
should not adopt provisions
that would confer regulatory
authority to the State in an
area of exclusive NRC
jurisdiction pursuant to the
Act, 10 CFR 8.4, 10 CFR Part
150, and other Federal laws,
regulations, or provisions.
10 CFR Part 21.................... Reporting of Defects Not applicable for The provisions in Part 21 are
and Noncompliance. all sections. derived from statutory
authority in the Energy
Reorganization Act, not the
Atomic Energy Act, which does
not apply to Agreement
States. Therefore, this part
cannot be addressed under
either compatibility or
adequacy. While it may be
argued that there are health
and safety reasons to require
States to adopt the
provisions of Part 21, States
may not have the statutory
authority to do so. States
that have the statutory
authority to implement
provisions similar to those
in Part 21 may adopt similar
provisions consistent with
their regulatory authority
but should not address areas
of exclusive NRC
jurisdiction.
10 CFR Part 25.................... Access Authorization. NRC for all sections. These provisions are
designated a Compatibility
Category NRC because they
address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 26.................... Fitness for Duty NRC for all sections. These provisions are
Programs. designated a Compatibility
Category NRC because they
address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 50.................... Domestic Licensing of NRC for all sections. These provisions are
Production and designated a Compatibility
Utilization Category NRC because they
Facilities. address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 51.................... Environmental NRC for all sections. These provisions are
Protection designated a Compatibility
Regulation for Category NRC because they
Domestic Licensing address areas reserved to the
and Related NRC. A State should not adopt
Regulatory Functions. provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 52.................... Licenses, NRC for all sections. These provisions are
Certifications, and designated a Compatibility
Approvals For Category NRC because they
Nuclear Power Plants. address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 54.................... Requirements for NRC for all sections. These provisions are
Renewal of Operating designated a Compatibility
License for Nuclear Category NRC because they
Power Plants. address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
[[Page 49467]]
10 CFR Part 55.................... Operators License.... NRC for all sections. These provisions are
designated a Compatibility
Category NRC because they
address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 72.................... Licensing NRC for all sections. These provisions are
Requirements for designated a Compatibility
Independent Storage Category NRC because they
of Spent Nuclear address areas reserved to the
Fuel and High-level NRC. A State should not adopt
Radioactive Waste provisions that would confer
and Greater than regulatory authority to the
Class C. State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 73.................... Physical Protection NRC for all sections. These provisions are
of Plants and designated a Compatibility
Materials. Category NRC because they
address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 75.................... Safeguards on Nuclear NRC for all sections. These provisions are
Material--Implementa designated a Compatibility
tion of US/IAEA Category NRC because they
Agreement. address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 95.................... Facility Security NRC for all sections. These provisions are
Clearance and designated a Compatibility
Safeguarding of Category NRC because they
National Security address areas reserved to the
Information and NRC. A State should not adopt
Restricted Data. provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 140................... Financial Protection NRC for all sections. These provisions are
Requirements and designated a Compatibility
Indemnity Agreements. Category NRC because they
address areas reserved to the
NRC. A State should not adopt
provisions that would confer
regulatory authority to the
State in an area of exclusive
NRC jurisdiction pursuant to
the Atomic Energy Act, 10 CFR
8.4, 10 CFR Part 150, and
other Federal laws,
regulations, or provisions.
10 CFR Part 170................... Fees for Facilities, D.................... These provisions are
Materials, Import designated a Category D
and Export Licenses, because they do not meet any
and Other Regulatory of the criteria of Category
Services under the A, B, or C. A State may adopt
Atomic Energy Act of similar provisions that are
1954, as Amended. compatible with the orderly
pattern of regulation
established by the Atomic
Energy Act, as amended (Act)
and are consistent with their
regulatory authority.
10 CFR Part 171................... Annual Fees: For D.................... These provisions are
Reactor Licenses and designated a Category D
Fuel Cycle Licenses because they do not meet any
and Material of the criteria of Category
Licenses, Including A, B, or C. A State may adopt
Holders of similar provisions that are
Certificates of compatible with the orderly
Compliance, pattern of regulation
Registrations, and established by the Atomic
Quality Assurance Energy Act, as amended (Act)
Program Approvals and are consistent with their
and Government regulatory authority.
Agencies Licensed by
NRC.
----------------------------------------------------------------------------------------------------------------
IX. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995, Pub.
L. 104-113, requires that Federal agencies use technical standards that
are developed or adopted by voluntary consensus standards bodies unless
using such a standard is inconsistent with applicable law or is
otherwise impractical. In this rule, the NRC is revising the procedural
requirements for early site permits, standard design approvals,
standard design certifications, combined licenses, and manufacturing
licenses to make certain corrections and changes based on the
experience of the previous design certification reviews and on
discussions
[[Page 49468]]
with stakeholders on these licensing processes. These procedural
requirements for rulemaking do not establish standards or substantive
requirements with which all applicants and licensees must comply. In
addition, portions of this rulemaking make conforming changes to
regulatory requirements throughout 10 CFR Chapter I, such as access to
national security information and the procedures governing the conduct
of hearings in proceedings. These changes also do not establish
standards or substantive requirements with which all applicants and
licensees must comply. Finally, portions of this rulemaking make
conforming changes to technical requirements throughout 10 CFR Chapter
I, in order to make clear their applicability to applicants and
licensees under part 52. Inasmuch as the purpose of this rulemaking was
not to establish or fundamentally alter these technical requirements,
the Commission considers it impractical to perform a reassessment of
the fundamental nature of these technical requirements in this
rulemaking. In addition, this rule amends certain portions of the three
design certification regulations in 10 CFR part 52, appendices A, B,
and C (for U.S. ABWR, System 80+, and AP600 designs, respectively).
Design certifications are not generic rulemakings in the sense that
design certifications do not establish standards or requirements with
which all applicants and licensees must comply. Rather, design
certifications are Commission approvals of specific nuclear power plant
designs by rulemaking. Furthermore, design certification rulemakings
are initiated by an applicant for a design certification, rather than
the NRC. For these reasons, the Commission concludes that this action
does not constitute the establishment of a standard that contains
generally applicable requirements.
X. Environmental Impact--Categorical Exclusion
The NRC has determined that these amendments fall within the types
of actions described as categorical exclusions 10 CFR 51.22(c)(1),
(c)(2), and (c)(3). Therefore, neither an environmental impact
statement nor an environmental assessment has been prepared for this
regulation.\19\
---------------------------------------------------------------------------
\19\ When 10 CFR part 52 was issued in 1989, the NRC determined
that the regulation met the eligibility criteria for the categorical
exclusion set forth in 10 CFR 51.22(c)(3). As stated in the Federal
Register notice for the final rule (54 FR 15384; April 18, 1989),
``It makes no substantive difference for the purpose of the
categorical exclusion that the amendments are in a new 10 CFR part
52 rather than in 10 CFR part 50. The amendments are, in fact,
amendments to the 10 CFR part 50 procedures and could have been
placed in that part.'' The categorical exclusion for the current
proposed change to 10 CFR part 52 is consistent with the original
categorical exclusion determination. To ensure that future changes
in part 52 are categorically excluded, this rule contains an
appropriate change to Sec. 51.22(c)(3).
---------------------------------------------------------------------------
XI. Paperwork Reduction Act Statement
This final rule contains new or amended information collection
requirements contained in 10 CFR parts 21, 25, 50, 51, 52, and 54 that
are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These requirements were approved by the Office of Management and
Budget approval numbers 3150-0035, 3150-0046, 3150-0011, 3150-0021,
3150-0151, and 3150-0155. The changes to 10 CFR parts 19, 20, 26, 55,
72, 73, 75, 95, and 140 do not contain new or amended information
collection requirements. Existing requirements were approved by the
Office of Management and Budget, approval numbers 3150-0044, 3150-0014,
3150-0146, 3150-0018, 3150-0132, 3150-0002, 3150-0055, 3150-0047, and
3150-0039.
The burden to the public for the information collections in 10 CFR
part 52 is estimated to average 11,277 hours per response. This
includes the time for reviewing instructions, searching existing data
sources, gathering and maintaining the data needed, and completing and
reviewing the information collection. Send comments on any aspect of
these information collections, including suggestions for reducing the
burden to the records and FOIA/Privacy Services Branch (T-5 F53, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001), or by
Internet electronic mail to [email protected]; and to the Desk
Officer, Office of Information and Regulatory Affairs, NEOB-10202
(3150-0035, 3150-0046, 3150-0011, 3150-0151, and 3150-0155 with revised
information collection requirements), Office of Management and Budget,
Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a request for information or an information collection
requirement unless the requesting document displays a currently valid
OMB control number.
XII. Regulatory Analysis
The Commission has prepared a regulatory analysis on this final
rule. Consistent with the Regulatory Analysis Guidelines, the NRC
performed an aggregate analysis of the rule. The analysis is based on
the assumption that the NRC will receive 19 COL applications during the
next 3 years and 1 COL application per year over the next 17 years. The
net present value of the part 52 rule modifications are estimated to
result in costs to the industry of $58,992 K and $30,952 K using a 3-
percent and a 7-percent discount rate, respectively. The provisions of
the rule relating to part 21 are estimated to result in net present
value costs of $3,873 K and $2,363 K to the industry, using a 3-percent
and a 7-percent discount rate, respectively. The net present value of
the entire rule is estimated to result in net costs to the industry of
$29,726 K and $204 K at a 3-percent and a 7-percent discount rate,
respectively. In addition, the rule is estimated to be a one time net
present value savings to the NRC of $10,443 K.
XIII. Regulatory Flexibility Certification
In accordance with the Regulatory Flexibility Act (5 U.S.C.
605(b)), the Commission certifies that this rule will not have a
significant economic impact on a substantial number of small entities.
This rule affects only the licensing of nuclear power plants. The
companies that will apply for an approval, certification, permit, site
report, or license in accordance with the regulations affected by this
rule do not fall within the scope of the definition of ``small
entities'' set forth in the Regulatory Flexibility Act or the size
standards established by the NRC (10 CFR 2.810).
XIV. Backfit Analysis
The NRC has determined that the backfit rule does not apply to this
rule and, therefore, a backfit analysis is not required, because the
rule does not contain any provisions that would impose backfitting as
defined in the backfit rule, 10 CFR 50.109.
There are no current holders of combined licenses or manufacturing
licenses that are protected by the backfitting restrictions in
Sec. Sec. 50.109, 52.39, 52.98, or 52.171. To the extent that this
rule revises the requirements for future early site permits, standard
design certifications, combined licenses, standard design approvals and
manufacturing licenses for nuclear power plants, these revisions do not
constitute backfits because they are prospective in nature and the
backfit rule is not intended to apply to every NRC action which
substantially changes the expectations of future applicants.
[[Page 49469]]
The NRC issued the first early site permits prior to the effective
date of this final part 52 rule. In addition, there are applications
for early site permits currently being considered by the NRC. As
discussed elsewhere, the NRC has included a ``grandfathering
provision'' in the final part 52 rulemaking which provides that the
early site permit provisions in subpart A of part 52 do not apply to
early site permits whose applications were docketed before the
effective date of the final part 52 rulemaking, unless requested by the
early site permit applicant. This grandfathering provision prohibits
any backfitting for these early site permits.
Other provisions in this rule would apply to currently-approved
standard design approvals and certifications, but they are not
protected by the backfitting restrictions in Sec. 50.104 or Sec.
52.63 because they are either corrections, administrative changes, or
provide additional flexibility to applicants or licensees who might
reference the design approvals or certifications, and thus constitute a
voluntary alternative or relaxation.
Finally, some of the provisions in this rule represent conforming
changes throughout 10 CFR Chapter I which are being made to reflect
Commission adoption of design approvals and design certification
processes which should have been made at the time the Commission first
adopted these processes by rulemaking. While these conforming changes
may, in some cases, affect the way in which a current design
certification or design approval may be referenced, they do not
directly affect the design approval nor are the conforming changes
result in any inconsistency with the finality provisions in the design
certifications or in part 52. Accordingly, the Commission believes that
these conforming changes with respect to design approvals and design
certifications do not raise new backfitting considerations.
XV. Congressional Review Act
Under the Congressional Review Act of 1996, the NRC has determined
that this action is not a major rule and has verified this
determination with the Office of Information and Regulatory Affairs of
OMB.
List of Subjects
10 CFR Part 1
Organization and functions (Government Agencies).
10 CFR Part 2
Administrative practice and procedure, Antitrust, Byproduct
material, Classified information, Environmental protection, Nuclear
materials, Nuclear power plants and reactors, Penalties, Sex
discrimination, Source material, Special nuclear material, Waste
treatment and disposal.
10 CFR Part 10
Administrative practice and procedure, Classified information,
Government employees, Security measures.
10 CFR Part 19
Criminal penalties, Environmental protection, Nuclear materials,
Nuclear power plants and reactors, Occupational safety and health,
Radiation protection, Reporting and recordkeeping requirements, Sex
discrimination.
10 CFR Part 20
Byproduct material, Criminal penalties, Licensed material, Nuclear
materials, Nuclear power plants and reactors, Occupational safety and
health, Packaging and containers, Radiation protection, Reporting and
recordkeeping requirements, Source material, Special nuclear material,
Waste treatment and disposal.
10 CFR Part 21
Nuclear power plants and reactors, Penalties, Radiation protection,
Reporting and recordkeeping requirements.
10 CFR Part 25
Classified information, Criminal penalties, Investigations,
Reporting and recordkeeping requirements, Security measures.
10 CFR Part 26
Alcohol abuse, Alcohol testing, Appeals, Chemical testing, Drug
abuse, Drug testing, Employee assistance programs, Fitness for duty,
Management actions, Nuclear power reactors, Protection of information,
Reporting and recordkeeping requirements.
10 CFR Part 50
Antitrust, Classified information, Criminal penalties, Emergency
Planning, Fire protection, Intergovernmental relations, Nuclear power
plants and reactors, Radiation protection, Reactor siting criteria,
Reporting and recordkeeping requirements.
10 CFR Part 51
Administrative practice and procedure, Environmental impact
statement, Nuclear materials, Nuclear power plants and reactors,
Reporting and recordkeeping requirements.
10 CFR Part 52
Administrative practice and procedure, Antitrust, Backfitting,
Combined license, Early site permit, Emergency planning, Fees,
Inspection, Limited work authorization, Nuclear power plants and
reactors, Probabilistic risk assessment, Prototype, Reactor siting
criteria, Redress of site, Reporting and recordkeeping requirements,
Standard design, Standard design certification.
10 CFR Part 54
Administrative practice and procedure, Age-related degradation,
Backfitting, Classified information, Criminal penalties, Environmental
protection, Nuclear power plants and reactors, Reporting and
recordkeeping requirements.
10 CFR Part 55
Criminal penalties, Manpower training programs, Nuclear power
plants and reactors, Reporting and recordkeeping requirements.
10 CFR Part 72
Administrative practice and procedure, Criminal penalties, Manpower
training programs, Nuclear materials, Occupational safety and health,
Penalties, Radiation protection, Reporting and recordkeeping
requirements, Security measures, Spent fuel, Whistleblowing.
10 CFR Part 73
Criminal penalties, Export, Hazardous materials transportation,
Import, Nuclear materials, Nuclear power plants and reactors, Reporting
and recordkeeping requirements, Security measures.
10 CFR Part 75
Criminal penalties, Intergovernmental relations, Nuclear materials,
Nuclear power plants and reactors, Reporting and recordkeeping
requirements, Security measures.
10 CFR Part 95
Classified information, Criminal penalties, Reporting and
recordkeeping requirements, Security measures.
10 CFR Part 140
Criminal penalties, Extraordinary nuclear occurrence, Insurance,
Intergovernmental relations, Nuclear materials, Nuclear power plants
and reactors, Reporting and recordkeeping requirements.
10 CFR Part 170
Byproduct material, Import and export licenses, Intergovernmental
[[Page 49470]]
relations, Non-payment penalties, Nuclear materials, Nuclear power
plants and reactors, Source material, Special nuclear material.
10 CFR Part 171
Nuclear power plants and reactors.
0
For the reasons set forth in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is adopting
the following amendments to 10 CFR parts 1, 2, 10, 19, 20, 21, 25, 26,
50, 51, 52, 54, 55, 72, 73, 75, 95, 140, 170, and 171.
PART 1--STATEMENT OF ORGANIZATION AND GENERAL INFORMATION
0
1. The authority citation for part 1 continues to read as follows:
Authority: Secs. 23, 161, 68 Stat. 925, 948, as amended (42
U.S.C. 2033, 2201); sec. 29, Pub. L. 85-256, 71 Stat. 579, Pub. L.
95-209, 91 Stat. 1483 (42 U.S.C. 2039); sec. 191, Pub. L. 87-615, 76
Stat. 409 (42 U.S.C. 2241); secs. 201, 203, 204, 205, 209, 88 Stat.
1242, 1244, 1245, 1246, 1248, as amended (42 U.S.C. 5841, 5843,
5844, 5845, 5849); 5 U.S.C. 552, 553; Reorganization Plan No. 1 of
1980, 45 FR 40561, June 16, 1980.
0
2. In Sec. 1.43, paragraph (a)(2) is revised to read as follows:
Sec. 1.43 Office of Nuclear Reactor Regulation.
* * * * *
(a) * * *
(2) Receipt, possession, and ownership of source, byproduct, and
special nuclear material used or produced at facilities licensed under
10 CFR parts 50, 52, and 54;
* * * * *
PART 2--RULES OF PRACTICE FOR DOMESTIC LICENSING PROCEEDINGS AND
ISSUANCE OF ORDERS
0
3. The authority citation for part 2 continues to read as follows:
Authority: Secs. 161, 181, 68 Stat. 948, 953, as amended (42
U.S.C. 2201, 2231); sec. 191, as amended, Pub. L. 87-615, 76 Stat.
409 (42 U.S.C. 2241); sec. 201, 88 Stat. 1242, as amended (42 U.S.C.
5841); 5 U.S.C. 552; sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504
note).
Section 2.101 also issued under secs. 53, 62, 63, 81, 103, 104,
105, 68 Stat. 930, 932, 933, 935, 936, 937, 938, as amended (42
U.S.C. 2073, 2092, 2093, 2111, 2133, 2134, 2135); sec. 114(f), Pub.
L. 97-425, 96 Stat. 2213, as amended (42 U.S.C. 10143(o)), sec. 102,
Pub. L. 91-190, 83 Stat. 853, as amended (42 U.S.C. 4332); sec. 301,
88 Stat. 1248 (42 U.S.C. 5871). Sections 2.102, 2.103, 2.104, 2.105,
2.721 also issued under secs. 102, 104, 105, 163, 183i, 189, 68
Stat. 936, 937, 938, 954, 955, as amended (42 U.S.C. 2132, 2133,
2134, 2135, 2233, 2239). Sections 2.105 also issued under Pub. L.
97-415, 96 Stat. 2073 (42 U.S.C. 2239). Sections 2.200-2.206 also
issued under secs. 161 b, i, o, 182, 186, 234, 68 Stat. 948-951,
955, 83 Stat. 444, as amended (42 U.S.C. 2201 (b), (i), (o), 2236,
2282); sec. 206, 88 Stat 1246 (42 U.S.C. 5846). Section 2.205(j)
also issued under Pub. L. 101-410, 104 Stat. 90, as amended by
Section 3100(s), Pub. L. 104-134, 110 Stat. 1321-373 (28 U.S.C. 2461
note). Subpart C also issued under sec. 189, 68 Stat. 955 (42 U.S.C.
2239). Sections 2.600-2.606 also issued under sec. 102, Pub. L. 91-
190, 83 Stat. 853, as amended (42 U.S.C. 4332). Section 2.700a also
issued under 5 U.S.C. 554. Sections 2.343, 2.346, 2.754, 2.712 also
issued under 5 U.S.C. 557. Section 2.764 also issued under secs.
135, 141, Pub. L. 97-425, 96 Stat. 2232, 2241 (42 U.S.C. 10155,
10161). Section 3.790 also issued under sec. 103, 68 Stat. 936, as
amended (42 U.S.C. 2133), and 5 U.S.C. 552. Sections 2.800 and 2.808
also issued under 5 U.S.C. 553. Section 2.809 also issued under 5
U.S.C. 553, and sec. 29, Pub. L. 85-256, 71 Stat. 579, as amended
(42 U.S.C. 2039). Subpart K also issued under sec. 189, 68 Stat. 955
(42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42 U.S.C.
10154). Subpart L also issued under sec. 189, 68 Stat. 955 (42
U.S.C. 2239). Subpart M also issued under sec. 184 (42 U.S.C. 2234)
and sec. 189, 68 Stat. 955 (42 U.S.C. 2239). Subpart N also issued
under sec. 189, 68 Stat. 955 (42 U.S.C. 2239). Appendix A also
issued under sec. 6, Pub. L. 91-550, 84 Stat. 1473 (42 U.S.C. 2135).
0
4. In Sec. 2.1, paragraphs (c) and (d) are revised and a new paragraph
(e) is added to read as follows:
Sec. 2.1 Scope.
* * * * *
(c) Imposing civil penalties under Section 234 of the Act;
(d) Rulemaking under the Act and the Administrative Procedure Act;
and
(e) Standard design approvals under part 52 of this chapter.
0
5. In Sec. 2.4, the definitions of contested proceeding, license and
licensee are revised to read as follows:
Sec. 2.4 Definitions.
* * * * *
Contested proceeding means--
(1) A proceeding in which there is a controversy between the NRC
staff and the applicant for a license or permit concerning the issuance
of the license or permit or any of the terms or conditions thereof;
(2) A proceeding in which the NRC is imposing a civil penalty or
other enforcement action, and the subject of the civil penalty or
enforcement action is an applicant for or holder of a license or
permit, or is or was an applicant for a standard design certification
under part 52 of this chapter; and
(3) A proceeding in which a petition for leave to intervene in
opposition to an application for a license or permit has been granted
or is pending before the Commission.
* * * * *
License means a license, including an early site permit,
construction permit, operating license, combined license, manufacturing
license, or renewed license issued by the Commission.
Licensee means a person who is authorized to conduct activities
under a license.
* * * * *
0
6. The heading of Subpart A is revised to read as follows:
Subpart A--Procedure for Issuance, Amendment, Transfer, or Renewal
of a License, and Standard Design Approval
0
7. Section 2.100 is revised to read as follows:
Sec. 2.100 Scope of subpart.
This subpart prescribes the procedure for issuance of a license;
amendment of a license at the request of the licensee; transfer and
renewal of a license; and issuance of a standard design approval under
subpart E of part 52 of this chapter.
0
8. In Sec. 2.101, paragraphs (a)(1), (a)(2), the introductory
paragraph of (a)(3), paragraph (a)(3)(ii), paragraph (a)(4), paragraph
(a)(5), and paragraph (a-1) are revised to read as follows:
Sec. 2.101 Filing of application.
(a)(1) An application for a permit, a license, a license transfer,
a license amendment, a license renewal, or a standard design approval,
shall be filed with the Director of New Reactors or Director of Nuclear
Reactor Regulation or Director of Nuclear Material Safety and
Safeguards, as prescribed by the applicable provisions of this chapter.
A prospective applicant may confer informally with the NRC staff before
filing an application.
(2) Each application for a license for a facility or for receipt of
waste radioactive material from other persons for the purpose of
commercial disposal by the waste disposal licensee will be assigned a
docket number. However, to allow a determination as to whether an
application for a construction permit, operating license, early site
permit, standard design approval, combined license, or manufacturing
license for a production or utilization facility is complete and
acceptable for docketing, it will be initially treated as a tendered
application. A copy of the tendered application will be available for
public inspection at the NRC Web site, http://www.nrc.gov, and/or at
the NRC
[[Page 49471]]
Public Document Room. Generally, the determination on acceptability for
docketing will be made within a period of 30 days. However, in selected
applications, the Commission may decide to determine acceptability
based on the technical adequacy of the application as well as its
completeness. In these cases, the Commission, under Sec. 2.104(a),
will direct that the notice of hearing be issued as soon as practicable
after the application has been tendered, and the determination of
acceptability will be made generally within a period of 60 days. For
docketing and other requirements for applications under part 61 of this
chapter, see paragraph (g) of this section.
(3) If the Director of New Reactors, Director of Nuclear Reactor
Regulation, or Director of Nuclear Material Safety and Safeguards, as
appropriate, determines that a tendered application for a construction
permit, operating license, early site permit, standard design approval,
combined license, or manufacturing license for a production or
utilization facility, and/or any environmental report required under
subpart A of part 51 of this chapter, or part thereof as provided in
paragraphs (a)(5) or (a-1) of this section are complete and acceptable
for docketing, a docket number will be assigned to the application or
part thereof, and the applicant will be notified of the determination.
With respect to the tendered application and/or environmental report or
part thereof that is acceptable for docketing, the applicant will be
requested to:
* * * * *
(ii) Serve a copy on the chief executive of the municipality in
which the facility or site which is the subject of an early site permit
is to be located or, if the facility or site which is the subject of an
early site permit is not to be located within a municipality, on the
chief executive of the county, and serve a notice of availability of
the application or environmental report on the chief executives of the
municipalities or counties which have been identified in the
application or environmental report as the location of all or part of
the alternative sites, containing the following information, as
applicable: Docket number of the application, a brief description of
the proposed site and facility; the location of the site and facility
as primarily proposed and alternatively listed; the name, address,
telephone number, and email address (if available) of the applicant's
representative who may be contacted for further information;
notification that a draft environmental impact statement will be issued
by the Commission and will be made available upon request to the
Commission; and notification that if a request is received from the
appropriate chief executive, the applicant will transmit a copy of the
application and environmental report, and any changes to these
documents which affect the alternative site location, to the executive
who makes the request. In complying with the requirements of this
paragraph, the applicant should not make public distribution of those
parts of the application subject to Sec. 2.390(d). The applicant shall
submit to the Director of New Reactors or the Director of Nuclear
Reactor Regulation an affidavit that service of the notice of
availability of the application or environmental report has been
completed along with a list of names and addresses of those executives
upon whom the notice was served; and
* * * * *
(4) The tendered application for a construction permit, operating
license, early site permit, standard design approval, combined license,
or manufacturing license will be formally docketed upon receipt by the
Director of New Reactors, Director of Nuclear Reactor Regulation, or
Director of Nuclear Material Safety and Safeguards, as appropriate, of
the required additional copies. Distribution of the additional copies
shall be deemed to be complete as of the time the copies are deposited
in the mail or with a carrier prepaid for delivery to the designated
addresses. The date of docketing shall be the date when the required
copies are received by the Director of New Reactors, Director of
Nuclear Reactor Regulation, or Director of Nuclear Material Safety and
Safeguards, as appropriate. Within 10 days after docketing, the
applicant shall submit to the Director of New Reactors, Director of
Nuclear Reactor Regulation, or Director of Nuclear Material Safety and
Safeguards, as appropriate, an affidavit that distribution of the
additional copies to Federal, State, and local officials has been
completed in accordance with the requirements of this chapter and
written instructions furnished to the applicant by the Director of New
Reactors, Director of Nuclear Reactor Regulation, or Director of
Nuclear Material Safety and Safeguards, as appropriate. Amendments to
the application and environmental report shall be filed and distributed
and an affidavit shall be furnished to the Director of New Reactors,
Director of Nuclear Reactor Regulation, or Director of Nuclear Material
Safety and Safeguards, as appropriate, in the same manner as for the
initial application and environmental report. If it is determined that
all or any part of the tendered application and/or environmental report
is incomplete and therefore not acceptable for processing, the
applicant will be informed of this determination, and the respects in
which the document is deficient.
(5) An applicant for a construction permit under part 50 of this
chapter or a combined license under part 52 of this chapter for a
production or utilization facility which is subject to Sec. 51.20(b)
of this chapter, and is of the type specified in Sec. 50.21(b)(2) or
(3) or Sec. 50.22 of this chapter or is a testing facility may submit
the information required of applicants by part 50 or part 52 of the
chapter in two parts. One part shall be accompanied by the information
required by Sec. 50.30(f) of this chapter, or Sec. 52.80(b) of this
chapter, as applicable. The other part shall include any information
required by Sec. 50.34(a) and, if applicable, Sec. 50.34a of this
chapter, or Sec. Sec. 52.79 and 52.80(a), as applicable. One part may
precede or follow other parts by no longer than 6 months. If it is
determined that either of the parts as described above is incomplete
and not acceptable for processing, the Director of New Reactors,
Director of Nuclear Reactor Regulation, or Director of Nuclear Material
Safety and Safeguards, as appropriate, will inform the applicant of
this determination and the respects in which the document is deficient.
Such a determination of completeness will generally be made within a
period of 30 days. Whichever part is filed first shall also include the
fee required by Sec. Sec. 50.30(e) and 170.21 of this chapter and the
information required by Sec. Sec. 50.33, 50.34(a)(1) or 52.79(a)(1),
as applicable, and Sec. 50.37 of this chapter. The Director of New
Reactors, Director of Nuclear Reactor Regulation, or Director of
Nuclear Material Safety and Safeguards, as appropriate, will accept for
docketing an application for a construction permit under part 50 or a
combined license under part 52 for a production or utilization facility
which is subject to Sec. 51.20(b) of this chapter, and is of the type
specified in Sec. 50.21(b)(2) or (3) or Sec. 50.22 of this chapter or
is a testing facility where one part of the application as described
above is complete and conforms to the requirements of part 50 of this
chapter. The additional parts will be docketed upon a determination by
the Director of New Reactors, Director of Nuclear Reactor Regulation,
or Director of Nuclear Material Safety and Safeguards, as appropriate,
that it is complete.
(a-1) Early consideration of site suitability issues. An applicant
for a
[[Page 49472]]
construction permit under part 50 of this chapter or a combined license
under part 52 of this chapter for a utilization facility which is
subject to Sec. 51.20(b) of this chapter and is of the type specified
in Sec. 50.21(b)(2) or (3) or Sec. 50.22 of this chapter or is a
testing facility, may request that the Commission conduct an early
review and hearing and render an early partial decision in accordance
with subpart F of this part on issues of site suitability within the
purview of the applicable provisions of parts 50, 51, 52, and 100 of
this chapter.
(1) Construction permit. The applicant for the construction permit
may submit the information required of applicants by the provisions of
this chapter in three parts:
(i) Part one shall include or be accompanied by any information
required by Sec. Sec. 50.34(a)(1) and 50.30(f) of this chapter which
relates to the issue(s) of site suitability for which an early review,
hearing, and partial decision are sought, except that information with
respect to operation of the facility at the projected initial power
level need not be supplied, and shall include the information required
by Sec. Sec. 50.33(a) through (e) and 50.37 of this chapter. The
information submitted shall also include:
(A) Proposed findings on the issues of site suitability on which
the applicant has requested review and a statement of the bases or the
reasons for those findings,
(B) A range of postulated facility design and operation parameters
that is sufficient to enable the Commission to perform the requested
review of site suitability issues under the applicable provisions of
parts 50, 51, and 100, and
(C) Information concerning the applicant's site selection process
and long-range plans for ultimate development of the site required by
Sec. 2.603(b)(1).
(ii) Part two shall include or be accompanied by the remaining
information required by Sec. Sec. 50.30(f), 50.33, and 50.34(a)(1) of
this chapter.
(iii) Part three shall include the remaining information required
by Sec. Sec. 50.34a and (in the case of a nuclear power reactor)
50.34(a) of this chapter.
(iv) The information required for part two or part three shall be
submitted during the period the partial decision on part one is
effective. Submittal of the information required for part three may
precede by no more than 6 months or follow by no more than 6 months the
submittal of the information required for part two.
(2) Combined license under part 52. An applicant for a combined
license under part 52 of this chapter may submit the information
required of applicants by the provisions of this chapter in three
parts:
(i) Part one shall include or be accompanied by any information
required by Sec. Sec. 52.79(a)(1) and 50.30(f) of this chapter which
relates to the issue(s) of site suitability for which an early review,
hearing, and partial decision are sought, except that information with
respect to operation of the facility at the projected initial power
level need not be supplied, and shall include the information required
by Sec. Sec. 50.33(a) through (e) and 50.37 of this chapter. The
information submitted shall also include:
(A) Proposed findings on the issues of site suitability on which
the applicant has requested review and a statement of the bases or the
reasons for those findings;
(B) A range of postulated facility design and operation parameters
that is sufficient to enable the Commission to perform the requested
review of site suitability issues under the applicable provisions of
parts 50, 51, 52, and 100; and
(C) Information concerning the applicant's site selection process
and long-range plans for ultimate development of the site required by
Sec. 2.621(b)(1).
(ii) Part two shall include or be accompanied by the remaining
information required by Sec. Sec. 50.30(f), 50.33, and 52.79(a)(1) of
this chapter.
(iii) Part three shall include the remaining information required
by Sec. Sec. 52.79 and 52.80 of this chapter.
(iv) The information required for part two or part three shall be
submitted during the period the partial decision on part one is
effective. Submittal of the information required for part three may
precede by no more than 6 months or follow by no more than 6 months the
submittal of the information required for part two.
* * * * *
0
9. In Sec. 2.102, paragraph (a) is revised to read as follows:
Sec. 2.102 Administrative review of application.
(a) During review of an application by the NRC staff, an applicant
may be required to supply additional information. The staff may request
any one party to the proceeding to confer with the staff informally. In
the case of a docketed application for a construction permit, operating
license, early site permit, standard design approval, combined license,
or manufacturing license of this chapter, the staff shall establish a
schedule for its review of the application, specifying the key
intermediate steps from the time of docketing until the completion of
its review.
* * * * *
0
10. Section 2.104 is revised to read as follows:
Sec. 2.104 Notice of hearing.
(a) In the case of an application on which a hearing is required by
the Act or this chapter, or in which the Commission finds that a
hearing is required in the public interest, the Secretary will issue a
notice of hearing to be published in the Federal Register. The notice
must be published at least 15 days, and in the case of an application
concerning a construction permit, early site permit, or combined
license for a facility of the type described in Sec. 50.21(b) or Sec.
50.22 of this chapter or a testing facility, at least 30 days before
the date set for hearing in the notice.\1\ In addition, in the case of
an application for a construction permit, early site permit, or
combined license for a facility of the type described in Sec. 50.22 of
this chapter, or a testing facility, the notice must be issued as soon
as practicable after the NRC has docketed the application; provided,
that if the NRC decides, under Sec. 2.101(a)(2), to determine the
acceptability of the application based upon its technical adequacy as
well as completeness, the notice shall be issued as soon as practicable
after the application has been tendered.
---------------------------------------------------------------------------
\1\ If the notice of hearing concerning an application for a
construction permit, early site permit, or combined license for a
facility of the type described in Sec. 50.22 of this chapter or a
testing facility does not specify the time and place of initial
hearing, a subsequent notice will be published in the Federal
Register which will provide at least 30 days notice of the time and
place of that hearing. After this notice is given, the presiding
officer may reschedule the commencement of the initial hearing for a
later date or reconvene a recessed hearing without again providing
at least 30 days notice.
---------------------------------------------------------------------------
(b) The notice of hearing must state:
(1) The nature of the hearing;
(2) The authority under which the hearing is to be held;
(3) The matters of fact and law to be considered;
(4) The date by which requests for hearing or petitions to
intervene must be filed;
(5) The presiding officer designated for the hearing, or the
procedure that the Commission will use to designate a presiding officer
for the hearing.
(c)(1) The Secretary will transmit a notice of hearing on an
application for a license for a production or utilization facility
including an early site permit, combined license (but not for a
manufacturing license), for a license for
[[Page 49473]]
receipt of waste radioactive material from other persons for the
purpose of commercial disposal by the waste disposal licensee, for a
license under part 61 of this chapter, for a construction authorization
for an HLW repository at a geologic repository operations area under
parts 60 or 63 of this chapter, for a license to receive and possess
high-level radioactive waste at a geologic repository operations area
under parts 60 or 63 of this chapter, and for a license under part 72
of this chapter to acquire, receive or possess spent fuel for the
purpose of storage in an independent spent fuel storage installation
(ISFSI) to the governor or other appropriate official of the State and
to the chief executive of the municipality in which the facility is to
be located or the activity is to be conducted or, if the facility is
not to be located or the activity conducted within a municipality, to
the chief executive of the county (or to the Tribal organization, if it
is to be located or conducted within an Indian reservation).
(2) The Secretary will transmit a notice of hearing on an
application for a license under part 72 of this chapter to acquire,
receive or possess spent fuel, high-level radioactive waste or
radioactive material associated with high-level radioactive waste for
the purpose of storage in a monitored retrievable storage installation
(MRS) to the same persons who received the notice of docketing under
Sec. 72.16(e) of this chapter.
0
11. In Sec. 2.105, the introductory text of paragraphs (a) and (a)(4)
are revised, and paragraphs (a)(12), (a)(13), and (b)(3) are added to
read as follows:
Sec. 2.105 Notice of proposed action.
(a) If a hearing is not required by the Act or this chapter, and if
the Commission has not found that a hearing is in the public interest,
it will, before acting thereon, publish in the Federal Register, as
applicable, either a notice of intended operation under Sec. 52.103(a)
of this chapter and a proposed finding that inspections, tests,
analysis, and acceptance criteria for a combined license under subpart
C of part 52 have been or will be met, or a notice of proposed action
with respect to an application for:
* * * * *
(4) An amendment to an operating license, combined license, or
manufacturing license for a facility licensed under Sec. Sec. 50.21(b)
or 50.22 of this chapter, or for a testing facility, as follows:
* * * * *
(12) An amendment to an early site permit issued under subpart A of
part 52 of this chapter, as follows:
(i) If the early site permit does not provide authority to conduct
the activities allowed under Sec. 50.10(e)(1) of this chapter, the
amendment will involve no significant hazards consideration, and though
the NRC will provide notice of opportunity for a hearing under this
section, it may make the amendment immediately effective and grant a
hearing thereafter; and
(ii) If the early site permit provides authority to conduct the
activities allowed under Sec. 50.10(e)(1) and the Commission
determines under Sec. Sec. 50.58 and 50.91 of this chapter that an
emergency situation exists or that exigent circumstances exist and that
the amendment involves no significant hazards consideration, it will
provide notice of opportunity for a hearing under Sec. 2.106 of this
chapter (if a hearing is requested, which will be held after issuance
of the amendment).
(13) A manufacturing license under subpart F of part 52 of this
chapter.
(b) * * *
(3) For a notice of intended operation under Sec. 52.103(a) of
this chapter, the following information:
(i) The identification of the NRC action as making the finding
required under Sec. 52.103(g) of this chapter;
(ii) The manner in which the licensee notifications under 10 CFR
52.99(c) which are required to be made available by 10 CFR 52.99(e)(2)
may be obtained and examined;
(iii) The manner in which copies of the safety analysis may be
obtained and examined; and
(iv) Any conditions, limitations, or restrictions to be placed on
the license in connection with the finding under Sec. 52.103(g) of
this chapter, and the expiration date or circumstances (if any) under
which the conditions, limitations or restrictions will no longer apply.
* * * * *
0
12. In Sec. 2.106, paragraphs (a) and (b) are revised to read as
follows:
Sec. 2.106 Notice of issuance.
(a) The Director of New Reactors, Director of Nuclear Reactor
Regulation, or Director of Nuclear Material Safety and Safeguards, as
appropriate, will inform the State and local officials specified in
Sec. 2.104(e) and publish a document in the Federal Register
announcing the issuance of:
(1) A license or an amendment of a license for which a notice of
proposed action has been previously published;
(2) An amendment of a license for a facility of the type described
in Sec. 50.21(b) or Sec. 50.22 of this chapter, or a testing
facility, whether or not a notice of proposed action has been
previously published; and
(3) The finding under Sec. 52.103(g) of this chapter.
(b) The notice of issuance will set forth:
(1) In the case of a license or amendment:
(i) The nature of the license or amendment;
(ii) The manner in which copies of the safety analysis, if any, may
be obtained and examined; and
(iii) A finding that the application for the license or amendment
complies with the requirements of the Act and this chapter.
(2) In the case of a finding under Sec. 52.103(g) of this chapter:
(i) The manner in which copies of the safety analysis, if any, may
be obtained and examined; and
(ii) A finding that the prescribed inspections, tests, and analyses
have been performed, the prescribed acceptance criteria have been met,
and that the license complies with the requirements of the Act and this
chapter.
* * * * *
0
13. Section 2.109 is revised to read as follows:
Sec. 2.109 Effect of timely renewal application.
(a) Except for the renewal of an operating license for a nuclear
power plant under 10 CFR 50.21(b) or 50.22, an early site permit under
subpart A of part 52 of this chapter, a manufacturing license under
subpart F of part 52 of this chapter, or a combined license under
subpart C of part 52 of this chapter, if at least 30 days before the
expiration of an existing license authorizing any activity of a
continuing nature, the licensee files an application for a renewal or
for a new license for the activity so authorized, the existing license
will not be deemed to have expired until the application has been
finally determined.
(b) If the licensee of a nuclear power plant licensed under 10 CFR
50.21(b) or 50.22 files a sufficient application for renewal of either
an operating license or a combined license at least 5 years before the
expiration of the existing license, the existing license will not be
deemed to have expired until the application has been finally
determined.
(c) If the holder of an early site permit licensed under subpart A
of part 52 of this chapter files a sufficient application for renewal
under Sec. 52.29 of this chapter at least 12 months before the
expiration of the existing early site permit, the
[[Page 49474]]
existing permit will not be deemed to have expired until the
application has been finally determined.
(d) If the licensee of a manufacturing license under subpart F of
part 52 of this chapter files a sufficient application for renewal
under Sec. 52.177 of this chapter at least 12 months before the
expiration of the existing license, the existing license will not be
deemed to have expired until the application has been finally
determined.
0
14. Section 2.110 is revised to read as follows:
Sec. 2.110 Filing and administrative action on submittals for
standard design approval or early review of site suitability issues.
(a)(1) A submittal for a standard design approval under subpart E
of part 52 of this chapter shall be subject to Sec. Sec. 2.101(a) and
2.390 to the same extent as if it were an application for a permit or
license.
(2) Except as specifically provided otherwise by the provisions of
appendix Q to parts 50 of this chapter, a submittal for early review of
site suitability issues under appendix Q to parts 50 of this chapter
shall be subject to Sec. Sec. 2.101(a)(2) through (4) to the same
extent as if it were an application for a permit or license.
(b) Upon initiation of review by the NRC staff of a submittal for
an early review of site suitability issues under appendix Q of parts 50
of this chapter, or for a standard design approval under subpart E of
part 52 of this chapter, the Director of New Reactors or the Director
of Nuclear Reactor Regulation shall publish in the Federal Register a
notice of receipt of the submittal, inviting comments from interested
persons within 60 days of publication or other time as may be
specified, for consideration by the NRC staff and ACRS in their review.
(c)(1) Upon completion of review by the NRC staff and the ACRS of a
submittal for a standard design approval, the Director of New Reactors
or the Director of the Office of Nuclear Reactor Regulation shall
publish in the Federal Register a determination as to whether or not
the design is acceptable, subject to terms and conditions as may be
appropriate, and shall make available at the NRC Web site, http://www.nrc.gov, a report that analyzes the design.
(2) Upon completion of review by the NRC staff and, if appropriate
by the ACRS, of a submittal for early review of site suitability
issues, the NRC staff shall prepare a staff site report which shall
identify the location of the site, state the site suitability issues
reviewed, explain the nature and scope of the review, state the
conclusions of the staff regarding the issues reviewed and state the
reasons for those conclusions. Upon issuance of an NRC staff site
report, the NRC staff shall publish a notice of the availability of the
report in the Federal Register and shall make the report available at
the NRC Web site, http://www.nrc.gov. The NRC staff shall also send a
copy of the report to the Governor or other appropriate official of the
State in which the site is located, and to the chief executive of the
municipality in which the site is located or, if the site is not
located in a municipality, to the chief executive of the county.
0
15. Section 2.111 is revised to read as follows:
Sec. 2.111 Prohibition of sex discrimination.
No person shall on the grounds of sex be excluded from
participation in, be denied a license, standard design approval, or
petition for rulemaking (including a design certification), be denied
the benefits of, or be subjected to discrimination under any program or
activity carried on or receiving Federal assistance under the Act or
the Energy Reorganization Act of 1974.
0
16. In Sec. 2.202, paragraph (e) is revised to read as follows:
Sec. 2.202 Orders.
* * * * *
(e)(1) If the order involves the modification of a part 50 license
and is a backfit, the requirements of Sec. 50.109 of this chapter
shall be followed, unless the licensee has consented to the action
required.
(2) If the order involves the modification of combined license
under subpart C of part 52 of this chapter, the requirements of Sec.
52.98 of this chapter shall be followed unless the licensee has
consented to the action required.
(3) If the order involves a change to an early site permit under
subpart A of part 52 of this chapter, the requirements of Sec. 52.39
of this chapter must be followed, unless the applicant or licensee has
consented to the action required.
(4) If the order involves a change to a standard design
certification rule referenced by that plant's application, the
requirements, if any, in the referenced design certification rule with
respect to changes must be followed, or, in the absence of these
requirements, the requirements of Sec. 52.63 of this chapter must be
followed, unless the applicant or licensee has consented to follow the
action required.
(5) If the order involves a change to a standard design approval
referenced by that plant's application, the requirements of Sec.
52.145 of this chapter must be followed unless the applicant or
licensee has consented to follow the action required.
(6) If the order involves a modification of a manufacturing license
under subpart F of part 52, the requirements of Sec. 52.171 of this
chapter must be followed, unless the applicant or licensee has
consented to the action required.
0
17. In Sec. 2.309, paragraphs (a), (f)(1)(i), (f)(1)(v), and
(f)(1)(vi) are revised, a new paragraph (f)(1)(vii) is added, and
paragraphs (g), (h)(2), and (i) are revised to read as follows:
Sec. 2.309 Hearing requests, petitions to intervene, requirements for
standing, and contentions.
(a) General requirements. Any person whose interest may be affected
by a proceeding and who desires to participate as a party must file a
written request for hearing and a specification of the contentions
which the person seeks to have litigated in the hearing. In a
proceeding under 10 CFR 52.103, the Commission, acting as the presiding
officer, will grant the request if it determines that the requestor has
standing under the provisions of paragraph (d) of this section and has
proposed at least one admissible contention that meets the requirements
of paragraph (f) of this section. For all other proceedings, except as
provided in paragraph (e) of this section, the Commission, presiding
officer, or the Atomic Safety and Licensing Board designated to rule on
the request for hearing and/or petition for leave to intervene, will
grant the request/petition if it determines that the requestor/
petitioner has standing under the provisions of paragraph (d) of this
section and has proposed at least one admissible contention that meets
the requirements of paragraph (f) of this section. In ruling on the
request for hearing/petition to intervene submitted by petitioners
seeking to intervene in the proceeding on the HLW repository, the
Commission, the presiding officer, or the Atomic Safety and Licensing
Board shall also consider any failure of the petitioner to participate
as a potential party in the pre-license application phase under subpart
J of this part in addition to the factors in paragraph (d) of this
section. If a request for hearing or petition to intervene is filed in
response to any notice of hearing or opportunity for hearing, the
applicant/licensee shall be deemed to be a party.
* * * * *
(f) * * *
(1) * * *
[[Page 49475]]
(i) Provide a specific statement of the issue of law or fact to be
raised or controverted, provided further, that the issue of law or fact
to be raised in a request for hearing under 10 CFR 52.103(b) must be
directed at demonstrating that one or more of the acceptance criteria
in the combined license have not been, or will not be met, and that the
specific operational consequences of nonconformance would be contrary
to providing reasonable assurance of adequate protection of the public
health and safety;
* * * * *
(v) Provide a concise statement of the alleged facts or expert
opinions which support the requestor's/petitioner's position on the
issue and on which the petitioner intends to rely at hearing, together
with references to the specific sources and documents on which the
requestor/petitioner intends to rely to support its position on the
issue;
(vi) In a proceeding other than one under 10 CFR 52.103, provide
sufficient information to show that a genuine dispute exists with the
applicant/licensee on a material issue of law or fact. This information
must include references to specific portions of the application
(including the applicant's environmental report and safety report) that
the petitioner disputes and the supporting reasons for each dispute,
or, if the petitioner believes that the application fails to contain
information on a relevant matter as required by law, the identification
of each failure and the supporting reasons for the petitioner's belief;
and
(vii) In a proceeding under 10 CFR 52.103(b), the information must
be sufficient, and include supporting information showing, prima facie,
that one or more of the acceptance criteria in the combined license
have not been, or will not be met, and that the specific operational
consequences of nonconformance would be contrary to providing
reasonable assurance of adequate protection of the public health and
safety. This information must include the specific portion of the
report required by 10 CFR 52.99(c) which the requestor believes is
inaccurate, incorrect, and/or incomplete (i.e., fails to contain the
necessary information required by Sec. 52.99(c)). If the requestor
identifies a specific portion of the Sec. 52.99(c) report as
incomplete and the requestor contends that the incomplete portion
prevents the requestor from making the necessary prima facie showing,
then the requestor must explain why this deficiency prevents the
requestor from making the prima facie showing.
* * * * *
(g) Selection of hearing procedures. A request for hearing and/or
petition for leave to intervene may, except in a proceeding under 10
CFR 52.103, also address the selection of hearing procedures, taking
into account the provisions of Sec. 2.310. If a request/petition
relies upon Sec. 2.310(d), the request/petition must demonstrate, by
reference to the contention and the bases provided and the specific
procedures in subpart G of this part, that resolution of the contention
necessitates resolution of material issues of fact which may be best
determined through the use of the identified procedures.
(h) * * *
(2) Except in a proceeding under 10 CFR 52.103, the requestor/
petitioner may file a reply to any answer. The reply must be filed
within 7 days after service of that answer.
* * * * *
(i) Decision on request/petition. In all proceedings other than a
proceeding under 10 CFR 52.103, the presiding officer shall, within 45
days after the filing of answers and replies under paragraph (h) of
this section, issue a decision on each request for hearing/petition to
intervene, absent an extension from the Commission. The Commission,
acting as the presiding officer, shall expeditiously grant or deny the
request for hearing in a proceeding under 10 CFR 52.103. The
Commission's decision may not be the subject of any appeal under 10 CFR
2.311.
0
18. In Sec. 2.310, paragraph (j) is redesignated as paragraph (k), and
a new paragraph (j) is added to read as follows:
Sec. 2.310 Selection of hearing procedures.
* * * * *
(j) Proceedings on a Commission finding under 10 CFR 52.103(c) and
(g) shall be conducted in accordance with the procedures designated by
the Commission in each proceeding.
* * * * *
0
19. In Sec. 2.339, paragraph (d) is revised to read as follows:
Sec. 2.339 Expedited decisionmaking procedure.
* * * * *
(d) The provisions of this section do not apply to an initial
decision directing the issuance of a limited work authorization under
10 CFR 50.10, an early site permit under subpart A of part 52 of this
chapter, a construction permit or construction authorization, a
combined license under subpart C of part 52 of this chapter, or a
manufacturing license under subpart F of part 52.
0
20. Section 2.340 is revised to read as follows:
Sec. 2.340 Initial decision in certain contested proceedings;
immediate effectiveness of initial decisions; issuance of
authorizations, permits, and licenses.
(a) Initial decision--production or utilization facility operating
license. In any initial decision in a contested proceeding on an
application for an operating license (including an amendment to or
renewal of an operating license) for a production or utilization
facility, the presiding officer shall make findings of fact and
conclusions of law on the matters put into controversy by the parties
to the proceeding, any matter designated by the Commission to be
decided by the presiding officer, and any matter not put into
controversy by the parties, but only to the extent that the presiding
officer determines that a serious safety, environmental, or common
defense and security matter exists, and the Commission approves of an
examination of and decision on the matter upon its referral by the
presiding officer. Depending on the resolution of those matters, the
Commission, the Director of Nuclear Reactor Regulation, or the Director
of New Reactors, as appropriate, after making the requisite findings,
will issue, deny or appropriately condition the license.
(b) Initial decision--combined license under 10 CFR part 52. In any
initial decision in a contested proceeding on an application for a
combined license (including an amendment to or renewal of a combined
license) under subpart C of part 52 of this chapter, the presiding
officer shall make findings of fact and conclusions of law on the
matters put into controversy by the parties to the proceeding, and any
matter designated by the Commission to be decided by the presiding
officer. Depending on the resolution of those matters, the Commission,
the Director of New Reactors, or the Director of Nuclear Reactor
Regulation, as appropriate, after making the requisite findings, will
issue, deny or appropriately condition the license.
(c) Initial decision on finding under 10 CFR 52.103 with respect to
acceptance criteria in nuclear power reactor combined licenses. In any
initial decision under Sec. 52.103(g) of this chapter with respect to
whether acceptance criteria have been or will be met, the presiding
officer shall make findings of fact and conclusions of law
[[Page 49476]]
on the matters put into controversy by the parties to the proceeding,
and on any matters designated by the Commission to be decided by the
presiding officer. Matters not put into controversy by the parties
shall be referred to the Commission for its determination. The
Commission may, in its discretion, treat the matter as a request for
action under 10 CFR 2.206 and process the matter in accordance with
Sec. 52.103(f). Depending on the resolution of those matters, the
Commission, the Director of New Reactors, or the Director of Nuclear
Reactor Regulation, as appropriate, will make the finding under 10 CFR
52.103, or appropriately condition that finding.
(d) Initial decision--manufacturing license under 10 CFR part 52.
In any initial decision in a contested proceeding on an application for
a manufactured license (including an amendment to or renewal of a
combined license) under subpart C of part 52 of this chapter, the
presiding officer shall make findings of fact and conclusions of law on
the matters put into controversy by the parties to the proceeding, and
any matter designated by the Commission to be decided by the presiding
officer. Depending on the resolution of those matters, the Commission,
the Director of New Reactors, or the Director of Nuclear Reactor
Regulation, as appropriate, after making the requisite findings, will
issue, deny, or appropriately condition the manufacturing license.
(e) Initial decision--other proceedings not involving production or
utilization facilities. In proceedings not involving production or
utilization facilities, the presiding officer shall make findings of
fact and conclusions of law on the matters put into controversy by the
parties to the proceeding, and on any matters designated by the
Commission to be decided by the presiding officer. Matters not put into
controversy by the parties must be referred to the Director of Nuclear
Material Safety and Safeguards, or the Director of the Office of
Federal and State Materials and Environmental Management Programs, as
appropriate. Depending on the resolution of those matters, the Director
of Nuclear Material Safety and Safeguards or the Director of the Office
of Federal and State Materials and Environmental Management Programs,
as appropriate, after making the requisite findings, will issue, deny,
revoke or appropriately condition the license, or take other action as
necessary or appropriate.
(f) Immediate effectiveness of certain decisions. An initial
decision directing the issuance or amendment of a limited work
authorization under 10 CFR 50.10, an early site permit under subpart A
of part 52 of this chapter, a construction permit or construction
authorization under part 50 of this chapter, an operating license under
part 50 of this chapter, a combined license under subpart C of part 52
of this chapter, a manufacturing license under subpart F of part 52 of
this chapter, or a license under 10 CFR part 72 to store spent fuel in
an independent spent fuel storage facility (ISFSI) or a monitored
retrievable storage installation (MRS), an initial decision directing
issuance of a license under part 61 of this chapter, or an initial
decision under 10 CFR 52.103(g) that acceptance criteria in a combined
license have been met, is immediately effective upon issuance unless
the presiding officer finds that good cause has been shown by a party
why the initial decision should not become immediately effective.
(g)-(h) [Reserved]
(i) Issuance of authorizations, permits, and licenses--production
and utilization facilities. The Commission, the Director of New
Reactors, or the Director of Nuclear Reactor Regulation, as
appropriate, shall issue a limited work authorization under 10 CFR
50.10, an early site permit under subpart A of part 52 of this chapter,
a construction permit or construction authorization under part 50 of
this chapter, an operating license under part 50 of this chapter, a
combined license under subpart C of part 52 of this chapter, or a
manufacturing license under subpart F of part 52 of this chapter within
10 days from the date of issuance of the initial decision:
(1) If the Commission or the appropriate Director has made all
findings necessary for issuance of the authorization, permit or
license, not within the scope of the initial decision of the presiding
officer; and
(2) Notwithstanding the pendency of a petition for reconsideration
under Sec. 2.345, a petition for review under Sec. 2.341, or a motion
for stay under Sec. 2.342, or the filing of a petition under Sec.
2.206.
(j) Issuance of finding on acceptance criteria under 10 CFR 52.103.
The Commission, the Director of New Reactors, or the Director of
Nuclear Reactor Regulation, as appropriate, shall make the finding
under 10 CFR 52.103(g) that acceptance criteria in a combined license
have been, or will be met, within 10 days from the date of issuance of
the initial decision:
(1) If the Commission or the appropriate Director has made the
finding under Sec. 52.103(g) that acceptance criteria have been, or
will be met, for those acceptance criteria which are not within the
scope of the initial decision of the presiding officer; and
(2) Notwithstanding the pendency of a petition for reconsideration
under Sec. 2.345, a petition for review under Sec. 2.341, or a motion
for stay under Sec. 2.342, or the filing of a petition under Sec.
2.206.
(k) Issuance of other licenses. The Commission or the Director of
Nuclear Material Safety and Safeguards, or the Director of the Office
of Federal and State Materials and Environmental Management Programs,
as appropriate, shall issue a license, including a license under 10 CFR
part 72 to store spent fuel in either an independent spent fuel storage
facility (ISFSI) located away from a reactor site or at a monitored
retrievable storage installation (MRS), within 10 days from the date of
issuance of the initial decision:
(1) If the Commission or the appropriate Director has made all
findings necessary for issuance of the license, not within the scope of
the initial decision of the presiding officer; and
(2) Notwithstanding the pendency of a petition for reconsideration
under Sec. 2.345, a petition for review under Sec. 2.341, or a motion
for stay under Sec. 2.342, or the filing of a petition under Sec.
2.206.
0
21. In Sec. 2.341, paragraph (a)(1) is revised to read as follows:
Sec. 2.341 Review of decisions and actions of a presiding officer.
(a)(1) Except for requests for review or appeals under Sec. 2.311
or in a proceeding on the high-level radioactive waste repository
(which are governed by Sec. 2.1015), review of decisions and actions
of a presiding officer are treated under this section, provided,
however, that no party may request a further Commission review of a
Commission determination to allow a period of interim operation under
10 CFR 52.103(c).
* * * * *
0
22. In Sec. 2.347, paragraph (a) is revised, and new paragraph (f)(5)
is added to read as follows:
Sec. 2.347 Ex parte communications.
* * * * *
(a)(1) Interested persons outside the agency may not make or
knowingly cause to be made to any Commission adjudicatory employee, any
ex parte communication relevant to the merits of the proceeding.
(2) For purposes of this section, merits of the proceeding
includes:
(i) A disputed issue;
[[Page 49477]]
(ii) A matter which a presiding officer seeks to be referred to the
Commission under 10 CFR 2.340(a); and
(iii) A matter for which the Commission has approved examination by
the presiding officer under Sec. 2.340(a).
* * * * *
(f) * * *
(5) Communications, in contested proceedings and uncontested
mandatory proceeding, regarding an undisputed issue.
0
23. In Sec. 2.348, the introductory text of paragraph (a) is revised,
and new paragraphs (d)(1)(iii), (d)(1)(iv), and (d)(3) are added to
read as follows:
Sec. 2.348 Separation of functions.
(a) In any proceeding under this part, any NRC officer or employee
engaged in the performance of any investigative or litigating function
in the proceeding or in a factually related proceeding with respect to
a disputed issue in that proceeding, may not participate in or advise a
Commission adjudicatory employee about the initial or final decision
with respect to that disputed issue, except--
* * * * *
(d) * * *
(1) * * *
(iii) A matter which a presiding officer seeks to be referred to
the Commission under 10 CFR 2.340(a); and
(iv) A matter for which the Commission has approved examination by
the presiding officer under Sec. 2.340(a).
* * * * *
(3) Separation of functions does not apply to uncontested
proceedings, or to an undisputed issue in contested initial licensing
proceedings.
* * * * *
0
24. In Sec. 2.390, the introductory text of paragraph (a) is revised
to read as follows:
Sec. 2.390 Public inspections, exemptions, requests for withholding.
(a) Subject to the provisions of paragraphs (b), (d), (e), and (f)
of this section, final NRC records and documents, including but not
limited to correspondence to and from the NRC regarding the issuance,
denial, amendment, transfer, renewal, modification, suspension,
revocation, or violation of a license, permit, order, or standard
design approval, or regarding a rulemaking proceeding subject to this
part shall not, in the absence of an NRC determination of a compelling
reason for nondisclosure after a balancing of the interests of the
person or agency urging nondisclosure and the public interest in
disclosure, be exempt from disclosure and will be made available for
inspection and copying at the NRC Web site, http://www.nrc.gov, and/or
at the NRC Public Document Room, except for matters that are:
* * * * *
0
25. Subpart D is revised to read as follows:
Subpart D--Additional Procedures Applicable to Proceedings for the
Issuance of Licenses To Construct and/or Operate Nuclear Power
Plants of Identical Design at Multiple Sites
Sec.
2.400 Scope of subpart.
2.401 Notice of hearing on construction permit or combined license
applications pursuant to appendix N of 10 CFR parts 50 or 52.
2.402 Separate hearings on separate issues; consolidation of
proceedings.
2.403 Notice of proposed action on applications for operating
licenses pursuant to appendix N of 10 CFR part 50.
2.404 Hearings on applications for operating licenses pursuant to
appendix N of 10 CFR part 50.
2.405 Initial decisions in consolidated hearings.
2.406 Finality of decisions on separate issues.
2.407 Applicability of other sections.
Sec. 2.400 Scope of subpart.
This subpart describes procedures applicable to licensing
proceedings which involve the consideration in hearings of a number of
applications, filed by one or more applicants pursuant to appendix N of
parts 50 or 52 of this chapter, for licenses to construct and/or
operate nuclear power reactors of identical design to be located at
multiple sites.
Sec. 2.401 Notice of hearing on construction permit or combined
license applications pursuant to appendix N of 10 CFR parts 50 or 52.
(a) In the case of applications pursuant to appendix N of part 50
of this chapter for construction permits for nuclear power reactors of
the type described in Sec. 50.22 of this chapter, or applications
pursuant to appendix N of part 52 of this chapter for combined
licenses, the Secretary will issue notices of hearing pursuant to Sec.
2.104.
(b) The notice of hearing will also state the time and place of the
hearings on any separate phase of the proceeding.
Sec. 2.402 Separate hearings on separate issues; consolidation of
proceedings.
(a) In the case of applications under appendix N of part 50 of this
chapter for construction permits for nuclear power reactors of a type
described in 10 CFR 50.22, or applications pursuant to appendix N of
part 52 of this chapter for combined licenses, the Commission or the
presiding officer may order separate hearings on particular phases of
the proceeding, such as matters related to the acceptability of the
design of the reactor, in the context of the site parameters postulated
for the design or environmental matters.
(b) If a separate hearing is held on a particular phase of the
proceeding, the Commission or presiding officers of each affected
proceeding may, under 10 CFR 2.317, consolidate for hearing on that
phase two or more proceedings to consider common issues relating to the
applications involved in the proceedings, if it finds that this action
will be conducive to the proper dispatch of its business and to the
ends of justice. In specifying the place of this consolidated hearing,
due regard will be given to the convenience and necessity of the
parties, petitioners for leave to intervene, or the attorneys or
representatives of such persons, and the public interest.
Sec. 2.403 Notice of proposed action on applications for operating
licenses pursuant to appendix N of 10 CFR part 50.
In the case of applications pursuant to appendix N of part 50 of
this chapter for operating licenses for nuclear power reactors, if the
Commission has not found that a hearing is in the public interest, the
Commission, the Director of New Reactors, or the Director of Nuclear
Reactor Regulation will, prior to acting thereon, cause to be published
in the Federal Register, pursuant to Sec. 2.105, a notice of proposed
action with respect to each application as soon as practicable after
the applications have been docketed.
Sec. 2.404 Hearings on applications for operating licenses pursuant
to appendix N of 10 CFR part 50.
If a request for a hearing and/or petition for leave to intervene
is filed within the time prescribed in the notice of proposed action on
an application for an operating license pursuant to appendix N of part
50 of this chapter with respect to a specific reactor(s) at a specific
site, and the Commission, the Chief Administrative Judge, or a
presiding officer has issued a notice of hearing or other appropriate
order, then the Commission, the Chief Administrative Judge, or the
presiding officer may order separate hearings on particular phases of
the proceeding and/or consolidate for hearing two or more proceedings
in the manner described in Sec. 2.402.
[[Page 49478]]
Sec. 2.405 Initial decisions in consolidated hearings.
At the conclusion of a hearing held under this subpart, the
presiding officer will render a partial initial decision on the common
design. The partial initial decision on the common design may be
appealed under Sec. 2.341. If the proceedings have also been
consolidated with respect to matters other than the common design under
Sec. 2.317(b), the presiding officer may issue a consolidated partial
initial decision for those proceedings. No construction permit, full-
power operating license, or combined license under part 52 of this
chapter will be issued until an initial decision has been issued on all
phases of the hearing and all issues under the Act and the National
Environmental Policy Act of 1969 appropriate to the proceeding have
been resolved.
Sec. 2.406 Finality of decisions on separate issues.
Notwithstanding any other provision of this chapter, in a
proceeding conducted pursuant to this subpart and appendices N of parts
50 or 52 of this chapter, no matter which has been reserved for
consideration in one phase of the hearing shall be considered at
another phase of the hearing except on the basis of significant new
information that substantially affects the conclusion(s) reached at the
other phase or other good cause.
Sec. 2.407 Applicability of other sections.
The provisions of subparts A, C, G, L, and N of this part relating
to construction permits, operating licenses, and combined licenses
apply, respectively, to construction permits, operating licenses, and
combined licenses subject to this subpart, except as may be qualified
by the provisions of this subpart.
0
26. Section 2.500 is revised to read as follows:
Sec. 2.500 Scope of subpart.
This subpart prescribes procedures applicable to licensing
proceedings which involve the consideration in separate hearings of an
application for a license to manufacture nuclear power reactors under
subpart F of part 52 of this chapter.
0
27. In Sec. 2.501, the section heading, the introductory text of
paragraph (a) and paragraph (b) are revised to read as follows:
Sec. 2.501 Notice of hearing on application under subpart F of 10 CFR
part 52 for a license to manufacture nuclear power reactors.
(a) In the case of an application under subpart F of part 52 of
this chapter for a license to manufacture nuclear power reactors of the
type described in Sec. 50.22 of this chapter to be operated at sites
not identified in the license application, the Secretary will issue a
notice of hearing to be published in the Federal Register at least 30
days before the date set for hearing in the notice.\1\ The notice shall
be issued as soon as practicable after the application has been
docketed. The notice will state:
---------------------------------------------------------------------------
\1\ The thirty-day (30) requirement of this paragraph is not
applicable to a notice of the time and place of hearing published by
the presiding officer after the notice of hearing described in this
section has been published.
---------------------------------------------------------------------------
* * * * *
(b) The notice of hearing shall comply with the requirements of
Sec. 2.104(f) of this chapter.
* * * * *
Sec. 2.502 [Removed]
0
28. Remove and reserve Sec. 2.502.
Sec. 2.503 [Removed]
0
29. Remove and reserve Sec. 2.503.
Sec. 2.504 [Removed]
0
30. Remove and reserve Sec. 2.504.
0
31. Subpart F is revised to read as follows:
Subpart F--Additional Procedures Applicable to Early Partial
Decisions on Site Suitability Issues in Connection With an
Application for a Construction Permit or Combined License for
Certain Utilization Facilities
Sec.
2.600 Scope of subpart.
2.601 Applicability of other sections.
Early Partial Decisions on Site Suitability--Construction Permit
2.602 Filing Fees.
2.603 Acceptance and docketing of application for early review of
site suitability issues in a construction permit proceeding.
2.604 Notice of hearing on application for early review of site
suitability issues in construction permit proceeding.
2.605 Additional considerations.
2.606 Partial decision on site suitability issues in construction
permit proceeding.
Early Partial Decisions on Site Suitability--Combined License Under 10
CFR Part 52
2.621 Acceptance and docketing of application for early review of
site suitability issues in a combined license proceeding.
2.623 Notice of hearing on application for early review of site
suitability issues in combined license proceeding.
2.625 Additional considerations.
2.627 Partial decision on site suitability issues in combined
license proceeding.
2.629 Finality of partial decision on site suitability issues in
combined license proceeding.
Sec. 2.600 Scope of subpart.
This subpart prescribes procedures applicable to licensing
proceedings which involve an early submittal of site suitability
information in accordance with Sec. 2.101(a-1) and (a-2), and a
hearing and early partial decision on issues of site suitability, in
connection with an application for a permit to construct a utilization
facility which is subject to Sec. 51.20(b) of this chapter and is of
the type specified in Sec. 50.21(b)(2) or (3) or Sec. 50.22 of this
chapter or is a testing facility; or an application for a combined
license under part 52 of this chapter for a nuclear power facility.
(a) The procedures in Sec. Sec. 2.601 through 2.609 apply to all
applications under this subpart.
(b) The procedures in Sec. Sec. 2.611 through 2.619 apply to
applications for a permit to construct a utilization facility which is
subject to Sec. 51.20(b) of this chapter and is of the type specified
in Sec. 50.21(b)(2) or (3) or Sec. 50.22 of this chapter or is a
testing facility.
(c) The procedures in Sec. Sec. 2.621 through 2.629 apply to
applications for combined license under part 52 of this chapter for a
nuclear power facility.
Sec. 2.601 Applicability of other sections.
The provisions of subparts A, C, G, L, and N relating to
applications for construction permits and combined licenses, and
proceedings thereon apply, respectively, to such applications and
proceedings in accordance with this subpart, except as specifically
provided otherwise by the provisions of this subpart.
Early Partial Decisions on Site Suitability--Construction Permit
Sec. 2.602 Filing fees.
Each application which contains a request for early review of site
suitability issues under the procedures of this subpart shall be
accompanied by any fee required by Sec. 50.30(e) and part 170 of this
chapter.
Sec. 2.603 Acceptance and docketing of application for early review
of site suitability issues in a construction permit proceeding.
(a) Each part of an application for a construction permit submitted
in accordance with Sec. 2.101(a-1) of this part will be initially
treated as a tendered
[[Page 49479]]
application. If it is determined that any one of the parts as described
in Sec. 2.101(a-1) is incomplete and not acceptable for processing,
the Director of the Office of New Reactors or the Director of the
Office of Nuclear Reactor Regulation, as appropriate, will inform the
applicant of this determination and the respects in which the document
is deficient. Such a determination of completeness will generally be
made within a period of 30 days.
(b)(1) The Director of the Office of New Reactors or the Director
of the Office of Nuclear Reactor Regulation, as appropriate, will
accept for docketing part one of an application for a construction
permit for a utilization facility which is subject to Sec. 51.20(b) of
this chapter and is of the type specified in Sec. 50.21(b)(2) or (3)
or Sec. 50.22 of this chapter, or is a testing facility where part one
of the application as described in Sec. 2.101(a-1) is complete. Part
one of any application will not be considered complete unless it
contains proposed findings as required by Sec. 2.101(a-1)(1)(i) and
unless it describes the applicant's site selection process, specifies
the extent to which that process involves the consideration of
alternative sites, explains the relationship between that process and
the application for early review of site suitability issues, and
briefly describes the applicant's long-range plans for ultimate
development of the site. Upon assignment of a docket number, the
procedures in Sec. 2.101(a)(3) and (4) relating to formal docketing
and the submission and distribution of additional copies of the
application shall be followed.
(2) Additional parts of the application will be docketed upon a
determination by the Director of the Office of New Reactors or the
Director of the Office of Nuclear Reactor Regulation, as appropriate,
that they are complete.
(c) If part one of the application is docketed, the Director of the
Office of New Reactors or the Director of the Office of Nuclear Reactor
Regulation, as appropriate, will cause to be published in the Federal
Register and send to the Governor or other appropriate official of the
State in which the site is located, a notice of docketing of the
application which states the purpose of the application, states the
location of the proposed site, states that a notice of hearing will be
published, requests comments within 120 days or such other time as may
be specified on the initiation or outcome of an early site review from
Federal, State, and local agencies and interested persons.
Sec. 2.604 Notice of hearing on application for early review of site
suitability issues in construction permit proceeding.
(a) Where an applicant for a construction permit requests an early
review and hearing and an early partial decision on issues of site
suitability pursuant to Sec. 2.101(a-1), the provisions in the notice
of hearing setting forth the matters of fact and law to be considered,
as required by Sec. 2.104, shall be modified so as to relate only to
the site suitability issue or issues under review.
(b) After docketing of part two of the application, as provided in
Sec. Sec. 2.101(a-1) and 2.603, a supplementary notice of hearing will
be published under Sec. 2.104 with respect to the remaining unresolved
issues in the proceeding within the scope of Sec. 2.104. This
supplementary notice of hearing will provide that any person whose
interest may be affected by the proceeding and who desires to
participate as a party in the resolution of the remaining issues shall
file a petition for leave to intervene pursuant to Sec. 2.309 within
the time prescribed in the notice. This supplementary notice will also
provide appropriate opportunities for participation by a representative
of an interested State under Sec. 2.315(c) and for limited appearances
under Sec. 2.315(a).
(c) Any person who was permitted to intervene as a party under the
initial notice of hearing on site suitability issues and who was not
dismissed or did not withdraw as a party may continue to participate as
a party to the proceeding with respect to the remaining unresolved
issues, provided that within the time prescribed for filing of
petitions for leave to intervene in the supplementary notice of
hearing, he or she files a notice of his intent to continue as a party,
along with a supporting affidavit identifying the specific aspect or
aspects of the subject matter of the proceeding as to which he or she
wishes to continue to participate as a party and setting forth with
particularity the basis for his contentions with regard to each aspect
or aspects. A party who files a non-timely notice of intent to continue
as a party may be dismissed from the proceeding, absent a determination
that the party has made a substantial showing of good cause for failure
to file on time, and with particular reference to the factors specified
in Sec. Sec. 2.309(c)(1)(i) through (iv) and 2.309(d). The notice will
be ruled upon by the Commission or presiding officer designated to rule
on petitions for leave to intervene.
(d) To the maximum extent practicable, the membership of any atomic
safety and licensing board designated to preside in the proceeding on
the remaining unresolved issues pursuant to the supplemental notice of
hearing will be the same as the membership designated to preside in the
initial notice of hearing on site suitability issues.
Sec. 2.605 Additional considerations.
(a) The Commission will not conduct more than one review of site
suitability issues with regard to a particular site prior to filing and
review of part two of the application described in Sec. 2.101(a-1) of
this part.
(b) The Commission, upon its own initiative, or upon the motion of
any party to the proceeding filed at least 60 days prior to the date of
the commencement of the evidentiary hearing on site suitability issues,
may decline to initiate an early hearing or render an early partial
decision on any issue or issues of site suitability:
(1) In cases where no partial decision on the relative merits of
the proposed site and alternative sites under subpart A of part 51 of
this chapter is requested, upon determination that there is a
reasonable likelihood that further review would identify one or more
preferable alternative sites and the partial decision on one or more
site suitability issues would lead to an irreversible and irretrievable
commitment of resources prior to the submittal of the remainder of the
information required by Sec. 50.30(f) of this chapter that would
prejudice the later review and decision on such alternative sites; or
(2) In cases where it appears that an early partial decision on any
issue or issues of site suitability would not be in the public interest
considering:
(i) The degree of likelihood that any early findings on those
issues would retain their validity in later reviews;
(ii) The objections, if any, of cognizant State or local government
agencies to the conduct of an early review on those issues; and
(iii) The possible effect on the public interest and the parties of
having an early, if not necessarily conclusive, resolution of those
issues.
Sec. 2.606 Partial decision on site suitability issues in
construction permit proceeding.
(a) The provisions of Sec. Sec. 2.331, 2.339, 2.340, 2.343, 2.712,
and 2.713 shall apply to any partial initial decision rendered in
accordance with this subpart. A limited work authorization may not be
issued under 10 CFR 50.10(e) and no construction permit may be issued
without completion of the full review required by Section 102(2) of the
National Environmental Policy Act of 1969, as amended, and
[[Page 49480]]
subpart A of part 51 of this chapter. The authority of the Commission
to review such a partial initial decision sua sponte, or to raise sua
sponte an issue that has not been raised by the parties, will be
exercised within the same time period as in the case of a full decision
relating to the issuance of a construction permit.
(b)(1) A partial decision on one or more site suitability issues
pursuant to the applicable provisions of part 50, subpart A of part 51,
and part 100 of this chapter issued in accordance with this subpart
shall:
(i) Clearly identify the site to which the partial decision
applies; and
(ii) Indicate to what extent additional information may be needed
and additional review may be required to enable the Commission to
determine in accordance with the provisions of the Act and the
applicable provisions of the regulations in this chapter whether a
construction permit for a facility to be located on the site identified
in the partial decision should be issued or denied.
(2) Following either the Commission (acting in the function of a
presiding officer) issuance of a partial initial decision, or
completion of Commission review of the partial initial decision of the
Atomic Safety and Licensing Board, after hearing, on the site
suitability issues, the partial decision shall remain in effect either
for a period of 5 years or, where the applicant for the construction
permit has made timely submittal of the information required to support
the application as provided in Sec. 2.101(a-1), until the proceeding
for a permit to construct a facility on the site identified in the
partial decision has been concluded,\3\ unless the Commission or Atomic
Safety and Licensing Board, upon its own initiative or upon motion by a
party to the proceeding, finds that there exists significant new
information that substantially affects the earlier conclusions and
reopens the hearing record on site suitability issues. Upon good cause
shown, the Commission may extend the 5-year period during which a
partial decision shall remain in effect for a reasonable period of time
not to exceed 1 year.
---------------------------------------------------------------------------
\3\ The partial decision on site suitability issues shall be
incorporated in the decision regarding issuance of the combined
license to the extent that it serves as a basis for the decision on
a specific site issue.
---------------------------------------------------------------------------
Early Partial Decisions on Site Suitability--Combined License Under 10
CFR Part 52
Sec. 2.621 Acceptance and docketing of application for early review
of site suitability issues in a combined license proceeding.
(a) Each part of an application submitted in accordance with Sec.
2.101(a-1) of this part will be initially treated as a tendered
application. If it is determined that any one of the parts as described
in Sec. 2.101(a-1) is incomplete and not acceptable for processing,
the Director of the Office of New Reactors or the Director of the
Office of Nuclear Reactor Regulation, as appropriate, will inform the
applicant of this determination and the respects in which the document
is deficient. Such a determination of completeness will generally be
made within a period of 30 days.
(b)(1) The Director of the Office of New Reactors or the Director
of the Office of Nuclear Reactor Regulation, as appropriate, will
accept for docketing an application for a combined license for a
nuclear power facility where part one of the application as described
in Sec. 2.101(a-1) is complete. Part one of any application will not
be considered complete unless it contains proposed findings as required
by Sec. 2.101(a-1)(1)(i) and unless it describes the applicant's site
selection process, specifies the extent to which that process involves
the consideration of alternative sites, explains the relationship
between that process and the application for early review of site
suitability issues, and briefly describes the applicant's long-range
plans for ultimate development of the site. Upon assignment of a docket
number, the procedures in Sec. 2.101(a)(3) and (4) relating to formal
docketing and the submission and distribution of additional copies of
the application shall be followed.
(2) Additional parts of the application will be docketed upon a
determination by the Director of the Office of New Reactors or the
Director of the Office of Nuclear Reactor Regulation, as appropriate,
that they are complete.
(c) If part one of the application is docketed, the Director of the
Office of New Reactors or the Director of the Office of Nuclear Reactor
Regulation, as appropriate, will cause to be published in the Federal
Register and send to the Governor or other appropriate official of the
State in which the site is located, a notice of docketing of the
application which states the purpose of the application, states the
location of the proposed site, states that a notice of hearing will be
published, requests comments within 120 days or such other time as may
be specified on the initiation or outcome of an early site review from
Federal, State, and local agencies and interested persons.
Sec. 2.623 Notice of hearing on application for early review of site
suitability issues in combined license proceeding.
(a) Where an applicant for a combined license under part 52 of this
chapter requests an early review and hearing and an early partial
decision on issues of site suitability pursuant to Sec. 2.101(a-2),
the provisions in the notice of hearing setting forth the matters of
fact and law to be considered, as required by Sec. 2.104, shall be
modified so as to relate only to the site suitability issue or issues
under review. The notice will provide appropriate opportunities for
participation by a representative of an interested State under Sec.
2.315(c) and for limited appearances under Sec. 2.315(a), limited
however, to the issues of site suitability for which early review has
been requested by the applicant.
(b) After docketing of part two of the application, as provided in
Sec. Sec. 2.101(a-1) and 2.603, a supplementary notice of hearing will
be published under Sec. 2.104 with respect to the remaining unresolved
issues in the proceeding within the scope of Sec. 2.104. This
supplementary notice of hearing will provide that any person whose
interest may be affected by the proceeding and who desires to
participate as a party in the resolution of the remaining issues shall
file a petition for leave to intervene pursuant to Sec. 2.309 within
the time prescribed in the notice. This supplementary notice will also
provide appropriate opportunities for participation by a representative
of an interested State under Sec. 2.315(c) and for limited appearances
under Sec. 2.315(a).
(c) Any person who was permitted to intervene as a party under the
initial notice of hearing on site suitability issues and who was not
dismissed or did not withdraw as a party may continue to participate as
a party to the proceeding without having to demonstrate standing under
Sec. 2.309(d), provided, however, that within the time prescribed for
filing of petitions for leave to intervene in the supplementary notice
of hearing, the party files a notice of intent to continue as a party.
The notice must include the information required by Sec. 2.309(f). A
party who files a non-timely notice of intent to continue as a party
may be dismissed from the proceeding, absent a determination that the
party has made a substantial showing of good cause for failure to file
on time, and with particular reference to the factors specified in
Sec. Sec. 2.309(c)(1)(i)
[[Page 49481]]
through (iv) and 2.309(d). The notice will be ruled upon by the
Commission or presiding officer designated to rule on petitions for
leave to intervene.
(d) To the maximum extent practicable, the presiding officer (as
applicable, the membership of the licensing board) designated to
preside in the proceeding on the remaining unresolved issues pursuant
to the supplemental notice of hearing will be the same as the presiding
officer (as applicable, the membership of the licensing board)
designated to preside in the initial notice of hearing on site
suitability issues.
Sec. 2.625 Additional considerations.
(a) The Commission will not conduct more than one review of site
suitability issues with regard to a particular site prior to filing and
review of part two of the application described in Sec. 2.101(a-1) of
this part.
(b) The Commission, upon its own initiative, or upon the motion of
any party to the proceeding filed at least 60 days prior to the date of
the commencement of the evidentiary hearing on site suitability issues,
may decline to initiate an early hearing or render an early partial
decision on any issue or issues of site suitability:
(1) In cases where no partial decision on the relative merits of
the proposed site and alternative sites under subpart A of part 51 is
requested, upon determination that there is a reasonable likelihood
that further review would identify one or more preferable alternative
sites and the partial decision on one or more site suitability issues
would lead to an irreversible and irretrievable commitment of resources
prior to the submittal of the remainder of the information required by
Sec. 50.30(f) of this chapter that would prejudice the later review
and decision on such alternative sites; or
(2) In cases where it appears that an early partial decision on any
issue or issues of site suitability would not be in the public interest
considering:
(i) The degree of likelihood that any early findings on those
issues would retain their validity in later reviews;
(ii) The objections, if any, of cognizant State or local government
agencies to the conduct of an early review on those issues; and
(iii) The possible effect on the public interest and the parties of
having an early, if not necessarily conclusive, resolution of those
issues.
Sec. 2.627 Partial decision on site suitability issues in combined
license proceeding.
(a) The provisions of Sec. Sec. 2.331, 2.339, 2.340(b), 2.343,
2.712, and 2.713 shall apply to any partial initial decision rendered
in accordance with this subpart. Section 2.340(c) shall not apply to
any partial initial decision rendered in accordance with this subpart.
A limited work authorization may not be issued under 10 CFR 50.10(e)
and no construction permit may be issued without completion of the full
review required by Section 102(2) of the National Environmental Policy
Act of 1969, as amended, and subpart A of part 51 of this chapter. The
authority of the Commission to review such a partial initial decision
sua sponte, or to raise sua sponte an issue that has not been raised by
the parties, will be exercised within the same time period as in the
case of a full decision relating to the issuance of a construction
permit.
(b)(1) A partial decision on one or more site suitability issues
pursuant to the applicable provisions of part 50, subpart A of part 51,
and part 100 of this chapter issued in accordance with this subpart
shall:
(i) Clearly identify the site to which the partial decision
applies; and
(ii) Indicate to what extent additional information may be needed
and additional review may be required to enable the Commission to
determine in accordance with the provisions of the Act and the
applicable provisions of the regulations in this chapter whether a
construction permit for a facility to be located on the site identified
in the partial decision should be issued or denied.
(2) Following either the Commission (acting in the function of a
presiding officer) issuance of a partial initial decision, or
completion of Commission review of the partial initial decision of the
presiding officer, after hearing, on the site suitability issues, the
partial decision shall remain in effect either for a period of 5 years
or, where the applicant for the combined license has made timely
submittal of the information required to support the application as
provided in Sec. 2.101(a-2), until the proceeding for a combined
license on the site identified in the partial decision has been
concluded, unless the Commission or presiding officer, upon its own
initiative or upon motion by a party to the proceeding, finds that
there exists significant new information that substantially affects the
earlier conclusions and reopens the hearing record on site suitability
issues. Upon good cause shown, the Commission may extend the 5-year
period during which a partial decision shall remain in effect for a
reasonable period of time not to exceed 1 year.
Sec. 2.629 Finality of partial decision on site suitability issues in
a combined license proceeding.
(a) The partial decision on site suitability issues in a combined
license proceeding shall be incorporated in the decision regarding
issuance of a combined license. Except as provided in 10 CFR 2.758, in
making the findings required for issuance of a combined license, the
Commission shall treat as resolved those matters resolved in connection
with the issuance of the partial decision on site suitability issues.
If the Commission reaches an adverse decision, the application shall be
denied without prejudice for resubmission, provided, however, that in
determining whether the resubmitted application is complete and
acceptable for docketing under Sec. 2.101(a)(3), the Director of the
Office of New Reactors or the Director of the Office of Nuclear Reactor
Regulation, as appropriate, shall determine whether the resubmitted
application addresses those matters identified as bases for denial of
the original application.
(b) Notwithstanding any provision in 10 CFR 50.109, while a partial
decision on site suitability is in effect under Sec. 2.617(b)(2), the
Commission may not modify, rescind, or impose new requirements with
respect to matters within the scope of the site suitability decision,
whether on its own motion, or in response to a request or petition from
any person, unless the Commission determines that a modification to the
original decision is necessary either for compliance with the
Commission's regulations applicable and in effect at the time the
partial decision was issued, or to assure adequate protection of the
public health and safety or the common defense and security.
0
32. Section 2.800 is revised to read as follows:
Sec. 2.800 Scope and applicability.
(a) This subpart governs the issuance, amendment, and repeal of
regulations in which participation by interested persons is prescribed
under Section 553 of title 5 of the U.S. Code.
(b) The procedures in Sec. Sec. 2.804 through 2.810 apply to all
rulemakings.
(c) The procedures in Sec. Sec. 2.802 through 2.803 apply to all
petitions for rulemaking except for initial applications for standard
design certification rulemaking under subpart B of part 52 of this
chapter, and subsequent petitions for amendment of an existing design
certification rule filed by the original applicant for the design
certification rule.
(d) The procedures in Sec. Sec. 2.811 through 2.819, as
supplemented by the
[[Page 49482]]
provisions of subpart B of part 52, apply to standard design
certification rulemaking.
0
33. Section 2.801 is revised to read as follows:
Sec. 2.801 Initiation of rulemaking.
Rulemaking may be initiated by the Commission at its own instance,
on the recommendation of another agency of the United States, or on the
petition of any other interested person, including an application for
design certification under subpart B of part 52 of this chapter.
0
34. In subpart H, Sec. Sec. 2.811, 2.813, 2.815, 2.817 and 2.819 are
added to read as follows:
Sec. 2.811 Filing of standard design certification application;
required copies.
(a) Serving of applications. The signed original of an application
for a standard design certification, including all amendments to the
applications, must be sent either by mail addressed: ATTN: Document
Control Desk, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001; by facsimile; by hand delivery to the NRC's offices at 11555
Rockville Pike, Rockville, Maryland, between the hours of 7:30 a.m. and
4:15 p.m. eastern time; or, where practicable, by electronic
submission, for example, via Electronic Information Exchange, e-mail,
or CD-ROM. Electronic submissions must be made in a manner that enables
the NRC to receive, read, authenticate, distribute, and archive the
submission, and process and retrieve it a single page at a time.
Detailed guidance on making electronic submissions can be obtained by
visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html, by calling (301) 415-0439, by e-mail at [email protected],
or by writing the Office of Information Services, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001. The guidance
discusses, among other topics, the formats the NRC can accept, the use
of electronic signatures, and the treatment of nonpublic information.
If the communication is on paper, the signed original must be sent.
(b) Form of application. Each original of an application and an
amendment of an application must meet the requirements in Sec. 2.813.
(c) Capability to provide additional copies. The applicant shall
maintain the capability to generate additional copies of the general
information and the safety analysis report, or part thereof or
amendment thereto, for subsequent distribution in accordance with the
written instructions of the Director, Office of New Reactors, the
Director, Office of Nuclear Reactor Regulation, or the Director, Office
of Nuclear Material Safety and Safeguards, as appropriate.
(d) Public hearing copy. In any hearing conducted under subpart O
of this part for a design certification rulemaking, the applicant must
make a copy of the updated application available at the public hearing
for the use of any other parties to the proceeding, and shall certify
that the updated copies of the application contain the current contents
of the application submitted in accordance with the requirements of
this part.
(e) Pre-application consultation. A prospective applicant for a
standard design certification may consult with the NRC before filing an
application by writing to the Director, Division of New Reactor
Licensing, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, with respect to the subject matters listed in Sec.
2.802(a)(1)(i) through (iii) of this chapter. A prospective petitioner
also may telephone the Rulemaking, Directives, and Editing Branch on
(301) 415-7163, or toll free on (800) 368-5642, or send e-mail to
[email protected] on these subject matters. In addition, a prospective
applicant may confer informally with the NRC staff BEFORE filing an
application for a standard design certification, and the limitations in
Sec. 2.802(a)(2) do not apply.
Sec. 2.813 Written communications.
(a) General requirements. All correspondence, reports, and other
written communications from the applicant to the Nuclear Regulatory
Commission concerning the regulations in this subpart, and parts 50,
52, and 100 of this chapter must be sent either by mail addressed:
ATTN: Document Control Desk, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001; by hand delivery to the NRC's offices at
11555 Rockville Pike, Rockville, Maryland, between the hours of 7:30
a.m. and 4:15 p.m. eastern time; or, where practicable, by electronic
submission, for example, via Electronic Information Exchange, e-mail,
or CD-ROM. Electronic submissions must be made in a manner that enables
the NRC to receive, read, authenticate, distribute, and archive the
submission, and process and retrieve it a single page at a time.
Detailed guidance on making electronic submissions can be obtained by
visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html, by calling (301) 415-0439, by e-mail at [email protected],
or by writing the Office of Information Services, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001. The guidance
discusses, among other topics, the formats the NRC can accept, the use
of electronic signatures, and the treatment of nonpublic information.
If the communication is on paper, the signed original must be sent. If
a submission due date falls on a Saturday, Sunday, or Federal holiday,
the next Federal working day becomes the official due date.
(b) Form of communications. All paper copies submitted to meet the
requirements set forth in paragraph (a) of this section must be
typewritten, printed or otherwise reproduced in permanent form on
unglazed paper. Exceptions to these requirements imposed on paper
submissions may be granted for the submission of micrographic,
photographic, or similar forms.
(c) Regulation governing submission. An applicant submitting
correspondence, reports, and other written communications under the
regulations of this chapter is requested but not required to cite
whenever practical, in the upper right corner of the first page of the
submission, the specific regulation or other basis requiring
submission.
Sec. 2.815 Docketing and acceptance review.
(a) Each application for a standard design certification will be
assigned a docket number. However, to allow a determination as to
whether an application is complete and acceptable for docketing, it
will be initially treated as a tendered application. A copy of the
tendered application will be available for public inspection at the NRC
Web site, http://www.nrc.gov, and/or at the NRC Public Document Room.
Generally, the determination on acceptability for docketing will be
made within a period of 30 days. The Commission may decide to determine
acceptability on the basis of the technical adequacy of the application
as well as its completeness.
(b) If the Commission determines that a tendered application is
complete and acceptable for docketing, a docket number will be assigned
to the application or part thereof, and the applicant will be notified
of the determination.
Sec. 2.817 Withdrawal of application.
(a) The Commission may permit an applicant to withdraw an
application for a standard design certification before the issuance of
a notice of proposed rulemaking on such terms and conditions as the
Commission may prescribe, or may, on receiving a request for withdrawal
of an application, deny the application or dismiss it without
prejudice. The NRC will publish in the
[[Page 49483]]
Federal Register a document withdrawing the application, if the notice
of receipt of the application, an advance notice of proposed
rulemaking, or a notice of proposed rulemaking for the standard design
certification has been previously published in the Federal Register. If
the notice of receipt, advance notice of proposed rulemaking or notice
of proposed rulemaking was published on the NRC Web site, then the
notice of action on the withdrawal will also be published on the NRC
Web site.
(b) The withdrawal of an application does not authorize the removal
of any document from the files of the Commission.
Sec. 2.819 Denial of application for failure to supply information.
(a) The Commission may deny an application for a standard design
certification if an applicant fails to respond to a request for
additional information within 30 days from the date of the request, or
within such other time as may be specified.
(b) If the Commission denies an application because the applicant
has failed to respond in a timely fashion to a request for additional
information, the NRC will publish in the Federal Register a notice of
denial and will notify the applicant with a simple statement of the
grounds of denial. If a notice of receipt of application, advance
notice of proposed rulemaking, or notice of proposed rulemaking for a
standard design certification was published on the NRC Web site, then
the notice of action on the denial will also be published on the NRC
Web site.
0
35. In Sec. 2.1202, paragraph (a) is revised to read as follows:
Sec. 2.1202 Authority and role of NRC staff.
(a) During the pendency of any hearing under this subpart,
consistent with the NRC staff's findings in its review of the
application or matter which is the subject of the hearing and as
authorized by law, the NRC staff is expected to issue its approval or
denial of the application promptly, or take other appropriate action on
the underlying regulatory matter for which a hearing was provided. When
the NRC staff takes its action, it shall notify the presiding officer
and the parties to the proceeding of its action. That notice must
include the NRC staff's position on the matters in controversy before
the presiding officer with respect to the staff action. The NRC staff's
action on the matter is effective upon issuance by the staff, except in
matters involving:
(1) An application to construct and/or operate a production or
utilization facility (including an application for a limited work
authorization under 10 CFR 50.12, or an application for a combined
license under subpart C of 10 CFR part 52);
(2) An application for an early site permit under subpart A of 10
CFR part 52;
(3) An application for a manufacturing license under subpart F of
10 CFR part 52;
(4) An application for an amendment to a construction authorization
for a high-level radioactive waste repository at a geologic repository
operations area falling under either 10 CFR 60.32(c)(1) or 10 CFR part
63;
(5) An application for the construction and operation of an
independent spent fuel storage installation (ISFSI) located at a site
other than a reactor site or a monitored retrievable storage
installation (MRS) under 10 CFR part 72; and
(6) Production or utilization facility licensing actions that
involve significant hazards considerations as defined in 10 CFR 50.92.
* * * * *
Sec. 2.1211 [Removed]
0
36. Section 2.1211 is removed.
PART 10--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR
ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN
EMPLOYMENT CLEARANCE
0
37. The authority citation for part 10 continues to read as follows:
Authority: Secs. 145, 161, 68 Stat. 942, 948, as amended (42
U.S.C. 2165, 2201); sec. 201, 88 Stat. 1242, as amended (42 U.S.C.
5841); E.O. 10450, 3 CFR parts 1949-1953 COMP., p. 936, as amended;
E.O. 10865, 3 CFR 1959-1963 COMP., p. 398, as amended; 3 CFR Table
4; E.O. 12968, 3 CFR 1995 COM., p. 396.
0
38. In Sec. 10.1, paragraphs (a)(1) and (a)(2) are revised and
paragraph (a)(3) is added to read as follows:
Sec. 10.1 Purpose.
(a) * * *
(1) The eligibility of individuals who are employed by or
applicants for employment with NRC contractors, agents, and other
individuals who are NRC employees or applicants for NRC employment, and
other persons designated by the Deputy Executive Director for
Information Services and Administration and Chief Information Officer
of the NRC, for access to Restricted Data under the Atomic Energy Act
of 1954, as amended, and the Energy Reorganization Act of 1974, or for
access to national security information;
(2) The eligibility of NRC employees, or the eligibility of
applicants for employment with the NRC, for employment clearance; and
(3) The eligibility of individuals who are employed by or are
applicants for employment with NRC licensees, certificate holders,
holders of standard design approvals under part 52 of this chapter,
applicants for licenses, certificates, and NRC approvals, and others
who may require access related to a license, certificate, or NRC
approval, or other activities as the Commission may determine, for
access to Restricted Data under the Atomic Energy Act of 1954, as
amended, and the Energy Reorganization Act of 1974, or for access to
national security information.
* * * * *
0
39. In Sec. 10.2, paragraph (b) is revised to read as follows:
Sec. 10.2 Scope.
* * * * *
(b) NRC licensees, certificate holders and holders of standard
design approvals under part 52 of this chapter, applicants for
licenses, certificates, and standard design approvals under part 52 of
this chapter, and their employees (including consultants) and
applicants for employment (including consulting);
* * * * *
PART 19--NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS; INSPECTION
AND INVESTIGATIONS
0
40. The authority citation for part 19 is revised to read as follows:
Authority: Secs. 53, 63, 81, 103, 104, 161, 186, 68 Stat. 930,
933, 935, 936, 937, 948, 955, as amended, sec. 234, 83 Stat. 444, as
amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42 U.S.C. 2073,
2093, 2111, 2133, 2134, 2201, 2236, 2282, 2297f); sec. 201, 88 Stat.
1242, as amended (42 U.S.C. 5841); Pub. L. 95-601, sec. 10, 92 Stat.
2951 (42 U.S.C. 5851); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504
note).
Section 19.32 is also issued under sec. 401, 88 Stat. 1254 (42
U.S.C. 2000d, 42 U.S.C. 5891).
0
41. Section 19.1 is revised to read as follows:
Sec. 19.1 Purpose.
The regulations in this part establish requirements for notices,
instructions, and reports by licensees and regulated entities to
individuals participating in NRC-licensed and regulated activities and
options available to these individuals in connection with Commission
inspections of licensees and regulated entities, and to ascertain
[[Page 49484]]
compliance with the provisions of the Atomic Energy Act of 1954, as
amended, titles II and IV of the Energy Reorganization Act of 1974, and
regulations, orders, and licenses thereunder. The regulations in this
part also establish the rights and responsibilities of the Commission
and individuals during interviews compelled by subpoena as part of
agency inspections or investigations under Section 161c of the Atomic
Energy Act of 1954, as amended, on any matter within the Commission's
jurisdiction.
0
42. Section 19.2 is revised to read as follows:
Sec. 19.2 Scope.
(a) The regulations in this part apply to:
(1) All persons who receive, possess, use, or transfer material
licensed by the NRC under the regulations in parts 30 through 36, 39,
40, 60, 61, 63, 70, or 72 of this chapter, including persons licensed
to operate a production or utilization facility under parts 50 or 52 of
this chapter, persons licensed to possess power reactor spent fuel in
an independent spent fuel storage installation (ISFSI) under part 72 of
this chapter, and in accordance with 10 CFR 76.60 to persons required
to obtain a certificate of compliance or an approved compliance plan
under part 76 of this chapter;
(2) All applicants for and holders of licenses (including
construction permits and early site permits) under parts 50, 52, and 54
of this chapter;
(3) All applicants for and holders of a standard design approval
under subpart E of part 52 of this chapter; and
(4) All applicants for a standard design certification under
subpart B of part 52 of this chapter, and those (former) applicants
whose designs have been certified under that subpart.
(b) The regulations in this part regarding interviews of
individuals under subpoena apply to all investigations and inspections
within the jurisdiction of the NRC other than those involving NRC
employees or NRC contractors. The regulations in this part do not apply
to subpoenas issued under 10 CFR 2.702.
0
43. In Sec. 19.3 the definitions of License and Worker are revised,
and the definitions of Regulated entities and Regulated activities are
added to read as follows:
Sec. 19.3 Definitions.
* * * * *
License means a license issued under the regulations in parts 30
through 36, 39, 40, 60, 61, 63, 70, or 72 of this chapter, including
licenses to manufacture, construct and/or operate a production or
utilization facility under parts 50, 52, or 54 of this chapter.
* * * * *
Regulated activities means any activity carried on which is under
the jurisdiction of the NRC under the Atomic Energy Act of 1954, as
amended, or any title of the Energy Reorganization Act of 1972, as
amended.
Regulated entities means any individual, person, organization, or
corporation that is subject to the regulatory jurisdiction of the NRC,
including (but not limited to) an applicant for or holder of a standard
design approval under subpart E of part 52 of this chapter or a
standard design certification under subpart B of part 52 of this
chapter.
* * * * *
Worker means an individual engaged in activities licensed or
regulated by the Commission and controlled by a licensee or regulated
entity, but does not include the licensee or regulated entity.
0
44. In Sec. 19.11, paragraph (c) is removed and reserved, and the
introductory text of paragraph (a), paragraphs (b), (d), and (e) are
revised, and paragraphs (f) and (g) are added to read as follows:
Sec. 19.11 Posting of notices to workers.
(a) Each licensee (except for a holder of an early site permit
under subpart A of part 52 of this chapter, or a holder of a
manufacturing license under subpart F of part 52 of this chapter) shall
post current copies of the following documents:
* * * * *
(b) Each applicant for and holder of a standard design approval
under subpart E of part 52 of this chapter, each applicant for an early
site permit under subpart A of part 52 of this chapter, each applicant
for a standard design certification under subpart B of part 52 of this
chapter, and each applicant for and holder of a manufacturing license
under subpart F of part 52 of this chapter shall post:
(1) The regulations in this part;
(2) The operating procedures applicable to the activities regulated
by the NRC which are being conducted by the applicant or holder; and
(3) Any notice of violation, proposed imposition of civil penalty,
or order issued under subpart B of part 2 of this chapter, and any
response from the applicant or holder.
(c) [Reserved]
(d) If posting of a document specified in paragraphs (a)(1), (2) or
(3), or (b)(1) or (2) of this section is not practicable, the licensee
or regulated entity may post a notice which describes the document and
states where it may be examined.
(e)(1) Each licensee, each applicant for a specific license, each
applicant for or holder of a standard design approval under subpart E
of part 52 of this chapter, each applicant for an early site permit
under subpart A of part 52 of this chapter, and each applicant for a
standard design certification under subpart B of part 52 of this
chapter shall prominently post NRC Form 3, ``Notice to Employees,''
dated August 1997. Later versions of NRC Form 3 that supersede the
August 1997 version shall replace the previously posted version within
30 days of receiving the revised NRC Form 3 from the Commission.
(2) Additional copies of NRC Form 3 may be obtained by writing to
the Regional Administrator of the appropriate U.S. Nuclear Regulatory
Commission Regional Office listed in appendix D to part 20 of this
chapter, by calling (301) 415-7232, via e-mail to [email protected], or by
visiting the NRC's Web site at http://www.nrc.gov and selecting forms
from the index found on the home page.
(f) Documents, notices, or forms posted under this section shall
appear in a sufficient number of places to permit individuals engaged
in NRC-licensed or regulated activities to observe them on the way to
or from any particular licensed or regulated activity location to which
the document applies, shall be conspicuous, and shall be replaced if
defaced or altered.
(g) Commission documents posted under paragraphs (a)(4) or (b)(3)
of this section shall be posted within 2 working days after receipt of
the documents from the Commission; the licensee's or regulated entity's
response, if any, shall be posted within 2 working days after dispatch
by the licensee or regulated entity. These documents shall remain
posted for a minimum of 5 working days or until action correcting the
violation has been completed, whichever is later.
0
45. Section 19.14 is revised to read as follows:
Sec. 19.14 Presence of representatives of licensees and regulated
entities, and workers during inspections.
(a) Each licensee, applicant for a license, applicant for or holder
of a standard design approval under subpart E of part 52 of this
chapter, applicant for an early site permit under subpart A of part 52
of this chapter, and applicant for a standard design certification
under subpart B of part 52 of this chapter shall afford to the
Commission at all
[[Page 49485]]
reasonable times opportunity to inspect materials, activities,
facilities, premises, and records under the regulations in this
chapter.
(b) During an inspection, Commission inspectors may consult
privately with workers as specified in Sec. 19.15. The licensee,
regulated entity, or the licensee's or regulated entity's
representative may accompany Commission inspectors during other phrases
of an inspection.
(c) If, at the time of inspection, an individual has been
authorized by the workers to represent them during Commission
inspections, the licensee or regulated entity shall notify the
inspectors of such authorization and shall give the workers'
representative an opportunity to accompany the inspectors during the
inspection of physical working conditions.
(d) Each workers' representative shall be routinely engaged in NRC-
licensed or regulated activities under control of the licensee or
regulated entity, and shall have received instructions as specified in
Sec. 19.12.
(e) Different representatives of licensees or regulated entities,
and workers may accompany the inspectors during different phases of an
inspection if there is no resulting interference with the conduct of
the inspection. However, only one workers' representative at a time may
accompany the inspectors.
(f) With the approval of the licensee or regulated entity, and the
workers' representative an individual who is not routinely engaged in
licensed or regulated activities under control of the license or
regulated entity (for example, a consultant to the licensee, the
regulated entity, or the workers' representative), shall be afforded
the opportunity to accompany Commission inspectors during the
inspection of physical working conditions.
(g) Notwithstanding the other provisions of this section,
Commission inspectors are authorized to refuse to permit accompaniment
by any individual who deliberately interferes with a fair and orderly
inspection. With regard to areas containing information classified by
an agency of the U.S. Government in the interest of national security,
an individual who accompanies an inspector may have access to such
information only if authorized to do so. With regard to any area
containing proprietary information, the workers' representative for
that area shall be an individual previously authorized by the licensee
or regulated entity to enter that area.
0
46. Section 19.20 is revised to read as follows:
Sec. 19.20 Employee protection.
Employment discrimination by a licensee, a holder of a certificate
of compliance issued under part 76 of this chapter or regulated entity
subject to the requirements in this part as delineated in Sec.
19.2(a), or a contractor or subcontractor of a licensee, a holder of a
certificate of compliance issued under part 76 of this chapter, or
regulated entity subject to the requirements in this part as delineated
in Sec. 19.2(a), against an employee for engaging in protected
activities under this part or parts 30, 40, 50, 52, 54, 60, 61, 63, 70,
72, 76, or 150 of this chapter is prohibited.
0
47. Section 19.31 is revised to read as follows:
Sec. 19.31 Application for exemptions.
The Commission may, upon application by any interested person or
upon its own initiative, grant such exemptions from the requirements of
the regulations in this part as it determines are authorized by law,
will not result in undue hazard to life and property.
0
48. Section 19.32 is revised to read as follows:
Sec. 19.32 Discrimination prohibited.
No person shall on the grounds of sex be excluded from
participation in, be denied a license, be denied the benefit of, or be
subjected to discrimination under any program or activity carried on
which is under the jurisdiction of the NRC under the Atomic Energy Act
of 1954, as amended, or under any title of the Energy Reorganization
Act of 1974, as amended. This provision will be enforced through agency
provisions and regulations similar to those already established, with
respect to racial and other discrimination, under Title VI of the Civil
Rights Act of 1964. This remedy is not exclusive, however, and will not
prejudice or cut off any other legal remedies available to a
discriminatee.
PART 20--STANDARDS FOR PROTECTION AGAINST RADIATION
0
49. The authority citation for Part 20 continues to read as follows:
Authority: Secs. 53, 63, 65, 81, 103, 104, 161, 182, 186, 68
Stat. 930, 933, 935, 936, 937, 948, 953, 955, as amended, sec. 1701,
106 Stat. 2951, 2952, 2953 (42 U.S.C. 2073, 2093, 2095, 2111, 2133,
2134, 2201, 2232, 2236, 2297f), secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
0
50. Section 20.1002 is revised to read as follows:
Sec. 20.1002 Scope.
The regulations in this part apply to persons licensed by the
Commission to receive, possess, use, transfer, or dispose of byproduct,
source, or special nuclear material or to operate a production or
utilization facility under parts 30 through 36, 39, 40, 50, 52, 60, 61,
63, 70, or 72 of this chapter, and in accordance with 10 CFR 76.60 to
persons required to obtain a certificate of compliance or an approved
compliance plan under part 76 of this chapter. The limits in this part
do not apply to doses due to background radiation, to exposure of
patients to radiation for the purpose of medical diagnosis or therapy,
to exposure from individuals administered radioactive material and
released under Sec. 35.75, or to exposure from voluntary participation
in medical research programs.
0
51. In Sec. 20.1401 paragraph (a) is revised to read as follows:
Sec. 20.1401 General provisions and scope.
(a) The criteria in this subpart apply to the decommissioning of
facilities licensed under parts 30, 40, 50, 52, 60, 61, 63, 70, and 72
of this chapter, and release of part of a facility or site for
unrestricted use in accordance with Sec. 50.83 of this chapter, as
well as other facilities subject to the Commission's jurisdiction under
the Atomic Energy Act of 1954, as amended, and the Energy
Reorganization Act of 1974, as amended. For high-level and low-level
waste disposal facilities (10 CFR parts 60, 61, and 63), the criteria
apply only to ancillary surface facilities that support radioactive
waste disposal activities. The criteria do not apply to uranium and
thorium recovery facilities already subject to appendix A to 10 CFR
part 40 or the uranium solution extraction facilities.
* * * * *
0
52. Section 20.1406 is revised to read as follows:
Sec. 20.1406 Minimization of contamination.
(a) Applicants for licenses, other than early site permits and
manufacturing licenses under part 52 of this chapter and renewals,
whose applications are submitted after August 20, 1997, shall describe
in the application how facility design and procedures for operation
will minimize, to the extent practicable, contamination of the facility
and the environment, facilitate eventual decommissioning, and minimize,
to the extent practicable, the generation of radioactive waste.
(b) Applicants for standard design certifications, standard design
[[Page 49486]]
approvals, and manufacturing licenses under part 52 of this chapter,
whose applications are submitted after August 20, 1997, shall describe
in the application how facility design will minimize, to the extent
practicable, contamination of the facility and the environment,
facilitate eventual decommissioning, and minimize, to the extent
practicable, the generation of radioactive waste.
0
53. In Sec. 20.2203, paragraphs (c) and (d) are revised to read as
follows:
Sec. 20.2203 Reports of exposures, radiation levels, and
concentrations of radioactive material exceeding the constraints or
limits.
* * * * *
(c) For holders of an operating license or a combined license for a
nuclear power plant, the occurrences included in paragraph (a) of this
section must be reported in accordance with the procedures described in
Sec. Sec. 50.73(b), (c), (d), (e), and (g) of this chapter, and must
include the information required by paragraph (b) of this section.
Occurrences reported in accordance with Sec. 50.73 of this chapter
need not be reported by a duplicate report under paragraph (a) of this
section.
(d) All licensees, other than those holding an operating license or
a combined license for a nuclear power plant, who make reports under
paragraph (a) of this section shall submit the report in writing either
by mail addressed to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; by hand delivery to
the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; or,
where practicable, by electronic submission, for example, Electronic
Information Exchange, or CD-ROM. Electronic submissions must be made in
a manner that enables the NRC to receive, read, authenticate,
distribute, and archive the submission, and process and retrieve it a
single page at a time. Detailed guidance on making electronic
submissions can be obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/e-submittals.html, by calling (301) 415-0439, by
e-mail to [email protected], or by writing the Office of Information
Services, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001. A copy should be sent to the appropriate NRC Regional Office
listed in appendix D to this part.
PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE
0
54. The authority citation for part 21 continues to read as follows:
Authority: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83
Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C.
2201, 2282, 2297f); secs. 201, as amended, 206, 88 Stat. 1242, as
amended 1246 (42 U.S.C. 5841, 5846); sec. 1704, 112 Stat. 2750 (44
U.S.C. 3504 note).
Section 21.2 also issued under secs. 135, 141, Pub. L. 97-425,
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
0
55. In Sec. 21.2, paragraphs (a), (b), and (c) are revised to read as
follows:
Sec. 21.2 Scope.
(a) The regulations in this part apply, except as specifically
provided otherwise in parts 31, 34, 35, 39, 40, 60, 61, 63, 70, or part
72 of this chapter, to:
(1) Each individual, partnership, corporation, or other entity
applying for or holding a license or permit under the regulations in
this chapter to possess, use, or transfer within the United States
source material, byproduct material, special nuclear material, and/or
spent fuel and high-level radioactive waste, or to construct,
manufacture, possess, own, operate, or transfer within the United
States, any production or utilization facility or independent spent
fuel storage installation (ISFSI) or monitored retrievable storage
installation (MRS); and each director and responsible officer of such a
licensee;
(2) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, that constructs a
production or utilization facility licensed for manufacture,
construction, or operation under parts 50 or 52 of this chapter, an
ISFSI for the storage of spent fuel licensed under part 72 of this
chapter, an MRS for the storage of spent fuel or high-level radioactive
waste under part 72 of this chapter, or a geologic repository for the
disposal of high-level radioactive waste under part 60 or 63 of this
chapter; or supplies basic components for a facility or activity
licensed, other than for export, under parts 30, 40, 50, 52, 60, 61,
63, 70, 71, or part 72 of this chapter;
(3) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, applying for a design
certification rule under part 52 of this chapter; or supplying basic
components with respect to that design certification, and each
individual, corporation, partnership, or other entity doing business
within the United States, and each director and responsible officer of
such an organization, whose application for design certification has
been granted under part 52 of this chapter, or who has supplied or is
supplying basic components with respect to that design certification;
(4) Each individual, corporation, partnership, or other entity
doing business within the United States, and each director and
responsible officer of such an organization, applying for or holding a
standard design approval under part 52 of this chapter; or supplying
basic components with respect to a standard design approval under part
52 of this chapter;
(b) For persons licensed to construct a facility under either a
construction permit issued under Sec. 50.23 of this chapter or a
combined license under part 52 of this chapter (for the period of
construction until the date that the Commission makes the finding under
Sec. 52.103(g) of this chapter), or to manufacture a facility under
part 52 of this chapter, evaluation of potential defects and failures
to comply and reporting of defects and failures to comply under Sec.
50.55(e) of this chapter satisfies each person's evaluation,
notification, and reporting obligation to report defects and failures
to comply under this part and the responsibility of individual
directors and responsible officers of these licensees to report defects
under Section 206 of the Energy Reorganization Act of 1974.
(c) For persons licensed to operate a nuclear power plant under
part 50 or part 52 of this chapter, evaluation of potential defects and
appropriate reporting of defects under Sec. Sec. 50.72, 50.73, or
Sec. 73.71 of this chapter, satisfies each person's evaluation,
notification, and reporting obligation to report defects under this
part, and the responsibility of individual directors and responsible
officers of these licensees to report defects under Section 206 of the
Energy Reorganization Act of 1974.
* * * * *
0
56. In Sec. 21.3 the definitions of basic component, defect,
deviation, and substantial safety hazard are revised to read as
follows:
Sec. 21.3 Definitions.
* * * * *
Basic component. (1)(i) When applied to nuclear power plants
licensed under 10 CFR part 50 or part 52 of this chapter, basic
component means a structure, system, or component, or part thereof that
affects its safety function necessary to assure:
(A) The integrity of the reactor coolant pressure boundary;
(B) The capability to shut down the reactor and maintain it in a
safe-shutdown condition; or
[[Page 49487]]
(C) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. Sec. 50.34(a)(1), 50.67(b)(2), or 100.11
of this chapter, as applicable.
(ii) Basic components are items designed and manufactured under a
quality assurance program complying with appendix B to part 50 of this
chapter, or commercial grade items which have successfully completed
the dedication process.
(2) When applied to standard design certifications under subpart C
of part 52 of this chapter and standard design approvals under part 52
of this chapter, basic component means the design or procurement
information approved or to be approved within the scope of the design
certification or approval for a structure, system, or component, or
part thereof, that affects its safety function necessary to assure:
(i) The integrity of the reactor coolant pressure boundary;
(ii) The capability to shut down the reactor and maintain it in a
safe-shutdown condition; or
(iii) The capability to prevent or mitigate the consequences of
accidents which could result in potential offsite exposures comparable
to those referred to in Sec. Sec. 50.34(a)(1), 50.67(b)(2), or 100.11
of this chapter, as applicable.
(3) When applied to other facilities and other activities licensed
under 10 CFR parts 30, 40, 50 (other than nuclear power plants), 60,
61, 63, 70, 71, or 72 of this chapter, basic component means a
structure, system, or component, or part thereof, that affects their
safety function, that is directly procured by the licensee of a
facility or activity subject to the regulations in this part and in
which a defect or failure to comply with any applicable regulation in
this chapter, order, or license issued by the Commission could create a
substantial safety hazard.
(4) In all cases, basic component includes safety-related design,
analysis, inspection, testing, fabrication, replacement of parts, or
consulting services that are associated with the component hardware,
design certification, design approval, or information in support of an
early site permit application under part 52 of this chapter, whether
these services are performed by the component supplier or others.
* * * * *
Defect means:
(1) A deviation in a basic component delivered to a purchaser for
use in a facility or an activity subject to the regulations in this
part if, on the basis of an evaluation, the deviation could create a
substantial safety hazard;
(2) The installation, use, or operation of a basic component
containing a defect as defined in this section;
(3) A deviation in a portion of a facility subject to the early
site permit, standard design certification, standard design approval,
construction permit, combined license or manufacturing licensing
requirements of part 50 or part 52 of this chapter, provided the
deviation could, on the basis of an evaluation, create a substantial
safety hazard and the portion of the facility containing the deviation
has been offered to the purchaser for acceptance;
(4) A condition or circumstance involving a basic component that
could contribute to the exceeding of a safety limit, as defined in the
technical specifications of a license for operation issued under part
50 or part 52 of this chapter; or
(5) An error, omission or other circumstance in a design
certification, or standard design approval that, on the basis of an
evaluation, could create a substantial safety hazard.
Deviation means a departure from the technical requirements
included in a procurement document, or specified in early site permit
information, a standard design certification or standard design
approval.
* * * * *
Substantial safety hazard means a loss of safety function to the
extent that there is a major reduction in the degree of protection
provided to public health and safety for any facility or activity
licensed or otherwise approved or regulated by the NRC, other than for
export, under parts 30, 40, 50, 52, 60, 61, 63, 70, 71, or 72 of this
chapter.
* * * * *
0
57. Section 21.5 is revised to read as follows:
Sec. 21.5 Communications.
Except where otherwise specified in this part, written
communications and reports concerning the regulations in this part must
be addressed to the NRC's Document Control Desk, and sent by mail to
the U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; by
hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville,
Maryland; or, where practicable, by electronic submission, for example,
Electronic Information Exchange, or CD-ROM. Electronic submissions must
be made in a manner that enables the NRC to receive, read,
authenticate, distribute, and archive the submission, and process and
retrieve it a single page at a time. Detailed guidance on making
electronic submissions can be obtained by visiting the NRC's Web site
at http://www.nrc.gov/site-help/eie.html, by calling (301) 415-6030, by
e-mail to [email protected], or by writing the Office of Information
Services, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001. The guidance discusses, among other topics, the formats the NRC
can accept, the use of electronic signatures, and the treatment of
nonpublic information. In the case of a licensee or permit holder, a
copy of the communication must also be sent to the appropriate Regional
Administrator at the address specified in appendix D to part 20 of this
chapter.
0
58. In Sec. 21.21 the introductory text of paragraph (a)(3), paragraph
(a)(3)(i), and paragraphs (d)(1)(i), (d)(1)(ii), and (d)(4)(vi) are
revised and paragraph (d)(4)(ix) is added to read as follows:
Sec. 21.21 Notification of failure to comply or existence of a defect
and its evaluation.
(a) * * *
(3) Ensure that a director or responsible officer subject to the
regulations of this part is informed as soon as practicable, and, in
all cases, within the 5 working days after completion of the evaluation
described in paragraphs (a)(1) or (a)(2) of this section if the
manufacture, construction, or operation of a facility or activity, a
basic component supplied for such facility or activity, or the design
certification or design approval under part 52 of this chapter--
(i) Fails to comply with the Atomic Energy Act of 1954, as amended,
or any applicable rule, regulation, order, or license of the Commission
or standard design approval under part 52 of this chapter, relating to
a substantial safety hazard, or
* * * * *
(d)(1) * * *
(i) The manufacture, construction or operation of a facility or an
activity within the United States that is subject to the licensing
requirements under parts 30, 40, 50, 52, 60, 61, 63, 70, 71, or 72 of
this chapter and that is within his or her organization's
responsibility; or
(ii) A basic component that is within his or her organization's
responsibility and is supplied for a facility or an activity within the
United States that is subject to the licensing, design certification,
or approval requirements under parts 30, 40, 50, 52, 60, 61, 63, 70,
71, or 72 of this chapter.
* * * * *
(4) * * *
(vi) In the case of a basic component which contains a defect or
fails to
[[Page 49488]]
comply, the number and location of these components in use at, supplied
for, being supplied for, or may be supplied for, manufactured, or being
manufactured for one or more facilities or activities subject to the
regulations in this part.
* * * * *
(ix) In the case of an early site permit, the entities to whom an
early site permit was transferred.
* * * * *
0
59. In Sec. 21.51 paragraphs (a)(4) and (a)(5) are added and paragraph
(b) is revised to read as follows:
Sec. 21.51 Maintenance and inspection of records.
(a) * * *
(4) Applicants for standard design certification under subpart B of
part 52 of this chapter and others providing a design which is the
subject of a design certification, during and following Commission
adoption of a final design certification rule for that design, shall
retain any notifications sent to purchasers and affected licensees for
a minimum of 5 years after the date of the notification, and retain a
record of the purchasers for 15 years after delivery of design which is
the subject of the design certification rule or service associated with
the design.
(5) Applicants for or holders of a standard design approval under
subpart E of part 52 of this chapter and others providing a design
which is the subject of a design approval shall retain any
notifications sent to purchasers and affected licensees for a minimum
of 5 years after the date of the notification, and retain a record of
the purchasers for 15 years after delivery of the design which is the
subject of the design approval or service associated with the design.
(b) Each individual, corporation, partnership, dedicating entity,
or other entity subject to the regulations in this part shall permit
the Commission the opportunity to inspect records pertaining to basic
components that relate to the identification and evaluation of
deviations, and the reporting of defects and failures to comply,
including (but not limited to) any advice given to purchasers or
licensees on the placement, erection, installation, operation,
maintenance, modification, or inspection of a basic component.
0
60. In Sec. 21.61, paragraph (b) is revised to read as follows:
Sec. 21.61 Failure to notify.
* * * * *
(b) Any NRC licensee or applicant for a license (including an
applicant for, or holder of, a permit), applicant for a design
certification under part 52 of this chapter during the pendency of its
application, applicant for a design certification after Commission
adoption of a final design certification rule for that design, or
applicant for or holder of a standard design approval under part 52 of
this chapter subject to the regulations in this part who fails to
provide the notice required by Sec. 21.21, or otherwise fails to
comply with the applicable requirements of this part shall be subject
to a civil penalty as provided by Section 234 of the Atomic Energy Act
of 1954, as amended.
* * * * *
PART 25--ACCESS AUTHORIZATION
0
61. The authority citation for part 25 continues to read as follows:
Authority: Secs. 145, 161, 68 Stat. 942, 948, as amended (42
U.S.C. 2165, 2201); sec. 201, 88 Stat. 1242, as amended (42 U.S.C.
5841); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); E.O. 10865,
as amended, 3 CFR 1959-1963 Comp., p. 398 (50 U.S.C. 401, note);
E.O. 12829, 3 CFR, 1993 Comp., p. 570; E.O. 12958, as amended, 3
CFR, 1995 Comp., p. 333 as amended by E.O. 13292, 3 CFR 2004 Comp.,
p. 196; E.O. 12968, 3 CFR, 1995 Comp., p. 396.
Appendix A also issued under 96 Stat. 1051 (31 U.S.C. 9701).
0
62. The heading of part 25 is revised to read as set forth above.
0
63. In Sec. 25.35, paragraph (a) is revised to read as follows:
Sec. 25.35 Classified visits.
(a) The number of classified visits must be held to a minimum. The
licensee, certificate holder, applicant for a standard design
certification under part 52 of this chapter (including an applicant
after the Commission has adopted a final standard design certification
rule under part 52 of this chapter), or other facility, or an applicant
for or holder of a standard design approval under part 52 of this
chapter shall determine that the visit is necessary and that the
purpose of the visit cannot be achieved without access to, or
disclosure of, classified information. All classified visits require
advance notification to, and approval of, the organization to be
visited. In urgent cases, visit information may be furnished by
telephone and confirmed in writing.
* * * * *
PART 26--FITNESS FOR DUTY PROGRAMS
0
64. The authority citation for part 26 continues to read as follows:
Authority: Secs. 53, 81, 103, 104, 107, 161, 68 Stat. 930, 935,
936, 937, 948, as amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42
U.S.C. 2073, 2111, 2112, 2133, 2134, 2137, 2201, 2297f); secs. 201,
202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 5841,
5842, 5846).
0
65. In Sec. 26.2, the introductory text of paragraph (a), and
paragraph (c) are revised to read as follows:
Sec. 26.2 Scope.
(a) The regulations in this part apply to licensees authorized to
operate a nuclear power reactor, including a holder of a combined
license after the Commission makes the finding under Sec. 52.103(g) of
this chapter, and licensees who are authorized to possess or use
formula quantities of SSNM, or to transport formula quantities of SSNM.
Each licensee shall implement a fitness-for-duty program which complies
with this part. The provisions of the fitness-for-duty program must
apply to all persons granted unescorted access to nuclear power plant
protected areas, to licensee, vendor, or contractor personnel required
to physically report to a licensee's Technical Support Center (TSC) or
Emergency Operations Facility (EOF) in accordance with licensee
emergency plans and procedures, and to SSNM licensee and transporter
personnel who:
* * * * *
(c) Certain regulations in this part apply to licensees holding
permits to construct a nuclear power plant, including a holder of a
combined license before the date that the Commission makes the finding
under Sec. 52.103(g) of this chapter, holders of manufacturing
licenses under part 52, and persons authorized to conduct the
activities under Sec. 50.10(e)(3) of this chapter. Each licensee with
a construction permit, a combined license before the Commission makes
the finding under Sec. 52.103(g) of this chapter, a manufacturing
license, or person authorized to conduct the activities under Sec.
50.10(e)(3) of this chapter, with a plant or reactor under active
construction or manufacture, shall--
(1) Comply with Sec. Sec. 26.10, 26.20, 26.23, 26.70, and 26.73;
(2) Implement a chemical testing program, including random tests;
and
(3) Make provisions for employee assistance programs, imposition of
sanctions, appeals procedures, the protection of information, and
recordkeeping.
* * * * *
0
66. In Sec. 26.10, paragraph (a) is revised to read as follows:
[[Page 49489]]
Sec. 26.10 General performance objectives.
* * * * *
(a) Provide reasonable assurance that nuclear power plant
personnel, personnel of a holder of a manufacturing license, personnel
of a person authorized to conduct activities under Sec. 50.10(e)(3) of
this chapter, transporter personnel, and personnel of licensees
authorized to possess or use formula quantities of SSNM, will perform
their tasks in a reliable and trustworthy manner and are not under the
influence of any substance, legal or illegal, or mentally or physically
impaired from any cause, which in any way adversely affects their
ability to safely and competently perform their duties;
* * * * *
0
67. In Appendix A of Part 26, paragraph (1) of Section 1.1 of Subpart A
is revised to read as follows:
Appendix A to Part 26--Guidelines for Drug and Alcohol Testing Programs
1.1 Applicability.
(1) These guidelines apply to licensees authorized to operate
nuclear power reactors, including a holder of a combined license
after the Commission makes the finding under Sec. 52.103(g) of this
chapter, and licensees who are authorized to possess, use, or
transport formula quantities of strategic special nuclear material
(SSNM).
* * * * *
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION
FACILITIES
0
68. The authority citation for part 50 continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234,
83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201,
2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
Section 50.7 also issued under Pub. L. 95--601, sec. 10, 92
Stat. 2951 (42 U.S.C. 5841). Section 50.10 also issued under secs.
101, 185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102,
Pub. L. 91--190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13,
50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as
amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56
also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections
50.33a, 50.55a and appendix Q also issued under sec. 102, Pub. L.
91--190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54
also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections
50.58, 50.91, and 50.92 also issued under Pub. L. 97--415, 96 Stat.
2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68
Stat. 939 (42 U.S.C. 2152). Sections 50.80--50.81 also issued under
sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also
issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
0
69. In Section 50.2, definitions of applicant, license, licensee, and
prototype plant, are added to read as follows:
Sec. 50.2 Definitions.
* * * * *
Applicant means a person or an entity applying for a license,
permit, or other form of Commission permission or approval under this
part or part 52 of this chapter.
* * * * *
License means a license, including a construction permit or
operating license under this part, an early site permit, combined
license or manufacturing license under part 52 of this chapter, or a
renewed license issued by the Commission under this part, part 52, or
part 54 of this chapter.
Licensee means a person who is authorized to conduct activities
under a license issued by the Commission.
* * * * *
Prototype plant means a nuclear reactor that is used to test design
features, such as the testing required under Sec. 50.43(e). The
prototype plant is similar to a first-of-a-kind or standard plant
design in all features and size, but may include additional safety
features to protect the public and the plant staff from the possible
consequences of accidents during the testing period.
* * * * *
0
70. In Sec. 50.10 the introductory text of paragraphs (b) and (c), and
paragraphs (e)(1), (e)(2), and (e)(3) are revised to read as follows:
Sec. 50.10 License required.
* * * * *
(b) No person shall begin the construction of a production or
utilization facility on a site on which the facility is to be operated
until either a construction permit under this part, or a combined
license under subpart C of part 52 of this chapter has been issued. As
used in this paragraph, the term ``construction'' includes pouring the
foundation for, or the installation of, any portion of the permanent
facility on the site, but does not include:
* * * * *
(c) Notwithstanding the provisions of paragraph (b) of this
section, and subject to paragraphs (d) and (e) of this section, no
person shall effect commencement of construction of a production or
utilization facility subject to the provisions of Sec. 51.20(b) of
this chapter on a site on which the facility is to be operated until an
early site permit, construction permit, or combined license has been
issued. As used in this paragraph, the term ``commencement of
construction'' means any clearing of land, excavation or other
substantial action that would adversely affect the environment of a
site, but does not include:
* * * * *
(e)(1) The Director of Nuclear Reactor Regulation may authorize an
applicant for a construction permit for a utilization facility which is
subject to Sec. 51.20(b) of this chapter, and is of the type specified
in Sec. Sec. 50.21(b)(2) or (3), or Sec. 50.22 or is a testing
facility, or an applicant for a combined license to conduct the
following activities:
(i) Preparation of the site for construction of the facility
(including activities as clearing, grading, construction of temporary
access roads and borrow areas);
(ii) Installation of temporary construction support facilities
(including items such as warehouse and shop facilities, utilities,
concrete mixing plants, docking and unloading facilities, and
construction support buildings);
(iii) Excavation for facility structures;
(iv) Construction of service facilities (including facilities such
as roadways, paving, railroad spurs, fencing, exterior utility and
lighting systems, transmission lines, and sanitary sewerage treatment
facilities); and
(v) The construction of structures, systems and components which do
not prevent or mitigate the consequences of postulated accidents that
could cause undue risk to the health and safety of the public.
(2) No authorization shall be granted unless the staff has
completed a final environmental impact statement on the issuance of the
construction permit or combined license as required by subpart A of
part 51 of this chapter. An authorization shall be granted only after
the presiding officer in the proceeding on the construction permit or
combined license application:
(i) Has made all the findings required by Sec. Sec. 51.104(b),
51.105, and 51.107 of this chapter to be made before issuance of the
construction permit, or combined license for the facility; and
(ii) Has determined that, based upon the available information and
review to date, there is reasonable assurance that the proposed site is
a suitable location for a reactor of the general size and type proposed
from the standpoint of radiological health and safety considerations
under the Act and regulations issued by the Commission.
(3)(i) The Director of New Reactors or the Director of Nuclear
Reactor
[[Page 49490]]
Regulation, as appropriate, may authorize an applicant for a
construction permit for a utilization facility which is subject to
Sec. 51.20(b) of this chapter, and is of the type specified in
Sec. Sec. 50.21(b)(2) or (3), or Sec. 50.22 or is a testing facility,
or an applicant for a combined license to conduct, in addition to the
activities described in paragraph (e)(1) of this section, the
installation of structural foundations, including any necessary
subsurface preparation, for structures, systems, and components which
prevent or mitigate the consequences of postulated accidents that could
cause undue risk to the health and safety of the public.
(ii) Such an authorization, which may be combined with the
authorization described in paragraph (e)(1) of this section, or may be
granted at a later time, shall be granted only after the presiding
officer in the proceeding on the construction permit or combined
license application has, in addition to making the findings and
determinations required by paragraph (e)(2) of this section, determined
that there are no unresolved safety issues relating to the additional
activities that may be authorized under this paragraph that would
constitute good cause for withholding authorization.
* * * * *
0
71. Section 50.23 is revised to read as follows:
Sec. 50.23 Construction permits.
A construction permit for the construction of a production or
utilization facility will be issued before the issuance of a license if
the application is otherwise acceptable, and will be converted upon
completion of the facility and Commission action, into a license as
provided in Sec. 50.56. However, if a combined license for a nuclear
power reactor is issued under part 52 of this chapter, the construction
permit and operating license are deemed to be combined in a single
license. A construction permit for the alteration of a production or
utilization facility will be issued before the issuance of an amendment
of a license, if the application for amendment is otherwise acceptable,
as provided in Sec. 50.91.
0
72. The undesignated center heading before Sec. 50.30 is revised to
read as follows:
Applications for Licenses, Certifications, and Regulatory Approvals;
Form; Contents; Ineligibility of Certain Applicants
0
73. In Sec. 50.30, the section heading and paragraphs (a)(1), (a)(3),
(a)(5), (a)(6), (b), (e), and (f) are revised to read as follows:
Sec. 50.30 Filing of application; oath or affirmation.
(a) * * *
(1) Each filing of an application for a standard design approval or
license to construct and/or operate, or manufacture, a production or
utilization facility (including an early site permit, combined license,
and manufacturing license under part 52 of this chapter), and any
amendments to the applications, must be submitted to the U.S. Nuclear
Regulatory Commission in accordance with Sec. 50.4 or Sec. 52.3 of
this chapter, as applicable.
* * * * *
(3) Each applicant for a construction permit under this part, or an
early site permit, combined license, or manufacturing license under
part 52 of this chapter, shall, upon notification by the Atomic Safety
and Licensing Board appointed to conduct the public hearing required by
the Atomic Energy Act, update the application and serve the updated
copies of the application or parts of it, eliminating all superseded
information, together with an index of the updated application, as
directed by the Atomic Safety and Licensing Board. Any subsequent
amendment to the application must be served on those served copies of
the application and must be submitted to the U.S. Nuclear Regulatory
Commission as specified in Sec. 50.4 or Sec. 52.3 of this chapter, as
applicable.
* * * * *
(5) At the time of filing an application, the Commission will make
available at the NRC Web site, http://www.nrc.gov, a copy of the
application, subsequent amendments, and other records pertinent to the
matter which is the subject of the application for public inspection
and copying.
(6) The serving of copies required by this section must not occur
until the application has been docketed under Sec. 2.101(a) of this
chapter. Copies must be submitted to the Commission, as specified in
Sec. 50.4 or Sec. 52.3 of this chapter, as applicable, to enable the
Director, Office of New Reactors, or the Director, Office of Nuclear
Reactor Regulation, or the Director, Office of Nuclear Material Safety
and Safeguards, as appropriate, to determine whether the application is
sufficiently complete to permit docketing.
(b) Oath or affirmation. Each application for a standard design
approval or license, including, whenever appropriate, a construction
permit or early site permit, or amendment of it, and each amendment of
each application must be executed in a signed original by the applicant
or duly authorized officer thereof under oath or affirmation.
* * * * *
(e) Filing Fees. Each application for a standard design approval or
production or utilization facility license, including, whenever
appropriate, a construction permit or early site permit, other than a
license exempted from part 170 of this chapter, shall be accompanied by
the fee prescribed in part 170 of this chapter. No fee will be required
to accompany an application for renewal, amendment, or termination of a
construction permit, operating license, combined license, or
manufacturing license, except as provided in Sec. 170.21 of this
chapter.
(f) Environmental report. An application for a construction permit,
operating license, early site permit, combined license, or
manufacturing license for a nuclear power reactor, testing facility,
fuel reprocessing plant, or other production or utilization facility
whose construction or operation may be determined by the Commission to
have a significant impact in the environment, shall be accompanied by
an Environmental Report required under subpart A of part 51 of this
chapter.
0
74. In Sec. 50.33, paragraphs (f)(3) and (f)(4) are redesignated as
(f)(4)and (f)(5), respectively, and are revised, a new paragraph (f)(3)
is added, and paragraphs (g), (h), and (k)(1) are revised to read as
follows:
Sec. 50.33 Contents of applications; general information.
* * * * *
(f) * * *
(3) If the application is for a combined license under subpart C of
part 52 of this chapter, the applicant shall submit the information
described in paragraphs (f)(1) and (f)(2) of this section.
(4) Each application for a construction permit, operating license,
or combined license submitted by a newly-formed entity organized for
the primary purpose of constructing and/or operating a facility must
also include information showing:
(i) The legal and financial relationships it has or proposes to
have with its stockholders or owners;
(ii) The stockholders' or owners' financial ability to meet any
contractual obligation to the entity which they have incurred or
proposed to incur; and
[[Page 49491]]
(iii) Any other information considered necessary by the Commission
to enable it to determine the applicant's financial qualification.
(5) The Commission may request an established entity or newly-
formed entity to submit additional or more detailed information
respecting its financial arrangements and status of funds if the
Commission considers this information appropriate. This may include
information regarding a licensee's ability to continue the conduct of
the activities authorized by the license and to decommission the
facility.
(g) If the application is for an operating license or combined
license for a nuclear power reactor, or if the application is for an
early site permit and contains plans for coping with emergencies under
Sec. 52.17(b)(2)(ii) of this chapter, the applicant shall submit
radiological emergency response plans of State and local governmental
entities in the United States that are wholly or partially within the
plume exposure pathway emergency planning zone (EPZ),\4\ as well as the
plans of State governments wholly or partially within the ingestion
pathway EPZ.\5\ If the application is for an early site permit that,
under 10 CFR 52.17(b)(2)(i), proposes major features of the emergency
plans describing the EPZs, then the descriptions of the EPZs must meet
the requirements of this paragraph. Generally, the plume exposure
pathway EPZ for nuclear power reactors shall consist of an area about
10 miles (16 km) in radius and the ingestion pathway EPZ shall consist
of an area about 50 miles (80 km) in radius. The exact size and
configuration of the EPZs surrounding a particular nuclear power
reactor shall be determined in relation to the local emergency response
needs and capabilities as they are affected by such conditions as
demography, topography, land characteristics, access routes, and
jurisdictional boundaries. The size of the EPZs also may be determined
on a case-by-case basis for gas-cooled reactors and for reactors with
an authorized power level less than 250 MW thermal. The plans for the
ingestion pathway shall focus on such actions as are appropriate to
protect the food ingestion pathway.
---------------------------------------------------------------------------
\4\ Emergency planning zones (EPZs) are discussed in NUREG-0396,
EPA 520/1-78-016, ``Planning Basis for the Development of State and
Local Government Radiological Emergency Response Plans in Support of
Light-Water Nuclear Power Plants,'' December 1978.
\5\ If the State and local emergency response plans have been
previously provided to the NRC for inclusion in the facility docket,
the applicant need only provide the appropriate reference to meet
this requirement.
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(h) If the applicant, other than an applicant for a combined
license, proposes to construct or alter a production or utilization
facility, the application shall state the earliest and latest dates for
completion of the construction or alteration.
* * * * *
(k)(1) For an application for an operating license or combined
license for a production or utilization facility, information in the
form of a report, as described in Sec. 50.75, indicating how
reasonable assurance will be provided that funds will be available to
decommission the facility.
* * * * *
0
75. In Sec. 50.34, the section heading, the introductory text of
paragraph (a)(1), paragraphs (a)(1)(ii)(E) and (a)(12), the
introductory text of paragraph (b), paragraphs (b)(10) and (b)(11), and
paragraphs (c), (d), and (e), the introductory text of paragraphs (f)
and(f)(1), and paragraphs (g), and (h)(1)(ii) are revised to read as
follows:
Sec. 50.34 Contents of construction permit and operating license
applications; technical information.
(a) * * *
(1) Stationary power reactor applicants for a construction permit
who apply on or after January 10, 1997, shall comply with paragraph
(a)(1)(ii) of this section. All other applicants for a construction
permit shall comply with paragraph (a)(1)(i) of this section.
* * * * *
(ii) * * *
(E) With respect to operation at the projected initial power level,
the applicant is required to submit information prescribed in
paragraphs (a)(2) through (a)(8) of this section, as well as the
information required by paragraph (a)(1)(i) of this section, in support
of the application for a construction permit.
* * * * *
(12) On or after January 10, 1997, stationary power reactor
applicants who apply for a construction permit, as partial conformance
to General Design Criterion 2 of appendix A to this part, shall comply
with the earthquake engineering criteria in appendix S to this part.
(b) Final safety analysis report. Each application for an operating
license shall include a final safety analysis report. The final safety
analysis report shall include information that describes the facility,
presents the design bases and the limits on its operation, and presents
a safety analysis of the structures, systems, and components and of the
facility as a whole, and shall include the following:
* * * * *
(10) On or after January 10, 1997, stationary power reactor
applicants who apply for an operating license, as partial conformance
to General Design Criterion 2 of appendix A to this part, shall comply
with the earthquake engineering criteria of appendix S to this part.
However, for those operating license applicants and holders whose
construction permit was issued before January 10, 1997, the earthquake
engineering criteria in Section VI of appendix A to part 100 of this
chapter continues to apply.
(11) On or after January 10, 1997, stationary power reactor
applicants who apply for an operating license, shall provide a
description and safety assessment of the site and of the facility as in
Sec. 50.34(a)(1)(ii). However, for either an operating license
applicant or holder whose construction permit was issued before January
10, 1997, the reactor site criteria in part 100 of this chapter and the
seismic and geologic siting criteria in appendix A to part 100 of this
chapter continues to apply.
(c) Physical Security Plan. Each application for an operating
license for a production or utilization facility must include a
physical security plan. The plan must describe how the applicant will
meet the requirements of part 73 of this chapter (and part 11 of this
chapter, if applicable, including the identification and description of
jobs as required by Sec. 11.11(a) of this chapter, at the proposed
facility). The plan must list tests, inspections, audits, and other
means to be used to demonstrate compliance with the requirements of 10
CFR parts 11 and 73, if applicable.
(d) Safeguards contingency plan. Each application for an operating
license for a production or utilization facility that will be subject
to Sec. Sec. 73.50, 73.55, or Sec. 73.60 of this chapter, must
include a licensee safeguards contingency plan in accordance with the
criteria set forth in appendix C to 10 CFR part 73. The safeguards
contingency plan shall include plans for dealing with threats, thefts,
and radiological sabotage, as defined in part 73 of this chapter,
relating to the special nuclear material and nuclear facilities
licensed under this chapter and in the applicant's possession and
control. Each application for such a license shall include the first
four categories of information contained in the applicant's safeguards
contingency plan. (The first four categories of information as set
forth in appendix C to 10 CFR part 73
[[Page 49492]]
of this chapter are Background, Generic Planning Base, Licensee
Planning Base, and Responsibility Matrix. The fifth category of
information, Procedures, does not have to be submitted for approval.)
\9\
---------------------------------------------------------------------------
\9\ A physical security plan that contains all the information
required in both Sec. 73.55 and appendix C to part 73 of this
chapter satisfies the requirement for a contingency plan.
---------------------------------------------------------------------------
(e) Protection against unauthorized disclosure. Each applicant for
an operating license for a production or utilization facility, who
prepares a physical security plan, a safeguards contingency plan, or a
guard qualification and training plan, shall protect the plans and
other related safeguards information against unauthorized disclosure in
accordance with the requirements of Sec. 73.21 of this chapter, as
appropriate.
(f) Additional TMI-related requirements. In addition to the
requirements of paragraph (a) of this section, each applicant for a
light-water-reactor construction permit or manufacturing license whose
application was pending as of February 16, 1982, shall meet the
requirements in paragraphs (f)(1) through (3) of this section. This
regulation applies to the pending applications by Duke Power Company
(Perkins Nuclear Station, Units 1, 2, and 3), Houston Lighting & Power
Company (Allens Creek Nuclear Generating Station, Unit 1), Portland
General Electric Company (Pebble Springs Nuclear Plant, Units 1 and 2),
Public Service Company of Oklahoma (Black Fox Station, Units 1 and 2),
Puget Sound Power & Light Company (Skagit/Hanford Nuclear Power
Project, Units 1 and 2), and Offshore Power Systems (License to
Manufacture Floating Nuclear Plants). The number of units that will be
specified in the manufacturing license above, if issued, will be that
number whose start of manufacture, as defined in the license
application, can practically begin within a 10-year period commencing
on the date of issuance of the manufacturing license, but in no event
will that number be in excess of ten. The manufacturing license will
require the plant design to be updated no later than 5 years after its
approval. Paragraphs (f)(1)(xii), (2)(ix), and (3)(v) of this section,
pertaining to hydrogen control measures, must be met by all applicants
covered by this regulation. However, the Commission may decide to
impose additional requirements and the issue of whether compliance with
these provisions, together with 10 CFR 50.44 and criterion 50 of
appendix A to 10 CFR part 50, is sufficient for issuance of that
manufacturing license which may be considered in the manufacturing
license proceeding. In addition, each applicant for a design
certification, design approval, combined license, or manufacturing
license under part 52 of this chapter shall demonstrate compliance with
the technically relevant portions of the requirements in paragraphs
(f)(1) through (3) of this section, except for paragraphs (f)(1)(xii),
(f)(2)(ix), and (f)(3)(v).
(1) To satisfy the following requirements, the application shall
provide sufficient information to describe the nature of the studies,
how they are to be conducted, estimated submittal dates, and a program
to ensure that the results of these studies are factored into the final
design of the facility. For licensees identified in the introduction to
paragraph (f) of this section, all studies must be completed no later
than 2 years following the issuance of the construction permit or
manufacturing license.\10\ For all other applicants, the studies must
be submitted as part of the final safety analysis report.
---------------------------------------------------------------------------
\10\ Alphanumeric designations correspond to the related action
plan items in NUREG 0718 and NUREG-0660, ``NRC Action Plan Developed
as a Result of the TMI-2 Accident.'' They are provided herein for
information only.
---------------------------------------------------------------------------
* * * * *
(g) Combustible gas control. All applicants for a reactor
construction permit or operating license whose application is submitted
after October 16, 2003, shall include the analyses, and the
descriptions of the equipment and systems required by Sec. 50.44 as a
part of their application.
(h) * * *
(1) * * *
(ii) Applications for light-water-cooled nuclear power plant
construction permits docketed after May 17, 1982, shall include an
evaluation of the facility against the SRP in effect on May 17, 1982,
or the SRP revision in effect six months before the docket date of the
application, whichever is later.
* * * * *
0
76. Section 50.34a is revised to read as follows:
Sec. 50.34a Design objectives for equipment to control releases of
radioactive material in effluents--nuclear power reactors.
(a) An application for a construction permit shall include a
description of the preliminary design of equipment to be installed to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations, including expected
operational occurrences. In the case of an application filed on or
after January 2, 1971, the application shall also identify the design
objectives, and the means to be employed, for keeping levels of
radioactive material in effluents to unrestricted areas as low as is
reasonably achievable. The term ``as low as is reasonably achievable''
as used in this part means as low as is reasonably achievable taking
into account the state of technology, and the economics of improvements
in relation to benefits to the public health and safety and other
societal and socioeconomic considerations, and in relation to the use
of atomic energy in the public interest. The guides set out in appendix
I to this part provide numerical guidance on design objectives for
light-water-cooled nuclear power reactors to meet the requirements that
radioactive material in effluents released to unrestricted areas be
kept as low as is reasonably achievable. These numerical guides for
design objectives and limiting conditions for operation are not to be
construed as radiation protection standards.
(b) Each application for a construction permit shall include:
(1) A description of the preliminary design of equipment to be
installed under paragraph (a) of this section;
(2) An estimate of:
(i) The quantity of each of the principal radionuclides expected to
be released annually to unrestricted areas in liquid effluents produced
during normal reactor operations; and
(ii) The quantity of each of the principal radionuclides of the
gases, halides, and particulates expected to be released annually to
unrestricted areas in gaseous effluents produced during normal reactor
operations.
(3) A general description of the provisions for packaging, storage,
and shipment offsite of solid waste containing radioactive materials
resulting from treatment of gaseous and liquid effluents and from other
sources.
(c) Each application for an operating license shall include:
(1) A description of the equipment and procedures for the control
of gaseous and liquid effluents and for the maintenance and use of
equipment installed in radioactive waste systems, under paragraph (a)
of this section; and
(2) A revised estimate of the information required in paragraph
(b)(2) of this section if the expected releases and exposures differ
significantly from the estimates submitted in the application for a
construction permit.
(d) Each application for a combined license under part 52 of this
chapter shall include:
[[Page 49493]]
(1) A description of the equipment and procedures for the control
of gaseous and liquid effluents and for the maintenance and use of
equipment installed in radioactive waste systems, under paragraph (a)
of this section; and
(2) The information required in paragraph (b)(2) of this section.
(e) Each application for a design approval, a design certification,
or a manufacturing license under part 52 of this chapter shall include:
(1) A description of the equipment for the control of gaseous and
liquid effluents and for the maintenance and use of equipment installed
in radioactive waste systems, under paragraph (a) of this section; and
(2) The information required in paragraph (b)(2) of this section.
0
77. In Sec. 50.36, paragraphs (c), (d), and (e) are redesignated as
paragraphs (d), (e), and (f), respectively, and a new paragraph (c) is
added to read as follows:
Sec. 50.36 Technical specifications.
* * * * *
(c) Each applicant for a design certification or manufacturing
license under part 52 of this chapter shall include in its application
proposed generic technical specifications in accordance with the
requirements of this section for the portion of the plant that is
within the scope of the design certification or manufacturing license
application.
* * * * *
0
78. In Sec. 50.36a, paragraph (a) is revised to read as follows:
Sec. 50.36a Technical specifications on effluents from nuclear power
reactors.
(a) To keep releases of radioactive materials to unrestricted areas
during normal conditions, including expected occurrences, as low as is
reasonably achievable, each licensee of a nuclear power reactor and
each applicant for a design certification or a manufacturing license
will include technical specifications that, in addition to requiring
compliance with applicable provisions of Sec. 20.1301 of this chapter,
require that:
(1) Operating procedures developed pursuant to Sec. 50.34a(c) for
the control of effluents be established and followed and that the
radioactive waste system, pursuant to Sec. 50.34a, be maintained and
used. The licensee shall retain the operating procedures in effect as a
record until the Commission terminates the license and shall retain
each superseded revision of the procedures for 3 years from the date it
was superseded.
(2) Each holder of an operating license, and each holder of a
combined license after the Commission has made the finding under Sec.
52.103(g) of this chapter, shall submit a report to the Commission
annually that specifies the quantity of each of the principal
radionuclides released to unrestricted areas in liquid and in gaseous
effluents during the previous 12 months, including any other
information as may be required by the Commission to estimate maximum
potential annual radiation doses to the public resulting from effluent
releases. The report must be submitted as specified in Sec. 50.4, and
the time between submission of the reports must be no longer than 12
months. If quantities of radioactive materials released during the
reporting period are significantly above design objectives, the report
must cover this specifically. On the basis of these reports and any
additional information the Commission may obtain from the licensee or
others, the Commission may require the licensee to take action as the
Commission deems appropriate.
* * * * *
0
79. Section 50.36b is revised to read as follows:
Sec. 50.36b Environmental conditions.
(a) Each construction permit under this part, each early site
permit under part 52 of this chapter, and each combined license under
part 52 of this chapter may include conditions to protect the
environment during construction. These conditions are to be set out in
an attachment to the permit or license, which is incorporated in and
made a part of the permit or license. These conditions will be derived
from information contained in the environmental report submitted
pursuant to Sec. 51.50 of this chapter as analyzed and evaluated in
the NRC record of decision, and will identify the obligations of the
licensee in the environmental area, including, as appropriate,
requirements for reporting and keeping records of environmental data,
and any conditions and monitoring requirement for the protection of the
nonaquatic environment.
(b) Each license authorizing operation of a production or
utilization facility, including a combined license under part 52 of
this chapter, and each license for a nuclear power reactor facility for
which the certification of permanent cessation of operations required
under Sec. 50.82(a)(1) or Sec. 52.110(a) of this chapter has been
submitted, which is of a type described in Sec. 50.21(b)(2) or (3) or
Sec. 50.22 or is a testing facility, may include conditions to protect
the environment during operation and decommissioning. These conditions
are to be set out in an attachment to the license which is incorporated
in and made a part of the license. These conditions will be derived
from information contained in the environmental report or the
supplement to the environmental report submitted pursuant to Sec. Sec.
51.50 and 51.53 of this chapter as analyzed and evaluated in the NRC
record of decision, and will identify the obligations of the licensee
in the environmental area, including, as appropriate, requirements for
reporting and keeping records of environmental data, and any conditions
and monitoring requirement for the protection of the nonaquatic
environment.
0
80. Section 50.37 is revised to read as follows:
Sec. 50.37 Agreement limiting access to Classified Information.
As part of its application and in any event before the receipt of
Restricted Data or classified National Security Information or the
issuance of a license, construction permit, early site permit, or
standard design approval, or before the Commission has adopted a final
standard design certification rule under part 52 of this chapter, the
applicant shall agree in writing that it will not permit any individual
to have access to any facility to possess Restricted Data or classified
National Security Information until the individual and/or facility has
been approved for access under the provisions of 10 CFR parts 25 and/or
95. The agreement of the applicant becomes part of the license, or
construction permit, or standard design approval.
0
81. The undesignated center heading before Sec. 50.40 is revised to
read as follows:
Standards for Licenses, Certifications, and Regulatory Approvals
0
82. Section 50.40 is revised to read as follows:
Sec. 50.40 Common standards.
In determining that a construction permit or operating license in
this part, or early site permit, combined license, or manufacturing
license in part 52 of this chapter will be issued to an applicant, the
Commission will be guided by the following considerations:
(a) Except for an early site permit or manufacturing license, the
processes to be performed, the operating procedures, the facility and
equipment, the use of the facility, and other technical specifications,
or the proposals, in regard to any of the foregoing
[[Page 49494]]
collectively provide reasonable assurance that the applicant will
comply with the regulations in this chapter, including the regulations
in part 20 of this chapter, and that the health and safety of the
public will not be endangered.
(b) The applicant for a construction permit, operating license,
combined license, or manufacturing license is technically and
financially qualified to engage in the proposed activities in
accordance with the regulations in this chapter. However, no
consideration of financial qualification is necessary for an electric
utility applicant for an operating license for a utilization facility
of the type described in Sec. 50.21(b) or Sec. 50.22 or for an
applicant for a manufacturing license.
(c) The issuance of a construction permit, operating license, early
site permit, combined license, or manufacturing license to the
applicant will not, in the opinion of the Commission, be inimical to
the common defense and security or to the health and safety of the
public.
(d) Any applicable requirements of subpart A of 10 CFR part 51 have
been satisfied.
0
83. In Sec. 50.43, the section heading, the introductory paragraph,
and paragraph (d) are revised, and paragraph (e) is added to read as
follows:
Sec. 50.43 Additional standards and provisions affecting class 103
licenses and certifications for commercial power.
In addition to applying the standards set forth in Sec. Sec. 50.40
and 50.42, paragraphs (a) through (e) of this section apply in the case
of a class 103 license for a facility for the generation of commercial
power. For a design certification under part 52 of this chapter, only
paragraph (e) of this section applies.
* * * * *
(d) Nothing shall preclude any government agency, now or hereafter
authorized by law to engage in the production, marketing, or
distribution of electric energy, if otherwise qualified, from obtaining
a construction permit or operating license under this part, or a
combined license under part 52 of this chapter for a utilization
facility for the primary purpose of producing electric energy for
disposition for ultimate public consumption.
(e) Applications for a design certification, combined license,
manufacturing license, or operating license that propose nuclear
reactor designs which differ significantly from light-water reactor
designs that were licensed before 1997, or use simplified, inherent,
passive, or other innovative means to accomplish their safety
functions, will be approved only if:
(1)(i) The performance of each safety feature of the design has
been demonstrated through either analysis, appropriate test programs,
experience, or a combination thereof;
(ii) Interdependent effects among the safety features of the design
are acceptable, as demonstrated by analysis, appropriate test programs,
experience, or a combination thereof; and
(iii) Sufficient data exist on the safety features of the design to
assess the analytical tools used for safety analyses over a sufficient
range of normal operating conditions, transient conditions, and
specified accident sequences, including equilibrium core conditions; or
(2) There has been acceptable testing of a prototype plant over a
sufficient range of normal operating conditions, transient conditions,
and specified accident sequences, including equilibrium core
conditions. If a prototype plant is used to comply with the testing
requirements, then the NRC may impose additional requirements on
siting, safety features, or operational conditions for the prototype
plant to protect the public and the plant staff from the possible
consequences of accidents during the testing period.
0
84. Section 50.45 is revised to read as follows:
Sec. 50.45 Standards for construction permits, operating licenses,
and combined licenses.
(a) An applicant for an operating license or an amendment of an
operating license who proposes to construct or alter a production or
utilization facility will be initially granted a construction permit if
the application is in conformity with and acceptable under the criteria
of Sec. Sec. 50.31 through 50.38, and the standards of Sec. Sec.
50.40 through 50.43, as applicable.
(b) A holder of a combined license who proposes, after the
Commission makes the finding under Sec. 52.103(g) of this chapter, to
alter the licensed facility will be initially granted a construction
permit if the application is in conformity with and acceptable under
the criteria of Sec. Sec. 50.30 through 50.33, Sec. 50.34(f),
Sec. Sec. 50.34a through 50.38, the standards of Sec. Sec. 50.40
through 50.43, as applicable, and Sec. Sec. 52.79 and 52.80 of this
chapter.
0
85. In Sec. 50.46, paragraph (a)(3) is revised to read as follows:
Sec. 50.46 Acceptance criteria for emergency core cooling systems for
light-water nuclear power reactors.
(a) * * *
(3)(i) Each applicant for or holder of an operating license or
construction permit issued under this part, applicant for a standard
design certification under part 52 of this chapter (including an
applicant after the Commission has adopted a final design certification
regulation), or an applicant for or holder of a standard design
approval, a combined license or a manufacturing license issued under
part 52 of this chapter, shall estimate the effect of any change to or
error in an acceptable evaluation model or in the application of such a
model to determine if the change or error is significant. For this
purpose, a significant change or error is one which results in a
calculated peak fuel cladding temperature different by more than 50
[deg]F from the temperature calculated for the limiting transient using
the last acceptable model, or is a cumulation of changes and errors
such that the sum of the absolute magnitudes of the respective
temperature changes is greater than 50 [deg]F.
(ii) For each change to or error discovered in an acceptable
evaluation model or in the application of such a model that affects the
temperature calculation, the applicant or holder of a construction
permit, operating license, combined license, or manufacturing license
shall report the nature of the change or error and its estimated effect
on the limiting ECCS analysis to the Commission at least annually as
specified in Sec. 50.4 or Sec. 52.3 of this chapter, as applicable.
If the change or error is significant, the applicant or licensee shall
provide this report within 30 days and include with the report a
proposed schedule for providing a reanalysis or taking other action as
may be needed to show compliance with Sec. 50.46 requirements. This
schedule may be developed using an integrated scheduling system
previously approved for the facility by the NRC. For those facilities
not using an NRC approved integrated scheduling system, a schedule will
be established by the NRC staff within 60 days of receipt of the
proposed schedule. Any change or error correction that results in a
calculated ECCS performance that does not conform to the criteria set
forth in paragraph (b) of this section is a reportable event as
described in Sec. Sec. 50.55(e), 50.72, and 50.73. The affected
applicant or licensee shall propose immediate steps to demonstrate
compliance or bring plant design or operation into compliance with
Sec. 50.46 requirements.
(iii) For each change to or error discovered in an acceptable
evaluation model or in the application of such a model that affects the
temperature
[[Page 49495]]
calculation, the applicant or holder of a standard design approval or
the applicant for a standard design certification (including an
applicant after the Commission has adopted a final design certification
rule) shall report the nature of the change or error and its estimated
effect on the limiting ECCS analysis to the Commission and to any
applicant or licensee referencing the design approval or design
certification at least annually as specified in Sec. 52.3 of this
chapter. If the change or error is significant, the applicant or holder
of the design approval or the applicant for the design certification
shall provide this report within 30 days and include with the report a
proposed schedule for providing a reanalysis or taking other action as
may be needed to show compliance with Sec. 50.46 requirements. The
affected applicant or holder shall propose immediate steps to
demonstrate compliance or bring plant design into compliance with Sec.
50.46 requirements.
* * * * *
0
86. In Sec. 50.47, paragraph (a)(1) is revised and paragraph (e) is
added to read as follows:
Sec. 50.47 Emergency plans.
(a)(1)(i) Except as provided in paragraph (d) of this section, no
initial operating license for a nuclear power reactor will be issued
unless a finding is made by the NRC that there is reasonable assurance
that adequate protective measures can and will be taken in the event of
a radiological emergency. No finding under this section is necessary
for issuance of a renewed nuclear power reactor operating license.
(ii) No initial combined license under part 52 of this chapter will
be issued unless a finding is made by the NRC that there is reasonable
assurance that adequate protective measures can and will be taken in
the event of a radiological emergency. No finding under this section is
necessary for issuance of a renewed combined license.
(iii) If an application for an early site permit under subpart A of
part 52 of this chapter includes complete and integrated emergency
plans under 10 CFR 52.17(b)(2)(ii), no early site permit will be issued
unless a finding is made by the NRC that the emergency plans provide
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency.
(iv) If an application for an early site permit proposes major
features of the emergency plans under 10 CFR 52.17(b)(2)(i), no early
site permit will be issued unless a finding is made by the NRC that the
major features are acceptable in accordance with the applicable
standards of 10 CFR 50.47 and 10 CFR part 50, appendix E, within the
scope of emergency preparedness matters addressed in the major
features.
* * * * *
(e) Notwithstanding the requirements of paragraph (b) of this
section and the provisions of Sec. 52.103 of this chapter, a holder of
a combined license under part 52 of this chapter may not load fuel or
operate except as provided in accordance with appendix E to part 50 and
Sec. 50.54(gg).
0
87. In Sec. 50.48, the introductory text of paragraph (a)(1) is
revised and paragraph (a)(4) is added to read as follows:
Sec. 50.48 Fire protection.
(a)(1) Each holder of an operating license issued under this part
or a combined license issued under part 52 of this chapter must have a
fire protection plan that satisfies Criterion 3 of appendix A to this
part. This fire protection plan must:
* * * * *
(a)(4) Each applicant for a design approval, design certification,
or manufacturing license under part 52 of this chapter must have a
description and analysis of the fire protection design features for the
standard plant necessary to demonstrate compliance with Criterion 3 of
appendix A to this part.
* * * * *
0
88. In Sec. 50.49, paragraph (a) is revised to read as follows:
Sec. 50.49 Environmental qualification of electric equipment
important to safety for nuclear power plants.
(a) Each holder of or an applicant for an operating license issued
under this part, or a combined license or manufacturing license issued
under part 52 of this chapter, other than a nuclear power plant for
which the certifications required under Sec. 50.82(a)(1) or Sec.
52.110(a)(1) of this chapter have been submitted, shall establish a
program for qualifying the electric equipment defined in paragraph (b)
of this section. For a manufacturing license, only electric equipment
defined in paragraph (b) which is within the scope of the manufactured
reactor must be included in the program.
* * * * *
0
89. In Sec. 50.54, the introductory text, and paragraphs (a)(1), (i-
1), (o), (p), and (q) are revised and paragraph (gg) is added to read
as follows:
Sec. 50.54 Conditions of licenses.
The following paragraphs with the exception of paragraphs (r) and
(gg) of this section are conditions in every nuclear power reactor
operating license issued under this part. The following paragraphs with
the exception of paragraph (r), (s), and (u) of this section are
conditions in every combined license issued under part 52 of this
chapter, provided, however, that paragraphs (i), (i-1), (j), (k), (l),
(m), (n), (w), (x), (y), and (z) of this section are only applicable
after the Commission makes the finding under Sec. 52.103(g) of this
chapter.
(a)(1) Each nuclear power plant or fuel reprocessing plant licensee
subject to the quality assurance criteria in appendix B of this part
shall implement, under Sec. 50.34(b)(6)(ii) or Sec. 52.79 of this
chapter, the quality assurance program described or referenced in the
safety analysis report, including changes to that report. However, a
holder of a combined license under part 52 of this chapter shall
implement the quality assurance program described or referenced in the
safety analysis report applicable to operation 30 days prior to the
scheduled date for the initial loading of fuel.
* * * * *
(i-1) Within 3 months after either the issuance of an operating
license or the date that the Commission makes the finding under Sec.
52.103(g) of this chapter for a combined license, as applicable, the
licensee shall have in effect an operator requalification program. The
operator requalification program must, as a minimum, meet the
requirements of Sec. 55.59(c) of this chapter. Notwithstanding the
provisions of Sec. 50.59, the licensee may not, except as specifically
authorized by the Commission decrease the scope of an approved operator
requalification program.
* * * * *
(o) Primary reactor containments for water cooled power reactors,
other than facilities for which the certifications required under
Sec. Sec. 50.82(a)(1) or 52.110(a)(1) of this chapter have been
submitted, shall be subject to the requirements set forth in appendix J
to this part.
(p)(1) The licensee shall prepare and maintain safeguards
contingency plan procedures in accordance with appendix C of part 73 of
this chapter for effecting the actions and decisions contained in the
Responsibility Matrix of the safeguards contingency plan. The licensee
may make no change which would decrease the effectiveness of a security
plan, or guard training and
[[Page 49496]]
qualification plan, prepared pursuant to Sec. 50.34(c) or Sec.
52.79(a), or part 73 of this chapter, or of the first four categories
of information (Background, Generic Planning Base, Licensee Planning
Base, Responsibility Matrix) contained in a licensee safeguards
contingency plan prepared pursuant to Sec. 50.34(d) or Sec. 52.79(a)
or part 73 of this chapter, as applicable, without prior approval of
the Commission. A licensee desiring to make such a change shall submit
an application for an amendment to the licensee's license pursuant to
Sec. 50.90.
(2) The licensee may make changes to the plans referenced in
paragraph (p)(1) of this section, without prior Commission approval if
the changes do not decrease the safeguards effectiveness of the plan.
The licensee shall maintain records of changes to the plans made
without prior Commission approval for a period of 3 years from the date
of the change, and shall submit, as specified in Sec. 50.4 or Sec.
52.3 of this chapter, a report containing a description of each change
within 2 months after the change is made. Prior to the safeguards
contingency plan being put into effect, the licensee shall have:
(i) All safeguards capabilities specified in the safeguards
contingency plan available and functional;
(ii) Detailed procedures developed according to appendix C to part
73 of this chapter available at the licensee's site; and
(iii) All appropriate personnel trained to respond to safeguards
incidents as outlined in the plan and specified in the detailed
procedures.
(3) The licensee shall provide for the development, revision,
implementation, and maintenance of its safeguards contingency plan. The
licensee shall ensure that all program elements are reviewed by
individuals independent of both security program management and
personnel who have direct responsibility for implementation of the
security program either:
(i) At intervals not to exceed 12 months; or
(ii) As necessary, based on an assessment by the licensee against
performance indicators, and as soon as reasonably practicable after a
change occurs in personnel, procedures, equipment, or facilities that
potentially could adversely affect security, but no longer than 12
months after the change. In any case, all elements of the safeguards
contingency plan must be reviewed at least once every 24 months.
(4) The review must include a review and audit of safeguards
contingency procedures and practices, an audit of the security system
testing and maintenance program, and a test of the safeguards systems
along with commitments established for response by local law
enforcement authorities. The results of the review and audit, along
with recommendations for improvements, must be documented, reported to
the licensee's corporate and plant management, and kept available at
the plant for inspection for a period of 3 years.
(q) A holder of a nuclear power reactor operating license under
this part, or a combined license under part 52 of this chapter after
the Commission makes the finding under Sec. 52.103(g) of this chapter,
shall follow and maintain in effect emergency plans which meet the
standards in Sec. 50.47(b) and the requirements in appendix E of this
part. A licensee authorized to possess and/or operate a research
reactor or a fuel facility shall follow and maintain in effect
emergency plans which meet the requirements in appendix E to this part.
The licensee shall retain the emergency plan and each change that
decreases the effectiveness of the plan as a record until the
Commission terminates the license for the nuclear power reactor. The
nuclear power reactor licensee may make changes to these plans without
Commission approval only if the changes do not decrease the
effectiveness of the plans and the plans, as changed, continue to meet
the standards of Sec. 50.47(b) and the requirements of appendix E to
this part. The research reactor and/or the fuel facility licensee may
make changes to these plans without Commission approval only if these
changes do not decrease the effectiveness of the plans and the plans,
as changed, continue to meet the requirements of appendix E to this
part. This nuclear power reactor, research reactor, or fuel facility
licensee shall retain a record of each change to the emergency plan
made without prior Commission approval for a period of 3 years from the
date of the change. Proposed changes that decrease the effectiveness of
the approved emergency plans may not be implemented without application
to and approval by the Commission. The licensee shall submit, as
specified in Sec. 50.4, a report of each proposed change for approval.
If a change is made without approval, the licensee shall submit, as
specified in Sec. 50.4, a report of each change within 30 days after
the change is made.
* * * * *
(gg)(1) Notwithstanding 10 CFR 52.103, if following the conduct of
the exercise required by paragraph IV.f.2.a of appendix E to part 50 of
this chapter, DHS identifies one or more deficiencies in the state of
offsite emergency preparedness, the holder of a combined license under
10 CFR part 52 may operate at up to 5 percent of rated thermal power
only if the Commission finds that the state of onsite emergency
preparedness provides reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological
emergency. The NRC will base this finding on its assessment of the
applicant's onsite emergency plans against the pertinent standards in
Sec. 50.47 and appendix E to this part. Review of the applicant's
emergency plans will include the following standards with offsite
aspects:
(i) Arrangements for requesting and effectively using offsite
assistance onsite have been made, arrangements to accommodate State and
local staff at the licensee's near-site Emergency Operations Facility
have been made, and other organizations capable of augmenting the
planned onsite response have been identified.
(ii) Procedures have been established for licensee communications
with State and local response organizations, including initial
notification of the declaration of emergency and periodic provision of
plant and response status reports.
(iii) Provisions exist for prompt communications among principal
response organizations to offsite emergency personnel who would be
responding onsite.
(iv) Adequate emergency facilities and equipment to support the
emergency response onsite are provided and maintained.
(v) Adequate methods, systems, and equipment for assessing and
monitoring actual or potential offsite consequences of a radiological
emergency condition are in use onsite.
(vi) Arrangements are made for medical services for contaminated
and injured onsite individuals.
(vii) Radiological emergency response training has been made
available to those offsite who may be called to assist in an emergency
onsite.
(2) The condition in this paragraph, regarding operation at up to 5
percent power, ceases to apply 30 days after DHS informs the NRC that
the offsite deficiencies have been corrected, unless the NRC notifies
the combined license holder before the expiration of the 30-day period
that the Commission finds under paragraphs (s)(2) and (3) of this
section that the state of emergency preparedness does not provide
reasonable assurance that adequate protective measures can and will be
taken in the event of a radiological emergency.
[[Page 49497]]
0
90. In Sec. 50.55, the heading, the introductory text and paragraphs
(a), (b), and (e) are revised, and a new paragraph (f)(4) is added to
read as follows:
Sec. 50.55 Conditions of construction permits, early site permits,
combined licenses, and manufacturing licenses.
Each construction permit is subject to the following terms and
conditions; each early site permit is subject to the terms and
conditions in paragraph (f) of this section; each manufacturing license
is subject to the terms and conditions in paragraphs (e) and (f) of
this section; and each combined license is subject to the terms and
conditions in paragraphs (e) and (f) of this section until the date
that the Commission makes the finding under Sec. 52.103(g) of this
chapter:
(a) The construction permit shall state the earliest and latest
dates for completion of the construction or modification.
(b) If the proposed construction or modification of the facility is
not completed by the latest completion date, the construction permit
shall expire and all rights are forfeited. However, upon good cause
shown, the Commission will extend the completion date for a reasonable
period of time. The Commission will recognize, among other things,
developmental problems attributable to the experimental nature of the
facility or fire, flood, explosion, strike, sabotage, domestic
violence, enemy action, an act of the elements, and other acts beyond
the control of the permit holder, as a basis for extending the
completion date.
* * * * *
(e)(1) Definitions. For purposes of this paragraph, the definitions
in Sec. 21.3 of this chapter apply.
(2) Posting requirements. (i) Each individual, partnership,
corporation, dedicating entity, or other entity subject to the
regulations in this part shall post current copies of the regulations
in this part; Section 206 of the Energy Reorganization Act of 1974
(ERA); and procedures adopted under the regulations in this part. These
documents must be posted in a conspicuous position on any premises
within the United States where the activities subject to this part are
conducted.
(ii) If posting of the regulations in this part or the procedures
adopted under the regulations in this part is not practicable, the
licensee or firm subject to the regulations in this part may, in
addition to posting Section 206 of the ERA, post a notice which
describes the regulations/procedures, including the name of the
individual to whom reports may be made, and states where the
regulation, procedures, and reports may be examined.
(3) Procedures. Each individual, corporation, partnership, or other
entity holding a facility construction permit subject to this part,
combined license (until the Commission makes the finding under 10 CFR
52.103(g)), and manufacturing license under 10 CFR part 52 must adopt
appropriate procedures to--
(i) Evaluate deviations and failures to comply to identify defects
and failures to comply associated with substantial safety hazards as
soon as practicable, and, except as provided in paragraph (e)(3)(ii) of
this section, in all cases within 60 days of discovery, to identify a
reportable defect or failure to comply that could create a substantial
safety hazard, were it to remain uncorrected.
(ii) Ensure that if an evaluation of an identified deviation or
failure to comply potentially associated with a substantial safety
hazard cannot be completed within 60 days from discovery of the
deviation or failure to comply, an interim report is prepared and
submitted to the Commission through a director or responsible officer
or designated person as discussed in paragraph (e)(4)(v) of this
section. The interim report should describe the deviation or failure to
comply that is being evaluated and should also state when the
evaluation will be completed. This interim report must be submitted in
writing within 60 days of discovery of the deviation or failure to
comply.
(iii) Ensure that a director or responsible officer of the holder
of a facility construction permit subject to this part, combined
license (until the Commission makes the finding under 10 CFR
52.103(g)), and manufacturing license under 10 CFR part 52 is informed
as soon as practicable, and, in all cases, within the 5 working days
after completion of the evaluation described in paragraph (e)(3)(i) or
(e)(3)(ii) of this section, if the construction or manufacture of a
facility or activity, or a basic component supplied for such facility
or activity--
(A) Fails to comply with the AEA, as amended, or any applicable
regulation, order, or license of the Commission, relating to a
substantial safety hazard;
(B) Contains a defect; or
(C) Undergoes any significant breakdown in any portion of the
quality assurance program conducted under the requirements of appendix
B to 10 CFR part 50 which could have produced a defect in a basic
component. These breakdowns in the quality assurance program are
reportable whether or not the breakdown actually resulted in a defect
in a design approved and released for construction, installation, or
manufacture.
(4) Notification. (i) The holder of a facility construction permit
subject to this part, combined license (until the Commission makes the
finding under 10 CFR 52.103(g)), and manufacturing license who obtains
information reasonably indicating that the facility fails to comply
with the AEA, as amended, or any applicable regulation, order, or
license of the Commission relating to a substantial safety hazard must
notify the Commission of the failure to comply through a director or
responsible officer or designated person as discussed in paragraph
(e)(10) of this section.
(ii) The holder of a facility construction permit subject to this
part, combined license, or manufacturing license, who obtains
information reasonably indicating the existence of any defect found in
the construction or manufacture, or any defect found in the final
design of a facility as approved and released for construction or
manufacture, must notify the Commission of the defect through a
director or responsible officer or designated person as discussed in
paragraph (e)(4)(v) of this section.
(iii) The holder of a facility construction permit subject to this
part, combined license, or manufacturing license, who obtains
information reasonably indicating that the quality assurance program
has undergone any significant breakdown discussed in paragraph
(e)(3)(ii)(C) of this section must notify the Commission of the
breakdown in the quality assurance program through a director or
responsible officer or designated person as discussed in paragraph
(4)(v) of this section.
(iv) A dedicating entity is responsible for identifying and
evaluating deviations and reporting defects and failures to comply
associated with substantial safety hazards for dedicated items; and
maintaining auditable records for the dedication process.
(v) The notification requirements of this paragraph apply to all
defects and failures to comply associated with a substantial safety
hazard regardless of whether extensive evaluation, redesign, or repair
is required to conform to the criteria and bases stated in the safety
analysis report, construction permit, combined license, or
manufacturing license. Evaluation of potential defects and failures to
comply and reporting of defects and failures to comply under this
section satisfies the construction permit holder's, combined license
holder's, and manufacturing license holder's evaluation and
notification
[[Page 49498]]
obligations under part 21 of this chapter, and satisfies the
responsibility of individual directors or responsible officers of
holders of construction permits issued under Sec. 50.23, holders of
combined licenses (until the Commission makes the finding under Sec.
52.103 of this chapter), and holders of manufacturing licenses to
report defects, and failures to comply associated with substantial
safety hazards under Section 206 of the ERA. The director or
responsible officer may authorize an individual to provide the
notification required by this section, provided that this must not
relieve the director or responsible officer of his or her
responsibility under this section.
(5) Notification--timing and where sent. The notification required
by paragraph (e)(4) of this section must consist of--
(i) Initial notification by facsimile, which is the preferred
method of notification, to the NRC Operations Center at (301) 816-5151
or by telephone at (301) 816-5100 within 2 days following receipt of
information by the director or responsible corporate officer under
paragraph (e)(3)(iii) of this section, on the identification of a
defect or a failure to comply. Verification that the facsimile has been
received should be made by calling the NRC Operations Center. This
paragraph does not apply to interim reports described in paragraph
(e)(3)(ii) of this section.
(ii) Written notification submitted to the Document Control Desk,
U.S. Nuclear Regulatory Commission, by an appropriate method listed in
Sec. 50.4, with a copy to the appropriate Regional Administrator at
the address specified in appendix D to part 20 of this chapter and a
copy to the appropriate NRC resident inspector within 30 days following
receipt of information by the director or responsible corporate officer
under paragraph (e)(3)(iii) of this section, on the identification of a
defect or failure to comply.
(6) Content of notification. The written notification required by
paragraph (e)(9)(ii) of this section must clearly indicate that the
written notification is being submitted under Sec. 50.55(e) and
include the following information, to the extent known.
(i) Name and address of the individual or individuals informing the
Commission.
(ii) Identification of the facility, the activity, or the basic
component supplied for the facility or the activity within the United
States which contains a defect or fails to comply.
(iii) Identification of the firm constructing or manufacturing the
facility or supplying the basic component which fails to comply or
contains a defect.
(iv) Nature of the defect or failure to comply and the safety
hazard which is created or could be created by the defect or failure to
comply.
(v) The date on which the information of a defect or failure to
comply was obtained.
(vi) In the case of a basic component which contains a defect or
fails to comply, the number and location of all the basic components in
use at the facility subject to the regulations in this part.
(vii) In the case of a completed reactor manufactured under part 52
of this chapter, the entities to which the reactor was supplied.
(viii) The corrective action which has been, is being, or will be
taken; the name of the individual or organization responsible for the
action; and the length of time that has been or will be taken to
complete the action.
(ix) Any advice related to the defect or failure to comply about
the facility, activity, or basic component that has been, is being, or
will be given to other entities.
(7) Procurement documents. Each individual, corporation,
partnership, dedicating entity, or other entity subject to the
regulations in this part shall ensure that each procurement document
for a facility, or a basic component specifies or is issued by the
entity subject to the regulations, when applicable, that the provisions
of 10 CFR part 21 or 10 CFR 50.55(e) applies, as applicable.
(8) Coordination with 10 CFR part 21. The requirements of Sec.
50.55(e) are satisfied when the defect or failure to comply associated
with a substantial safety hazard has been previously reported under
part 21 of this chapter, under Sec. 73.71 of this chapter, or under
Sec. Sec. 50.55(e) or 50.73. For holders of construction permits
issued before October 29, 1991, evaluation, reporting and recordkeeping
requirements of Sec. 50.55(e) may be met by complying with the
comparable requirements of part 21 of this chapter.
(9) Records retention. The holder of a construction permit,
combined license, and manufacturing license must prepare and maintain
records necessary to accomplish the purposes of this section,
specifically--
(i) Retain procurement documents, which define the requirements
that facilities or basic components must meet in order to be considered
acceptable, for the lifetime of the facility or basic component.
(ii) Retain records of evaluations of all deviations and failures
to comply under paragraph (e)(3)(i) of this section for the longest of:
(A) Ten (10) years from the date of the evaluation;
(B) Five (5) years from the date that an early site permit is
referenced in an application for a combined license; or
(C) Five (5) years from the date of delivery of a manufactured
reactor.
(iii) Retain records of all interim reports to the Commission made
under paragraph (e)(3)(ii) of this section, or notifications to the
Commission made under paragraph (e)(4) of this section for the minimum
time periods stated in paragraph (e)(9)(ii) of this section;
(iv) Suppliers of basic components must retain records of:
(A) All notifications sent to affected licensees or purchasers
under paragraph (e)(4)(iv) of this section for a minimum of ten (10)
years following the date of the notification;
(B) The facilities or other purchasers to whom basic components or
associated services were supplied for a minimum of fifteen (15) years
from the delivery of the basic component or associated services.
(v) Maintaining records in accordance with this section satisfies
the recordkeeping obligations under part 21 of this chapter of the
entities, including directors or responsible officers thereof, subject
to this section.
(f) * * *
(4) Each holder of an early site permit or a manufacturing license
under part 52 of this chapter shall implement the quality assurance
program described or referenced in the safety analysis report,
including changes to that report. Each holder of a combined license
shall implement the quality assurance program for design and
construction described or referenced in the safety analysis report,
including changes to that report, provided, however, that the holder of
a combined license is not subject to the terms and conditions in this
paragraph after the Commission makes the finding under Sec. 52.103(g)
of this chapter.
(i) Each holder described in paragraph (f)(4) of this section may
make a change to a previously accepted quality assurance program
description included or referenced in the safety analysis report, if
the change does not reduce the commitments in the program description
previously accepted by the NRC. Changes to the quality assurance
program description that do not reduce the commitments must be
submitted to NRC within 90 days. Changes to the quality assurance
program description that reduce the commitments must be submitted to
NRC and receive NRC
[[Page 49499]]
approval before implementation, as follows:
(A) Changes to the safety analysis report must be submitted for
review as specified in Sec. 50.4. Changes made to NRC-accepted quality
assurance topical report descriptions must be submitted as specified in
Sec. 50.4.
(B) The submittal of a change to the safety analysis report quality
assurance program description must include all pages affected by that
change and must be accompanied by a forwarding letter identifying the
change, the reason for the change, and the basis for concluding that
the revised program incorporating the change continues to satisfy the
criteria of appendix B of this part and the safety analysis report
quality assurance program description commitments previously accepted
by the NRC (the letter need not provide the basis for changes that
correct spelling, punctuation, or editorial items).
(C) A copy of the forwarding letter identifying the changes must be
maintained as a facility record for three (3) years.
(D) Changes to the quality assurance program description included
or referenced in the safety analysis report shall be regarded as
accepted by the Commission upon receipt of a letter to this effect from
the appropriate reviewing office of the Commission or 60 days after
submittal to the Commission, whichever occurs first.
(ii) [Reserved]
0
91. In Section 50.55a, the introductory paragraphs (b)(1)(i),
(b)(1)(ii), (b)(1)(iii), (b)(1)(v), the introductory text of paragraphs
(b)(4) and (d)(1), paragraph (e)(1), the introductory text of paragraph
(f)(3), paragraphs (f)(3)(iii), (f)(3)(iv)(B), (f)(4)(i), the
introductory text of paragraph (g)(3), paragraphs (g)(4)(i), the
introductory text of paragraph (g)(4)(v), and paragraph (h)(3) are
revised to read as follows:
Sec. 50.55a Codes and standards.
Each construction permit for a utilization facility is subject to
the following conditions in addition to those specified in Sec. 50.55.
Each combined license for a utilization facility is subject to the
following conditions in addition to those specified in Sec. 50.55,
except that each combined license for a boiling or pressurized water-
cooled nuclear power facility is subject to the conditions in
paragraphs (f) and (g) of this section, but only after the Commission
makes the finding under Sec. 52.103(g) of this chapter. Each operating
license for a boiling or pressurized water-cooled nuclear power
facility is subject to the conditions in paragraphs (f) and (g) of this
section in addition to those specified in Sec. 50.55. Each
manufacturing license, standard design approval, and standard design
certification application under part 52 of this chapter is subject to
the conditions in paragraphs (a), (b)(1), (b)(4), (c), (d), (e),
(f)(3), and (g)(3) of this section.
* * * * *
(b) * * *
(1) * * *
(i) Section III Materials. When applying the 1992 Edition of
Section III, applicants or licensees must apply the 1992 Edition with
the 1992 Addenda of Section II of the ASME Boiler and Pressure Vessel
Code.
(ii) Weld leg dimensions. When applying the 1989 Addenda through
the latest edition, and addenda incorporated by reference in paragraph
(b)(1) of this section, applicants or licensees may not apply paragraph
NB-3683.4(c)(1), Footnote 11 to Figure NC-3673.2(b)-1, and Figure ND-
3673.2(b)-1.
(iii) Seismic design. Applicants or licensees may use Articles NB-
3200, NB-3600, NC-3600, and ND-3600 up to and including the 1993
Addenda, subject to the limitation specified in paragraph (b)(1)(ii) of
this section. Applicants or licensees may not use these articles in the
1994 Addenda through the latest edition and addenda incorporated by
reference in paragraph (b)(1) of this section.
* * * * *
(v) Independence of inspection. Applicants or licensees may not
apply NCA-4134.10(a) of Section III, 1995 Edition, through the latest
edition and addenda incorporated by reference in paragraph (b)(1) of
this section.
* * * * *
(4) Design, Fabrication, and Materials Code Cases. Applicants or
licensees may apply the ASME Boiler and Pressure Vessel Code cases
listed in NRC Regulatory Guide 1.84, Revision 33, without prior NRC
approval subject to the following:
* * * * *
(d) * * *
(1) For a nuclear power plant whose application for a construction
permit under this part, or a combined license or manufacturing license
under part 52 of this chapter is docketed after May 14, 1984, or for an
application for a standard design approval or a standard design
certification docketed after May 14, 1984, components classified
Quality Group B \9\ must meet the requirements for Class 2 Components
in Section III of the ASME Boiler and Pressure Vessel Code.
---------------------------------------------------------------------------
\9\ See footnotes at end of section.
---------------------------------------------------------------------------
* * * * *
(e) * * *
(1) For a nuclear power plant whose application for a construction
permit under this part, or a combined license or manufacturing license
under part 52 of this chapter is docketed after May 14, 1984, or for an
application for a standard design approval or a standard design
certification docketed after May 14, 1984, components classified
Quality Group C 9 must meet the requirements for Class 3
components in Section III of the ASME Boiler and Pressure Vessel Code.
* * * * *
(f) * * *
(3) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit under this part or design approval,
design certification, combined license, or manufacturing license under
part 52 of this chapter, was issued on or after July 1, 1974:
* * * * *
(iii)(A) Pumps and valves, in facilities whose construction permit
under this part, or design certification or design approval under part
52 of this chapter was issued before November 22, 1999, which are
classified as ASME Code Class 1 must be designed and be provided with
access to enable the performance of inservice testing of the pumps and
valves for assessing operational readiness set forth in the editions
and addenda of Section XI of the ASME Boiler and Pressure Vessel Code
incorporated by reference in paragraph (b) of this section (or the
optional ASME Code cases that are listed in NRC Regulatory Guide 1.147,
through Revision 14 or Regulatory Guide 1.192, that are incorporated by
reference in paragraph (b) of this section) applied to the construction
of the particular pump or valve or the summer 1973 Addenda, whichever
is later.
(B) Pumps and valves, in facilities whose construction permit under
this part, or design certification, design approval, combined license,
or manufacturing license under part 52 of this chapter, is issued on or
after November 22, 1999, which are classified as ASME Code Class 1 must
be designed and be provided with access to enable the performance of
inservice testing of the pumps and valves for assessing operational
readiness set forth in editions and addenda of the ASME OM Code (or the
optional ASME Code cases listed in the NRC Regulatory Guide 1.192 that
is incorporated by reference in paragraph (b) of this section)
[[Page 49500]]
referenced in paragraph (b)(3) of this section at the time the
construction permit, combined license, manufacturing license, design
certification, or design approval is issued.
(iv) * * *
(B) Pumps and valves, in facilities whose construction permit under
this part or design certification or combined license under part 52 of
this chapter is issued on or after November 22, 1999, which are
classified as ASME Code Class 2 and 3 must be designed and be provided
with access to enable the performance of inservice testing of the pumps
and valves for assessing operational readiness set forth in editions
and addenda of the ASME OM Code (or the optional ASME Code cases listed
in the NRC Regulatory Guide 1.192 that is incorporated by reference in
paragraph (b) of this section) referenced in paragraph (b)(3) of this
section at the time the construction permit, combined license, or
design certification is issued.
* * * * *
(4) * * *
(i) Inservice tests to verify operational readiness of pumps and
valves, whose function is required for safety, conducted during the
initial 120-month interval must comply with the requirements in the
latest edition and addenda of the Code incorporated by reference in
paragraph (b) of this section on the date 12 months before the date of
issuance of the operating license under this part, or 12 months before
the date scheduled for initial loading fuel under a combined license
under part 52 of this chapter (or the optional ASME Code cases listed
in NRC Regulatory Guide 1.192, that is incorporated by reference in
paragraph (b) of this section), subject to the limitations and
modifications listed in paragraph (b) of this section.
* * * * *
(g) * * *
(3) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit under this part, or design
certification, design approval, combined license, or manufacturing
license under part 52 of this chapter, was issued on or after July 1,
1974:
* * * * *
(4) * * *
(i) Inservice examinations of components and system pressure tests
conducted during the initial 120-month inspection interval must comply
with the requirements in the latest edition and addenda of the Code
incorporated by reference in paragraph (b) of this section on the date
12 months before the date of issuance of the operating license under
this part, or 12 months before the date scheduled for initial loading
of fuel under a combined license under part 52 of this chapter (or the
optional ASME Code cases listed in NRC Regulatory Guide 1.147, through
Revision 14, that are incorporated by reference in paragraph (b) of
this section), subject to the limitations and modifications listed in
paragraph (b) of this section.
* * * * *
(v) For a boiling or pressurized water-cooled nuclear power
facility whose construction permit under this part or combined license
under part 52 of this chapter was issued after January 1, 1956:
* * * * *
(h) * * *
(3) Safety systems. Applications filed on or after May 13, 1999,
for construction permits and operating licenses under this part, and
for design approvals, design certifications, and combined licenses
under part 52 of this chapter, must meet the requirements for safety
systems in IEEE Std. 603-1991 and the correction sheet dated January
30, 1995.
0
92. In Sec. 50.59, paragraphs (b), (d)(2), and (d)(3) are revised to
read as follows:
Sec. 50.59 Changes, tests, and experiments.
* * * * *
(b) This section applies to each holder of an operating license
issued under this part or a combined license issued under part 52 of
this chapter, including the holder of a license authorizing operation
of a nuclear power reactor that has submitted the certification of
permanent cessation of operations required under Sec. 50.82(a)(1) or
Sec. 50.110 or a reactor licensee whose license has been amended to
allow possession of nuclear fuel but not operation of the facility.
* * * * *
(d) * * *
(2) The licensee shall submit, as specified in Sec. 50.4 or Sec.
52.3 of this chapter, as applicable, a report containing a brief
description of any changes, tests, and experiments, including a summary
of the evaluation of each. A report must be submitted at intervals not
to exceed 24 months. For combined licenses, the report must be
submitted at intervals not to exceed 6 months during the period from
the date of application for a combined license to the date the
Commission makes its findings under 10 CFR 52.103(g).
(3) The records of changes in the facility must be maintained until
the termination of an operating license issued under this part, a
combined license issued under part 52 of this chapter, or the
termination of a license issued under 10 CFR part 54, whichever is
later. Records of changes in procedures and records of tests and
experiments must be maintained for a period of 5 years.
0
93. In Sec. 50.61, paragraph (b)(1) is revised to read as follows:
Sec. 50.61 Fracture toughness requirements for protection against
pressurized thermal shock events.
* * * * *
(b) * * *
(1) For each pressurized water nuclear power reactor for which an
operating license has been issued under this part or a combined license
has been issued under part 52 of this chapter, other than a nuclear
power reactor facility for which the certifications required under
Sec. 50.82(a)(1) have been submitted, the licensee shall have
projected values of RTPTS, accepted by the NRC, for each
reactor vessel beltline material for the EOL fluence of the material.
The assessment of RTPTS must use the calculation procedures
given in paragraph (c)(1) of this section, except as provided in
paragraphs (c)(2) and (c)(3) of this section. The assessment must
specify the bases for the projected value of RTPTS for each
vessel beltline material, including the assumptions regarding core
loading patterns, and must specify the copper and nickel contents and
the fluence value used in the calculation for each beltline material.
This assessment must be updated whenever there is a significant \2\
change in projected values of RTPTS, or upon request for a
change in the expiration date for operation of the facility.
---------------------------------------------------------------------------
\2\ Changes to RTPTS values are considered
significant if either the previous value or the current value, or
both values, exceed the screening criterion before the expiration of
the operating license or the combined license under part 52 of this
chapter, including any renewed term, if applicable for the plant.
---------------------------------------------------------------------------
* * * * *
0
94. In Sec. 50.62, paragraph (d) is revised to read as follows:
Sec. 50.62 Requirements for reduction of risk from anticipated
transients without scram (ATWS) events for light-water-cooled nuclear
power plants.
* * * * *
(d) Implementation. For each light-water-cooled nuclear power plant
operating license issued before September 27, 2007, by 180 days after
the issuance of the QA guidance for non-safety related components, each
licensee shall develop and submit to the Commission, as specified in
Sec. 50.4, a proposed schedule for meeting the
[[Page 49501]]
requirements of paragraphs (c)(1) through (c)(5) of this section. Each
shall include an explanation of the schedule along with a justification
if the schedule calls for final implementation later than the second
refueling outage after July 26, 1984, or the date of issuance of a
license authorizing operation above 5 percent of full power. A final
schedule shall then be mutually agreed upon by the Commission and
licensee. For each light-water-cooled nuclear power plant operating
license application submitted after September 27, 2007, the applicant
shall submit information in its final safety analysis report
demonstrating how it will comply with paragraphs (c)(1) through (c)(5)
of this section.
0
95. In Sec. 50.63, the introductory text of paragraphs (a)(1) and
(c)(1) are revised to read as follows:
Sec. 50.63 Loss of all alternating current power.
(a) * * *
(1) Each light-water-cooled nuclear power plant licensed to operate
under this part, each light-water-cooled nuclear power plant licensed
under subpart C of 10 CFR part 52 after the Commission makes the
finding under Sec. 52.103(g) of this chapter, and each design for a
light-water-cooled nuclear power plant approved under a standard design
approval, standard design certification, and manufacturing license
under part 52 of this chapter must be able to withstand for a specified
duration and recover from a station blackout as defined in Sec. 50.2.
The specified station blackout duration shall be based on the following
factors:
* * * * *
(c) * * *
(1) Information Submittal. For each light-water-cooled nuclear
power plant licensed to operate on or before July 21, 1988, the
licensee shall submit the information defined below to the Director of
the Office of Nuclear Reactor Regulation by April 17, 1989. For each
light-water-cooled nuclear power plant licensed to operate after July
21, 1988, but before September 27, 2007, the licensee shall submit the
information defined in this section to the Director of the Office of
Nuclear Reactor Regulation, by 270 days after the date of license
issuance. For each light-water-cooled nuclear power plant operating
license application submitted after September 27, 2007, the applicant
shall submit the information defined below in its final safety analysis
report.
* * * * *
0
96. In Sec. 50.65, paragraph (a)(1) is revised to read as follows:
Sec. 50.65 Requirements for monitoring the effectiveness of
maintenance at nuclear power plants.
* * * * *
(a)(1) Each holder of an operating license for a nuclear power
plant under this part and each holder of a combined license under part
52 of this chapter after the Commission makes the finding under Sec.
52.103(g) of this chapter, shall monitor the performance or condition
of structures, systems, or components, against licensee-established
goals, in a manner sufficient to provide reasonable assurance that
these structures, systems, and components, as defined in paragraph (b)
of this section, are capable of fulfilling their intended functions.
These goals shall be established commensurate with safety and, where
practical, take into account industry-wide operating experience. When
the performance or condition of a structure, system, or component does
not meet established goals, appropriate corrective action shall be
taken. For a nuclear power plant for which the licensee has submitted
the certifications specified in Sec. 50.82(a)(1) or 52.110(a)(1) of
this chapter, as applicable, this section shall only apply to the
extent that the licensee shall monitor the performance or condition of
all structures, systems, or components associated with the storage,
control, and maintenance of spent fuel in a safe condition, in a manner
sufficient to provide reasonable assurance that these structures,
systems, and components are capable of fulfilling their intended
functions.
* * * * *
0
97. In Sec. 50.70 paragraphs (a) and (b)(2) are revised to read as
follows:
Sec. 50.70 Inspections.
(a) Each applicant for or holder of a license, including a
construction permit or an early site permit, shall permit inspection,
by duly authorized representatives of the Commission, of his records,
premises, activities, and of licensed materials in possession or use,
related to the license or construction permit or early site permit as
may be necessary to effectuate the purposes of the Act, as amended,
including Section 105 of the Act, and the Energy Reorganization Act of
1974, as amended.
(b) * * *
(2) For a site with a single power reactor or fuel facility
licensed under part 50 or part 52 of this chapter, or a facility issued
a manufacturing license under part 52, the space provided shall be
adequate to accommodate a full-time inspector, a part-time secretary
and transient NRC personnel and will be generally commensurate with
other office facilities at the site. A space of 250 square feet either
within the site's office complex or in an office trailer or other
onsite space is suggested as a guide. For sites containing multiple
power reactor units or fuel facilities, additional space may be
requested to accommodate additional full-time inspector(s). The office
space that is provided shall be subject to the approval of the
Director, Office of New Reactors, or the Director, Office of Nuclear
Reactor Regulation. All furniture, supplies and communication equipment
will be furnished by the Commission.
* * * * *
0
98. In Sec. 50.71, paragraphs (a), (c), (d)(1), and the introductory
text of paragraph (e) are revised, paragraph (e)(3)(iii) is added,
paragraph (f) is redesignated as paragraph (g) and revised, and new
paragraphs (f) and (h) are added to read as follows:
Sec. 50.71 Maintenance of records, making of reports.
(a) Each licensee, including each holder of a construction permit
or early site permit, shall maintain all records and make all reports,
in connection with the activity, as may be required by the conditions
of the license or permit or by the regulations, and orders of the
Commission in effectuating the purposes of the Act, including Section
105 of the Act, and the Energy Reorganization Act of 1974, as amended.
Reports must be submitted in accordance with Sec. 50.4 or 10 CFR 52.3,
as applicable.
* * * * *
(c) Records that are required by the regulations in this part or
part 52 of this chapter, by license condition, or by technical
specifications must be retained for the period specified by the
appropriate regulation, license condition, or technical specification.
If a retention period is not otherwise specified, these records must be
retained until the Commission terminates the facility license or, in
the case of an early site permit, until the permit expires.
(d)(1) Records which must be maintained under this part or part 52
of this chapter may be the original or a reproduced copy or microform
if the reproduced copy or microform is duly authenticated by authorized
personnel and the microform is capable of producing a clear and legible
copy after storage for the period specified by Commission regulations.
The record may also be stored in electronic media with the capability
of producing legible, accurate, and complete records during
[[Page 49502]]
the required retention period. Records such as letters, drawings, and
specifications, must include all pertinent information such as stamps,
initials, and signatures. The licensee shall maintain adequate
safeguards against tampering with, and loss of records.
* * * * *
(e) Each person licensed to operate a nuclear power reactor under
the provisions of Sec. 50.21 or Sec. 50.22, and each applicant for a
combined license under part 52 of this chapter, shall update
periodically, as provided in paragraphs (e) (3) and (4) of this
section, the final safety analysis report (FSAR) originally submitted
as part of the application for the license, to assure that the
information included in the report contains the latest information
developed. This submittal shall contain all the changes necessary to
reflect information and analyses submitted to the Commission by the
applicant or licensee or prepared by the applicant or licensee pursuant
to Commission requirement since the submittal of the original FSAR, or
as appropriate, the last update to the FSAR under this section. The
submittal shall include the effects \1\ of all changes made in the
facility or procedures as described in the FSAR; all safety analyses
and evaluations performed by the applicant or licensee either in
support of approved license amendments or in support of conclusions
that changes did not require a license amendment in accordance with
Sec. 50.59(c)(2) or, in the case of a license that references a
certified design, in accordance with Sec. 52.98(c) of this chapter;
and all analyses of new safety issues performed by or on behalf of the
applicant or licensee at Commission request. The updated information
shall be appropriately located within the update to the FSAR.
---------------------------------------------------------------------------
\1\ Effects of changes includes appropriate revisions of
descriptions in the FSAR such that the FSAR (as updated) is complete
and accurate.
---------------------------------------------------------------------------
* * * * *
(3) * * *
(iii) During the period from the docketing of an application for a
combined license under subpart C of part 52 of this chapter until the
Commission makes the finding under Sec. 52.103(g) of this chapter, the
update to the FSAR must be submitted annually.
* * * * *
(f) Each person licensed to manufacture a nuclear power reactor
under subpart F of 10 CFR part 52 shall update the FSAR originally
submitted as part of the application to reflect any modification to the
design that is approved by the Commission under Sec. 52.171 of this
chapter, and any new analyses of the design performed by or on behalf
of the licensee at the NRC's request. This submittal shall contain all
the changes necessary to reflect information and analyses submitted to
the Commission by the licensee or prepared by the licensee with respect
to the modification approved under Sec. 52.171 of this chapter or the
analyses requested by the Commission under Sec. 52.171 of this
chapter. The updated information shall be appropriately located within
the update to the FSAR.
(g) The provisions of this section apply to nuclear power reactor
licensees that have submitted the certification of permanent cessation
of operations required under Sec. Sec. 50.82(a)(1)(i) or 52.110(a)(1)
of this chapter. The provisions of paragraphs (a), (c), and (d) of this
section also apply to non-power reactor licensees that are no longer
authorized to operate.
(h)(1) No later than the scheduled date for initial loading of
fuel, each holder of a combined license under subpart C of 10 CFR part
52 shall develop a level 1 and a level 2 probabilistic risk assessment
(PRA). The PRA must cover those initiating events and modes for which
NRC-endorsed consensus standards on PRA exist one year prior to the
scheduled date for initial loading of fuel.
(2) Each holder of a combined license shall maintain and upgrade
the PRA required by paragraph (h)(1) of this section. The upgraded PRA
must cover initiating events and modes of operation contained in NRC-
endorsed consensus standards on PRA in effect one year prior to each
required upgrade. The PRA must be upgraded every four years until the
permanent cessation of operations under Sec. 52.110(a) of this
chapter.
(3) Each holder of a combined license shall, no later than the date
on which the licensee submits an application for a renewed license,
upgrade the PRA required by paragraph (h)(1) of this section to cover
all modes and all initiating events.
0
99. In Sec. 50.72, the introductory text of paragraph (a)(1) is
revised to read as follows:
Sec. 50.72 Immediate notification requirements for operating nuclear
power reactors.
(a) * * *
(1) Each nuclear power reactor licensee licensed under Sec. Sec.
50.21(b) or 50.22 holding an operating license under this part or a
combined license under part 52 of this chapter after the Commission
makes the finding under Sec. 52.103(g), shall notify the NRC
Operations Center via the Emergency Notification System of:
* * * * *
0
100. In Sec. 50.73, paragraph (a)(1) is revised to read as follows:
Sec. 50.73 Licensee event report system.
(a) * * *
(1) The holder of an operating license under this part or a
combined license under part 52 of this chapter (after the Commission
has made the finding under Sec. 52.103(g) of this chapter) for a
nuclear power plant (licensee) shall submit a Licensee Event Report
(LER) for any event of the type described in this paragraph within 60
days after the discovery of the event. In the case of an invalid
actuation reported under Sec. 50.73(a)(2)(iv), other than actuation of
the reactor protection system (RPS) when the reactor is critical, the
licensee may, at its option, provide a telephone notification to the
NRC Operations Center within 60 days after discovery of the event
instead of submitting a written LER. Unless otherwise specified in this
section, the licensee shall report an event if it occurred within 3
years of the date of discovery regardless of the plant mode or power
level, and regardless of the significance of the structure, system, or
component that initiated the event.
* * * * *
0
101. In Sec. 50.75, paragraphs (a) and (b) are revised, paragraph
(e)(3) is added, paragraphs (f)(1), (f)(2), (f)(3), and (f)(4) are
redesignated as paragraphs (f)(2), (f)(3), (f)(4), and (f)(5),
respectively, and paragraph (f)(1) is added to read as follows:
Sec. 50.75 Reporting and recordkeeping for decommissioning planning.
(a) This section establishes requirements for indicating to NRC how
a licensee will provide reasonable assurance that funds will be
available for the decommissioning process. For power reactor licensees
(except a holder of a manufacturing license under part 52 of this
chapter), reasonable assurance consists of a series of steps as
provided in paragraphs (b), (c), (e), and (f) of this section. Funding
for the decommissioning of power reactors may also be subject to the
regulation of Federal or State Government agencies (e.g., Federal
Energy Regulatory Commission (FERC) and State Public Utility
Commissions) that have jurisdiction over rate regulation. The
requirements of this section, in particular paragraph (c) of this
section, are in addition to, and not substitution for, other
requirements, and are not intended to be used by themselves or by other
agencies to establish rates.
[[Page 49503]]
(b) Each power reactor applicant for or holder of an operating
license, and each applicant for a combined license under subpart C of
10 CFR part 52 for a production or utilization facility of the type and
power level specified in paragraph (c) of this section shall submit a
decommissioning report, as required by Sec. 50.33(k).
(1) For an applicant for or holder of an operating license under
part 50, the report must contain a certification that financial
assurance for decommissioning will be (for a license applicant), or has
been (for a license holder), provided in an amount which may be more,
but not less, than the amount stated in the table in paragraph (c)(1)
of this section adjusted using a rate at least equal to that stated in
paragraph (c)(2) of this section. For an applicant for a combined
license under subpart C of 10 CFR part 52, the report must contain a
certification that financial assurance for decommissioning will be
provided no later than 30 days after the Commission publishes notice in
the Federal Register under Sec. 52.103(a) in an amount which may be
more, but not less, than the amount stated in the table in paragraph
(c)(1) of this section, adjusted using a rate at least equal to that
stated in paragraph (c)(2) of this section.
(2) The amount to be provided must be adjusted annually using a
rate at least equal to that stated in paragraph (c)(2) of this section.
(3) The amount must be covered by one or more of the methods
described in paragraph (e) of this section as acceptable to the NRC.
(4) The amount stated in the applicant's or licensee's
certification may be based on a cost estimate for decommissioning the
facility. As part of the certification, a copy of the financial
instrument obtained to satisfy the requirements of paragraph (e) of
this section must be submitted to NRC; provided, however, that an
applicant for or holder of a combined license need not obtain such
financial instrument or submit a copy to the Commission except as
provided in paragraph (e)(3) of this section.
* * * * *
(e) * * *
(3) Each holder of a combined license under subpart C of 10 CFR
part 52 shall, 2 years before and 1 year before the scheduled date for
initial loading of fuel, consistent with the schedule required by Sec.
52.99(a), submit a report to the NRC containing a certification
updating the information described under paragraph (b)(1) of this
section, including a copy of the financial instrument to be used. No
later than 30 days after the Commission publishes notice in the Federal
Register under 10 CFR 52.103(a), the licensee shall submit a report
containing a certification that financial assurance for decommissioning
is being provided in an amount specified in the licensee's most recent
updated certification, including a copy of the financial instrument
obtained to satisfy the requirements of paragraph (e) of this section.
(f)(1) Each power reactor licensee shall report, on a calendar-year
basis, to the NRC by March 31, 1999, and at least once every 2 years on
the status of its decommissioning funding for each reactor or part of a
reactor that it owns. However, each holder of a combined license under
part 52 of this chapter need not begin reporting until the date that
the Commission has made the finding under Sec. 52.103(g) of this
chapter. The information in this report must include, at a minimum the
amount of decommissioning funds estimated to be required under 10 CFR
50.75(b) and (c); the amount accumulated to the end of the calendar
year preceding the date of the report; a schedule of the annual amounts
remaining to be collected; the assumptions used regarding rates of
escalation in decommissioning costs, rates of earnings on
decommissioning funds, and rates of other factors used in funding
projections; any contracts upon which the licensee is relying under
paragraph (e)(1)(v) of this section; any modifications occurring to a
licensee's current method of providing financial assurance since the
last submitted report; and any material changes to trust agreements.
Any licensee for a plant that is within 5 years of the projected end of
its operation, or where conditions have changed so that it will close
within 5 years (before the end of its licensed life), or has already
closed (before the end of its licensed life), or for plants involved in
mergers or acquisitions shall submit this report annually.
* * * * *
0
102. Section 50.78 is revised to read as follows:
Sec. 50.78 Installation information and verification.
Each holder of a construction permit and each holder of a combined
license shall, if requested by the Commission, submit installation
information on Form-71, permit verification thereof by the
International Atomic Energy Agency, and take other action as may be
necessary to implement the US/IAEA Safeguards Agreement, in the manner
set forth in Sec. 75.6 and Sec. Sec. 75.11 through 75.14 of this
chapter.
0
103. In Sec. 50.80, paragraphs (a) and (b) are revised to read as
follows:
Sec. 50.80 Transfer of licenses.
(a) No license for a production or utilization facility (including,
but not limited to, permits under this part and part 52 of this
chapter, and licenses under parts 50 and 52 of this chapter), or any
right thereunder, shall be transferred, assigned, or in any manner
disposed of, either voluntarily or involuntarily, directly or
indirectly, through transfer of control of the license to any person,
unless the Commission gives its consent in writing.
(b)(1) An application for transfer of a license shall include:
(i) For a construction permit or operating license under this part,
as much of the information described in Sec. Sec. 50.33 and 50.34 of
this part with respect to the identity and technical and financial
qualifications of the proposed transferee as would be required by those
sections if the application were for an initial license. The Commission
may require additional information such as data respecting proposed
safeguards against hazards from radioactive materials and the
applicant's qualifications to protect against such hazards.
(ii) For an early site permit under part 52 of this chapter, as
much of the information described in Sec. Sec. 52.16 and 52.17 of this
chapter with respect to the identity and technical qualifications of
the proposed transferee as would be required by those sections if the
application were for an initial license.
(iii) For a combined license under part 52 of this chapter, as much
of the information described in Sec. Sec. 52.77 and 52.79 of this
chapter with respect to the identity and technical and financial
qualifications of the proposed transferee as would be required by those
sections if the application were for an initial license. The Commission
may require additional information such as data respecting proposed
safeguards against hazards from radioactive materials and the
applicant's qualifications to protect against such hazards.
(iv) For a manufacturing license under part 52 of this chapter, as
much of the information described in Sec. Sec. 52.156 and 52.157 of
this chapter with respect to the identity and technical qualifications
of the proposed transferee as would be required by those sections if
the application were for an initial license.
(2) The application shall include also a statement of the purposes
for which the transfer of the license is requested,
[[Page 49504]]
the nature of the transaction necessitating or making desirable the
transfer of the license, and an agreement to limit access to Restricted
Data pursuant to Sec. 50.37. The Commission may require any person who
submits an application for license pursuant to the provisions of this
section to file a written consent from the existing licensee or a
certified copy of an order or judgment of a court of competent
jurisdiction attesting to the person's right (subject to the licensing
requirements of the Act and these regulations) to possession of the
facility or site involved.
* * * * *
0
104. In Sec. 50.81, paragraph (d)(1) is revised, and a new paragraph
(d)(3) is added to read as follows:
Sec. 50.81 Creditor regulations.
* * * * *
(d) * * *
(1) License includes any license under this chapter, any
construction permit under this part, and any early site permit under
part 52 of this chapter, which may be issued by the Commission with
regard to a facility;
* * * * *
(3) Facility includes but is not limited to, a site which is the
subject of an early site permit under subpart A of part 52 of this
chapter, and a reactor manufactured under a manufacturing license under
subpart F of part 52 of this chapter.
0
105. Section 50.90 is revised to read as follows:
Sec. 50.90 Application for amendment of license, construction permit,
or early site permit.
Whenever a holder of a license, including a construction permit and
operating license under this part, and an early site permit, combined
license, and manufacturing license under part 52 of this chapter,
desires to amend the license or permit, application for an amendment
must be filed with the Commission, as specified in Sec. Sec. 50.4 or
52.3 of this chapter, as applicable, fully describing the changes
desired, and following as far as applicable, the form prescribed for
original applications.
0
106. In Sec. 50.91, the introductory text is revised to read as
follows:
Sec. 50.91 Notice for public comment; State consultation.
The Commission will use the following procedures for an application
requesting an amendment to an operating license under this part or a
combined license under part 52 of this chapter for a facility licensed
under Sec. Sec. 50.21(b) or 50.22, or for a testing facility, except
for amendments subject to hearings governed by 10 CFR part 2, subpart
L. For amendments subject to 10 CFR part 2, subpart L, the following
procedures will apply only to the extent specifically referenced in
Sec. 2.309(b) of this chapter, except that notice of opportunity for
hearing must be published in the Federal Register at least 30 days
before the requested amendment is issued by the Commission:
* * * * *
0
107. Section 50.92 paragraph (a), and the introductory text of
paragraph (c) are revised to read as follows:
Sec. 50.92 Issuance of amendment.
(a) In determining whether an amendment to a license, construction
permit, or early site permit will be issued to the applicant, the
Commission will be guided by the considerations which govern the
issuance of initial licenses, construction permits, or early site
permits to the extent applicable and appropriate. If the application
involves the material alteration of a licensed facility, a construction
permit will be issued before the issuance of the amendment to the
license, provided however, that if the application involves a material
alteration to a nuclear power reactor manufactured under part 52 of
this chapter before its installation at a site, or a combined license
before the date that the Commission makes the finding under Sec.
52.103(g) of this chapter, no application for a construction permit is
required. If the amendment involves a significant hazards
consideration, the Commission will give notice of its proposed action:
(1) Under Sec. 2.105 of this chapter before acting thereon; and
(2) As soon as practicable after the application has been docketed.
* * * * *
(c) The Commission may make a final determination, under the
procedures in Sec. 50.91, that a proposed amendment to an operating
license or a combined license for a facility or reactor licensed under
Sec. Sec. 50.21(b) or 50.22, or for a testing facility involves no
significant hazards consideration, if operation of the facility in
accordance with the proposed amendment would not:
* * * * *
0
108. Section 50.100 is revised to read as follows:
Sec. 50.100 Revocation, suspension, modification of licenses,
permits, and approvals for cause.
A license, permit, or standard design approval under parts 50 or 52
of this chapter may be revoked, suspended, or modified, in whole or in
part, for any material false statement in the application or in the
supplemental or other statement of fact required of the applicant; or
because of conditions revealed by the application or statement of fact
of any report, record, inspection, or other means which would warrant
the Commission to refuse to grant a license, permit, or approval on an
original application (other than those relating to Sec. Sec. 50.51,
50.42(a), and 50.43(b)); or for failure to manufacture a reactor, or
construct or operate a facility in accordance with the terms of the
permit or license, provided, however, that failure to make timely
completion of the proposed construction or alteration of a facility
under a construction permit under part 50 of this chapter or a combined
license under part 52 of this chapter shall be governed by the
provisions of Sec. 50.55(b); or for violation of, or failure to
observe, any of the terms and provisions of the act, regulations,
license, permit, approval, or order of the Commission.
0
109. In Sec. 50.109, paragraph (a)(1) is revised to read as follows:
Sec. 50.109 Backfitting.
(a)(1) Backfitting is defined as the modification of or addition to
systems, structures, components, or design of a facility; or the design
approval or manufacturing license for a facility; or the procedures or
organization required to design, construct or operate a facility; any
of which may result from a new or amended provision in the Commission's
regulations or the imposition of a regulatory staff position
interpreting the Commission's regulations that is either new or
different from a previously applicable staff position after:
(i) The date of issuance of the construction permit for the
facility for facilities having construction permits issued after
October 21, 1985;
(ii) Six (6) months before the date of docketing of the operating
license application for the facility for facilities having construction
permits issued before October 21, 1985;
(iii) The date of issuance of the operating license for the
facility for facilities having operating licenses;
(iv) The date of issuance of the design approval under subpart E of
part 52 of this chapter;
(v) The date of issuance of a manufacturing license under subpart F
of part 52 of this chapter;
(vi) The date of issuance of the first construction permit issued
for a duplicate design under appendix N of this part; or
[[Page 49505]]
(vii) The date of issuance of a combined license under subpart C of
part 52 of this chapter, provided that if the combined license
references an early site permit, the provisions in Sec. 52.39 of this
chapter apply with respect to the site characteristics, design
parameters, and terms and conditions specified in the early site
permit. If the combined license references a standard design
certification rule under subpart B of 10 CFR part 52, the provisions in
Sec. 52.63 of this chapter apply with respect to the design matters
resolved in the standard design certification rule, provided however,
that if any specific backfitting limitations are included in a
referenced design certification rule, those limitations shall govern.
If the combined license references a standard design approval under
subpart E of 10 CFR part 52, the provisions in Sec. 52.145 of this
chapter apply with respect to the design matters resolved in the
standard design approval. If the combined license uses a reactor
manufactured under a manufacturing license under subpart F of 10 CFR
part 52, the provisions of Sec. 52.171 of this chapter apply with
respect to matters resolved in the manufacturing license proceeding.
* * * * *
0
110. Section 50.120 is revised to read as follows:
Sec. 50.120 Training and qualification of nuclear power plant
personnel.
(a) Applicability. The requirements of this section apply to each
applicant for and each holder of an operating license issued under this
part and each holder of a combined license issued under part 52 of this
chapter for a nuclear power plant of the type specified in Sec.
50.21(b) or Sec. 50.22.
(b) Requirements. (1)(i) Each nuclear power plant operating license
applicant, by 18 months prior to fuel load, and each holder of an
operating license shall establish, implement, and maintain a training
program that meets the requirements of paragraphs (b)(2) and (b)(3) of
this section.
(ii) Each holder of a combined license shall establish, implement,
and maintain the training program that meets the requirements of
paragraphs (b)(2) and (b)(3) of this section, as described in the final
safety analysis report no later than 18 months before the scheduled
date for initial loading of fuel.
(2) The training program must be derived from a systems approach to
training as defined in 10 CFR 55.4, and must provide for the training
and qualification of the following categories of nuclear power plant
personnel:
(i) Non-licensed operator.
(ii) Shift supervisor.
(iii) Shift technical advisor.
(iv) Instrument and control technician.
(v) Electrical maintenance personnel.
(vi) Mechanical maintenance personnel.
(vii) Radiological protection technician.
(viii) Chemistry technician.
(ix) Engineering support personnel.
(3) The training program must incorporate the instructional
requirements necessary to provide qualified personnel to operate and
maintain the facility in a safe manner in all modes of operation. The
training program must be developed to be in compliance with the
facility license, including all technical specifications and applicable
regulations. The training program must be periodically evaluated and
revised as appropriate to reflect industry experience as well as
changes to the facility, procedures, regulations, and quality assurance
requirements. The training program must be periodically reviewed by
licensee management for effectiveness. Sufficient records must be
maintained by the licensee to maintain program integrity and kept
available for NRC inspection to verify the adequacy of the program.
0
111. In Appendix A to Part 50, the first paragraph under the
introduction and the second paragraph under Criterion 19 are revised to
read as follows:
Appendix A to Part 50--General Design Criteria for Nuclear Power Plants
* * * * *
Introduction
Under the provisions of Sec. 50.34, an application for a
construction permit must include the principal design criteria for a
proposed facility. Under the provisions of 10 CFR 52.47, 52.79,
52.137, and 52.157, an application for a design certification,
combined license, design approval, or manufacturing license,
respectively, must include the principal design criteria for a
proposed facility. The principal design criteria establish the
necessary design, fabrication, construction, testing, and
performance requirements for structures, systems, and components
important to safety; that is, structures, systems, and components
that provide reasonable assurance that the facility can be operated
without undue risk to the health and safety of the public.
* * * * *
Criterion 19--Control Room.
* * * * *
Applicants for and holders of construction permits and operating
licenses under this part who apply on or after January 10, 1997,
applicants for design approvals or certifications under part 52 of
this chapter who apply on or after January 10, 1997, applicants for
and holders of combined licenses or manufacturing licenses under
part 52 of this chapter who do not reference a standard design
approval or certification, or holders of operating licenses using an
alternative source term under Sec. 50.67, shall meet the
requirements of this criterion, except that with regard to control
room access and occupancy, adequate radiation protection shall be
provided to ensure that radiation exposures shall not exceed 0.05 Sv
(5 rem) total effective dose equivalent (TEDE) as defined in Sec.
50.2 for the duration of the accident.
* * * * *
0
112. In Appendix B to Part 50, the Introduction and Section I are
revised to read as follows:
Appendix B to Part 50--Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Introduction. Every applicant for a construction permit is
required by the provisions of Sec. 50.34 to include in its
preliminary safety analysis report a description of the quality
assurance program to be applied to the design, fabrication,
construction, and testing of the structures, systems, and components
of the facility. Every applicant for an operating license is
required to include, in its final safety analysis report,
information pertaining to the managerial and administrative controls
to be used to assure safe operation. Every applicant for a combined
license under part 52 of this chapter is required by the provisions
of Sec. 52.79 of this chapter to include in its final safety
analysis report a description of the quality assurance applied to
the design, and to be applied to the fabrication, construction, and
testing of the structures, systems, and components of the facility
and to the managerial and administrative controls to be used to
assure safe operation. For applications submitted after September
27, 2007, every applicant for an early site permit under part 52 of
this chapter is required by the provisions of Sec. 52.17 of this
chapter to include in its site safety analysis report a description
of the quality assurance program applied to site activities related
to the design, fabrication, construction, and testing of the
structures, systems, and components of a facility or facilities that
may be constructed on the site. Every applicant for a design
approval or design certification under part 52 of this chapter is
required by the provisions of 10 CFR 52.137 and 52.47, respectively,
to include in its final safety analysis report a description of the
quality assurance program applied to the design of the structures,
systems, and components of the facility. Every applicant for a
manufacturing license under part 52 of this chapter is required by
the provisions of 10 CFR 52.157 to include in its final safety
analysis report a description of the quality assurance program
applied to the design, and to be applied to the manufacture of, the
structures, systems, and components of the reactor. Nuclear power
plants and fuel reprocessing plants include structures, systems, and
components that prevent or mitigate the consequences of
[[Page 49506]]
postulated accidents that could cause undue risk to the health and
safety of the public. This appendix establishes quality assurance
requirements for the design, manufacture, construction, and
operation of those structures, systems, and components. The
pertinent requirements of this appendix apply to all activities
affecting the safety-related functions of those structures, systems,
and components; these activities include designing, purchasing,
fabricating, handling, shipping, storing, cleaning, erecting,
installing, inspecting, testing, operating, maintaining, repairing,
refueling, and modifying.
As used in this appendix, ``quality assurance'' comprises all
those planned and systematic actions necessary to provide adequate
confidence that a structure, system, or component will perform
satisfactorily in service. Quality assurance includes quality
control, which comprises those quality assurance actions related to
the physical characteristics of a material, structure, component, or
system which provide a means to control the quality of the material,
structure, component, or system to predetermined requirements.
I. Organization
The applicant \1\ shall be responsible for the establishment and
execution of the quality assurance program. The applicant may
delegate to others, such as contractors, agents, or consultants, the
work of establishing and executing the quality assurance program, or
any part thereof, but shall retain responsibility for the quality
assurance program. The authority and duties of persons and
organizations performing activities affecting the safety-related
functions of structures, systems, and components shall be clearly
established and delineated in writing. These activities include both
the performing functions of attaining quality objectives and the
quality assurance functions. The quality assurance functions are
those of (1) assuring that an appropriate quality assurance program
is established and effectively executed; and (2) verifying, such as
by checking, auditing, and inspecting, that activities affecting the
safety-related functions have been correctly performed. The persons
and organizations performing quality assurance functions shall have
sufficient authority and organizational freedom to identify quality
problems; to initiate, recommend, or provide solutions; and to
verify implementation of solutions. There persons and organizations
performing quality assurance functions shall report to a management
level so that the required authority and organizational freedom,
including sufficient independence from cost and schedule when
opposed to safety considerations, are provided. Because of the many
variables involved, such as the number of personnel, the type of
activity being performed, and the location or locations where
activities are performed, the organizational structure for executing
the quality assurance program may take various forms, provided that
the persons and organizations assigned the quality assurance
functions have the required authority and organizational freedom.
Irrespective of the organizational structure, the individual(s)
assigned the responsibility for assuring effective execution of any
portion of the quality assurance program at any location where
activities subject to this appendix are being performed, shall have
direct access to the levels of management necessary to perform this
function.
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\1\ While the term ``applicant'' is used in these criteria, the
requirements are, of course, applicable after such a person has
received a license to construct and operate a nuclear power plant or
a fuel reprocessing plant or has received an early site permit,
design approval, design certification, or manufacturing license, as
applicable. These criteria will also be used for guidance in
evaluating the adequacy of quality assurance programs in use by
holders of construction permits, operating licenses, early site
permits, design approvals, combined licenses, and manufacturing
licenses.
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* * * * *
0
113. In Appendix C to Part 50, the heading, the first paragraph of
General Information, and the headings of Sections I.A and II.A, and
Section III are revised to read as follows:
Appendix C to Part 50--A Guide for the Financial Data and Related
Information Required To Establish Financial Qualifications for
Construction Permits and Combined Licenses
General Information
This appendix is intended to appraise applicants for
construction permits and combined licenses for production or
utilization facilities of the types described in Sec. 50.21(b) or
Sec. 50.22, or testing facilities, of the general kinds of
financial data and other related information that will demonstrate
the financial qualification of the applicant to carry out the
activities for which the permit or license is sought. The kind and
depth of information described in this guide is not intended to be a
rigid and absolute requirement. In some instances, additional
pertinent material may be needed. In any case, the applicant should
include information other than that specified, if the information is
pertinent to establishing the applicant's financial ability to carry
out the activities for which the permit or license is sought.
* * * * *
I. * * *
A. Applications for Construction Permits or Combined Licenses
* * * * *
II. * * *
A. Applications for Construction Permits or Combined Licenses
* * * * *
III. Annual Financial Statement
Each holder of a construction permit for a production or
utilization facility of a type described in Sec. 50.21(b) or Sec.
50.22 or a testing facility, and each holder of a combined license
issued under part 52 of this chapter, is required by Sec. 50.71(b)
to file its annual financial report with the Commission at the time
of issuance. This requirement does not apply to licensees or holders
of construction permits for medical and research reactors.
* * * * *
0
114. In Appendix E to Part 50, Sections I, III, IV.F.2.a, IV.F.2.c, and
V are revised, and footnotes 6, 7, 8, 9, 10, and 11 are redesignated as
7, 8, 9, 10, 11, and 12, respectively, and a new footnote 6 is added to
read as follows:
Appendix E to Part 50--Emergency Planning and Preparedness for
Production and Utilization Facilities
* * * * *
I. Introduction
Each applicant for a construction permit is required by Sec.
50.34(a) to include in the preliminary safety analysis report a
discussion of preliminary plans for coping with emergencies. Each
applicant for an operating license is required by Sec. 50.34(b) to
include in the final safety analysis report plans for coping with
emergencies. Each applicant for a combined license under subpart C
of part 52 of this chapter is required by Sec. 52.79 of this
chapter to include in the application plans for coping with
emergencies. Each applicant for an early site permit under subpart A
of part 52 of this chapter may submit plans for coping with
emergencies under Sec. 52.17 of this chapter.
This appendix establishes minimum requirements for emergency
plans for use in attaining an acceptable state of emergency
preparedness. These plans shall be described generally in the
preliminary safety analysis report for a construction permit and
submitted as part of the final safety analysis report for an
operating license. These plans, or major features thereof, may be
submitted as part of the site safety analysis report for an early
site permit.
* * * * *
III. The Final Safety Analysis Report; Site Safety Analysis Report
The final safety analysis report or the site safety analysis
report for an early site permit that includes complete and
integrated emergency plans under Sec. 52.17(b)(2)(ii) of this
chapter shall contain the plans for coping with emergencies. The
plans shall be an expression of the overall concept of operation;
they shall describe the essential elements of advance planning that
have been considered and the provisions that have been made to cope
with emergency situations. The plans shall incorporate information
about the emergency response roles of supporting organizations and
offsite agencies. That information shall be sufficient to provide
assurance of coordination among the supporting groups and with the
licensee. The site safety analysis report for an early site permit
which proposes major features must address the relevant provisions
of 10 CFR 50.47 and 10 CFR part 50, appendix E, within the scope of
emergency preparedness matters addressed in the major features.
The plans submitted must include a description of the elements
set out in Section IV for the emergency planning zones (EPZs)
[[Page 49507]]
to an extent sufficient to demonstrate that the plans provide
reasonable assurance that adequate protective measures can and will
be taken in the event of an emergency.
IV. Content of Emergency Plans
* * * * *
F. * * *
2. * * *
a. A full participation \4\ exercise which tests as much of the
licensee, State, and local emergency plans as is reasonably
achievable without mandatory public participation shall be conducted
for each site at which a power reactor is located.
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\4\ Full participation when used in conjunction with emergency
preparedness exercises for a particular site means appropriate
offsite local and State authorities and licensee personnel
physically and actively take part in testing their integrated
capability to adequately assess and respond to an accident at a
commercial nuclear power plant. Full participation includes testing
major observable portions of the onsite and offsite emergency plans
and mobilization of State, local and licensee personnel and other
resources in sufficient numbers to verify the capability to respond
to the accident scenario.
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(i) For an operating license issued under this part, this
exercise must be conducted within two years before the issuance of
the first operating license for full power (one authorizing
operation above 5 percent of rated power) of the first reactor and
shall include participation by each State and local government
within the plume exposure pathway EPZ and each state within the
ingestion exposure pathway EPZ. If the full participation exercise
is conducted more than 1 year prior to issuance of an operating
licensee for full power, an exercise which tests the licensee's
onsite emergency plans must be conducted within one year before
issuance of an operating license for full power. This exercise need
not have State or local government participation.
(ii) For a combined license issued under part 52 of this
chapter, this exercise must be conducted within two years of the
scheduled date for initial loading of fuel. If the first full
participation exercise is conducted more than one year before the
scheduled date for initial loading of fuel, an exercise which tests
the licensee's onsite emergency plans must be conducted within one
year before the scheduled date for initial loading of fuel. This
exercise need not have State or local government participation. If
DHS identifies one or more deficiencies in the state of offsite
emergency preparedness as the result of the first full participation
exercise, or if the Commission finds that the state of emergency
preparedness does not provide reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency, the provisions of Sec. 50.54(gg) apply.
(iii) For a combined licensee issued under part 52 of this
chapter, if the applicant currently has an operating reactor at the
site, an exercise, either full or partial participation,\5\ shall be
conducted for each subsequent reactor constructed on the site. This
exercise may be incorporated in the exercise requirements of
Sections IV.F.2.b. and c. in this appendix. If DHS identifies one or
more deficiencies in the state of offsite emergency preparedness as
the result of this exercise for the new reactor, or if the
Commission finds that the state of emergency preparedness does not
provide reasonable assurance that adequate protective measures can
and will be taken in the event of a radiological emergency, the
provisions of Sec. 50.54(gg) apply.
---------------------------------------------------------------------------
\5\ Partial participation when used in conjunction with
emergency preparedness exercises for a particular site means
appropriate offsite authorities shall actively take part in the
exercise sufficient to test direction and control functions; i.e.,
(a) protective action decision making related to emergency action
levels, and (b) communication capabilities among affected State and
local authorities and the licensee.
---------------------------------------------------------------------------
* * * * *
c. Offsite plans for each site shall be exercised biennially
with full participation by each offsite authority having a role
under the radiological response plan. Where the offsite authority
has a role under a radiological response plan for more than one
site, it shall fully participate in one exercise every two years and
shall, at least, partially participate in other offsite plan
exercises in this period. If two different licensees whose licensed
facilities are located either on the same site or on adjacent,
contiguous sites, and that share most of the elements defining co-
located licensees,\6\ each licensee shall:
---------------------------------------------------------------------------
\6\ Co-located licensees are two different licensees whose
licensed facilities are located either on the same site or on
adjacent, contiguous sites, and that share most of the following
emergency planning and siting elements:
a. Plume exposure and ingestion emergency planning zones;
b. Offsite governmental authorities;
c. Offsite emergency response organizations;
d. Public notification system; and/or
e. Emergency facilities.
---------------------------------------------------------------------------
(1) Conduct an exercise biennially of its onsite emergency plan;
and
(2) Participate quadrennially in an offsite biennial full or
partial participation exercise; and
(3) Conduct emergency preparedness activities and interactions
in the years between its participation in the offsite full or
partial participation exercise with offsite authorities, to test and
maintain interface among the affected State and local authorities
and the licensee. Co-located licensees shall also participate in
emergency preparedness activities and interaction with offsite
authorities for the period between exercises.
* * * * *
V. Implementing Procedures
No less than 180 days before the scheduled issuance of an
operating license for a nuclear power reactor or a license to
possess nuclear material, or the scheduled date for initial loading
of fuel for a combined license under part 52 of this chapter, the
applicant's or licensee's detailed implementing procedures for its
emergency plan shall be submitted to the Commission as specified in
Sec. 50.4. Licensees who are authorized to operate a nuclear power
facility shall submit any changes to the emergency plan or
procedures to the Commission, as specified in Sec. 50.4, within 30
days of such changes.
* * * * *
0
115. In Appendix I to Part 50, the first paragraphs of Sections I, II,
IV, and V are revised to read as follows:
Appendix I to Part 50--Numerical Guides for Design Objectives and
Limiting Conditions for Operation To Meet the Criterion ``as Low as Is
Reasonably Achievable'' for Radioactive Material in Light-Water-Cooled
Nuclear Power Reactor Effluents
SECTION I. Introduction. Section 50.34a provides that an
application for a construction permit shall include a description of
the preliminary design of equipment to be installed to maintain
control over radioactive materials in gaseous and liquid effluents
produced during normal conditions, including expected occurrences.
In the case of an application filed on or after January 2, 1971, the
application must also identify the design objectives, and the means
to be employed, for keeping levels of radioactive material in
effluents to unrestricted areas as low as practicable. Sections
52.47, 52.79, 52.137, and 52.157 of this chapter provide that
applications for design certification, combined license, design
approval, or manufacturing license, respectively, shall include a
description of the equipment and procedures for the control of
gaseous and liquid effluents and for the maintenance and use of
equipment installed in radioactive waste systems.
* * * * *
SECTION II. Guides on design objectives for light-water-cooled
nuclear power reactors licensed under 10 CFR part 50 or part 52 of
this chapter. The guides on design objectives set forth in this
section may be used by an applicant for a construction permit as
guidance in meeting the requirements of Sec. 50.34a(a), or by an
applicant for a combined license under part 52 of this chapter as
guidance in meeting the requirements of Sec. 50.34a(d), or by an
applicant for a design approval, a design certification, or a
manufacturing license as guidance in meeting the requirements of
Sec. 50.34a(e). The applicant shall provide reasonable assurance
that the following design objectives will be met.
* * * * *
SECTION IV. Guides on technical specifications for limiting
conditions for operation for light-water-cooled nuclear power
reactors licensed under 10 CFR part 50 or part 52 of this chapter.
The guides on limiting conditions for operation for light-water-
cooled nuclear power reactors set forth below may be used by an
applicant for an operating license under this part or a design
certification or combined license under part 52 of this chapter, or
a licensee who has submitted a certification of permanent cessation
of operations under Sec. 50.82(a)(1) or Sec. 52.110 of this
chapter as guidance in developing technical specifications under
Sec. 50.36a(a) to keep levels of radioactive materials in effluents
to unrestricted areas as low as is reasonably achievable.
* * * * *
[[Page 49508]]
SECTION V. Effective dates. A. The guides for limiting
conditions for operation set forth in this appendix shall be
applicable in any case in which an application was filed on or after
January 2, 1971, for a construction permit for a light-water-cooled
nuclear power reactor under this part, or a design certification, a
combined license, or a manufacturing license for a light-water-
cooled nuclear power reactor under part 52 of this chapter.
* * * * *
0
116. In Appendix J to Part 50 in Option A, Section I, and paragraph
II.K are revised and in Option B, Section I, and paragraphs V.B.2 and 3
are revised to read as follows:
Appendix J to Part 50--Primary Reactor Containment Leakage Testing for
Water-Cooled Reactors
* * * * *
Option A--Prescriptive Requirements
* * * * *
I. Introduction
One of the conditions of all operating licenses under this part
and combined licenses under part 52 of this chapter for water-cooled
power reactors as specified in Sec. 50.54(o) is that primary
reactor containments shall meet the containment leakage test
requirements set forth in this appendix. These test requirements
provide for preoperational and periodic verification by tests of the
leak-tight integrity of the primary reactor containment, and systems
and components which penetrate containment of water-cooled power
reactors, and establish the acceptance criteria for these tests. The
purposes of the tests are to assure that (a) leakage through the
primary reactor containment and systems and components penetrating
primary containment shall not exceed allowable leakage rate values
as specified in the technical specifications or associated bases;
and (b) periodic surveillance of reactor containment penetrations
and isolation valves is performed so that proper maintenance and
repairs are made during the service life of the containment, and
systems and components penetrating primary containment. These test
requirements may also be used for guidance in establishing
appropriate containment leakage test requirements in technical
specifications or associated bases for other types of nuclear power
reactors.
II. * * *
K. La (percent/24 hours) means the maximum allowable leakage
rate at pressure Pa as specified for preoperational tests in the
technical specifications or associated bases, and as specified for
periodic tests in the operating license or combined license,
including the technical specifications in any referenced design
certification or manufactured reactor used at the facility.
* * * * *
Option B--Performance-Based Requirements
* * * * *
I. Introduction
One of the conditions required of all operating licenses and
combined licenses for light-water-cooled power reactors as specified
in Sec. 50.54(o) is that primary reactor containments meet the
leakage-rate test requirements in either Option A or B of this
appendix. These test requirements ensure that (a) leakage through
these containments or systems and components penetrating these
containments does not exceed allowable leakage rates specified in
the technical specifications; and (b) integrity of the containment
structure is maintained during its service life. Option B of this
appendix identifies the performance-based requirements and criteria
for preoperational and subsequent periodic leakage-rate testing.\3\
---------------------------------------------------------------------------
\3\ Specific guidance concerning a performance-based leakage-
test program, acceptable leakage-rate test methods, procedures, and
analyses that may be used to implement these requirements and
criteria are provided in Regulatory Guide 1.163, ``Performance-Based
Containment Leak-Test Program.''
---------------------------------------------------------------------------
* * * * *
V. * * *
B. * * *
2. A licensee or applicant for an operating license under this
part or a combined license under part 52 of this chapter may adopt
Option B, or parts thereof, as specified in Section V.A of this
appendix, by submitting its implementation plan and request for
revision to technical specifications (see paragraph B.3 of this
section) to the Director of the Office of Nuclear Reactor Regulation
or the Director of the Office of New Reactors, as appropriate.
3. The regulatory guide or other implementation document used by
a licensee or applicant for an operating license under this part or
a combined license under part 52 of this chapter to develop a
performance-based leakage-testing program must be included, by
general reference, in the plant technical specifications. The
submittal for technical specification revisions must contain
justification, including supporting analyses, if the licensee
chooses to deviate from methods approved by the Commission and
endorsed in a regulatory guide.
* * * * *
Appendix M to Part 50 [Removed]
0
117. Appendix M to Part 50 is removed and reserved.
0
118. The heading for appendix N to part 50 is revised to read as
follows:
Appendix N to Part 50--Standardization of Nuclear Power Plant Designs:
Permits To Construct and Licenses To Operate Nuclear Power Reactors of
Identical Design at Multiple Sites
Appendix O to Part 50 [Removed]
0
119. Appendix O to Part 50 is removed and reserved.
0
120. In Appendix S to Part 50, the first paragraph titled ``General
Information,'' Section I(a), and Section III are revised to read as
follows:
Appendix S to Part 50--Earthquake Engineering Criteria for Nuclear
Power Plants
General Information
This appendix applies to applicants for a construction permit or
operating license under part 50, or a design certification, combined
license, design approval, or manufacturing license under part 52 of
this chapter, on or after January 10, 1997. However, for either an
operating license applicant or holder whose construction permit was
issued before January 10, 1997, the earthquake engineering criteria
in Section VI of appendix A to 10 CFR part 100 continue to apply.
Paragraphs IV.a.1.i, IV.a.1.ii, IV.4.b, and IV.4.c of this appendix
apply to applicants for an early site permit under part 52.
I. Introduction
(a) Each applicant for a construction permit, operating license,
design certification, combined license, design approval, or
manufacturing license is required by Sec. Sec. 50.34(a)(12),
50.34(b)(10), or 10 CFR 52.47, 52.79, 52.137, or 52.157, and General
Design Criterion 2 of appendix A to this part, to design nuclear
power plant structures, systems, and components important to safety
to withstand the effects of natural phenomena, such as earthquakes,
without loss of capability to perform their safety functions. Also,
as specified in Sec. 50.54(ff), nuclear power plants that have
implemented the earthquake engineering criteria described herein
must shut down if the criteria in paragraph IV(a)(3) of this
appendix are exceeded.
* * * * *
III. Definitions
As used in these criteria:
Combined license means a combined construction permit and
operating license with conditions for a nuclear power facility
issued under subpart C of part 52 of this chapter.
Design Approval means an NRC staff approval, issued under
subpart E of part 52 of this chapter, of a final standard design for
a nuclear power reactor of the type described in 10 CFR 50.22.
Design Certification means a Commission approval, issued under
subpart B of part 52 of this chapter, of a standard design for a
nuclear power facility.
Manufacturing license means a license, issued under subpart F of
part 52 of this chapter, authorizing the manufacture of nuclear
power reactors but not their installation into facilities located at
the sites on which the facilities are to be operated.
Operating basis earthquake ground motion (OBE) is the vibratory
ground motion for which those features of the nuclear power plant
necessary for continued operation without undue risk to the health
and safety of the public will remain functional. The operating basis
earthquake ground motion is only associated with plant shutdown and
inspection unless specifically selected by the applicant as a design
input.
[[Page 49509]]
Response spectrum is a plot of the maximum responses
(acceleration, velocity, or displacement) of idealized single-
degree-of-freedom oscillators as a function of the natural
frequencies of the oscillators for a given damping value. The
response spectrum is calculated for a specified vibratory motion
input at the oscillators' supports.
Safe-shutdown earthquake ground motion (SSE) is the vibratory
ground motion for which certain structures, systems, and components
must be designed to remain functional.
Structures, systems, and components required to withstand the
effects of the safe-shutdown earthquake ground motion or surface
deformation are those necessary to assure:
(1) The integrity of the reactor coolant pressure boundary;
(2) The capability to shut down the reactor and maintain it in a
safe-shutdown condition; or
(3) The capability to prevent or mitigate the consequences of
accidents that could result in potential offsite exposures
comparable to the guideline exposures of Sec. 50.34(a)(1).
Surface deformation is distortion of geologic strata at or near
the ground surface by the processes of folding or faulting as a
result of various earth forces. Tectonic surface deformation is
associated with earthquake processes.
* * * * *
PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC
LICENSING AND RELATED REGULATORY FUNCTIONS
0
121. The authority citation for part 51 continues to read as follows:
Authority: Sec. 161, 68 Stat. 948, as amended, sec. 1701, 106
Stat. 2951, 2952, 2953 (42 U.S.C. 2201, 2297f); secs. 201, as
amended, 202, 88 Stat. 1242, as amended, 1244 (42 U.S.C. 5841,
5842); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note). Subpart A
also issued under National Environmental Policy Act of 1969, secs.
102, 104, 105, 83 Stat. 853-854, as amended (42 U.S.C. 4332, 4334,
4335); and Pub. L. 95-604, Title II, 92 Stat. 3033-3041; and sec.
193, Pub. L. 101-575, 104 Stat. 2835 (42 U.S.C. 2243). Sections
51.20, 51.30, 51.60, 51.80, and 51.97 also issued under secs. 135,
141, Pub. L. 97-425, 96 Stat. 2232, 2241, and sec. 148, Pub. L. 100-
203, 101 Stat. 1330-223 (42 U.S.C. 10155, 10161, 10168). Section
51.22 also issued under sec. 274, 73 Stat. 688, as amended by 92
Stat. 3036-3038 (42 U.S.C. 2021) and under Nuclear Waste Policy Act
of 1982, sec. 121, 96 Stat. 2228 (42 U.S.C. 10141). Sections 51.43,
51.67, and 51.109 also issued under Nuclear Waste Policy Act of
1982, sec. 114(f), 96 Stat. 2216, as amended (42 U.S.C. 10134(f)).
0
122. In Sec. 51.17, paragraph (b) is revised to read as follows:
Sec. 51.17 Information collection requirements; OMB approval.
* * * * *
(b) The approved information collection requirements in this part
appear in Sec. Sec. 51.6, 51.16, 51.41, 51.45, 51.50, 51.51, 51.52,
51.53, 51.54, 51.55, 51.58, 51.60, 51.61, 51.62, 51.66, 51.68, and
51.69.
0
123. In Sec. 51.20, paragraphs (b)(1) and (b)(2) are revised, and
paragraph (b)(6) is removed and reserved.
The revisions read as follows:
Sec. 51.20 Criteria for and identification of licensing and
regulatory actions requiring environmental impact statements.
* * * * *
(b)* * *
(1) Issuance of a limited work authorization or a permit to
construct a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 50 of this chapter, or issuance of an
early site permit under part 52 of this chapter.
(2) Issuance or renewal of a full power or design capacity license
to operate a nuclear power reactor, testing facility, or fuel
reprocessing plant under part 50 of this chapter, or a combined license
under part 52 of this chapter.
* * * * *
(6) [Reserved]
* * * * *
0
124. In Sec. 51.22, the introductory text of paragraph (c)(3),
paragraphs (c)(3)(i) and (c)(9), the introductory text of paragraphs
(c)(10) and (c)(12), and paragraph (c)(17) are revised, and paragraphs
(c)(22) and (c)(23) are added to read as follows:
Sec. 51.22 Criterion for categorical exclusion; identification of
licensing and regulatory actions eligible for categorical exclusion or
otherwise not requiring environmental review.
* * * * *
(c) * * *
(3) Amendments to parts 20, 30, 31, 32, 33, 34, 35, 39, 40, 50, 51,
52, 54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and 100 of this chapter
which relate to--
(i) Procedures for filing and reviewing applications for licenses
or construction permits or early site permits or other forms of
permission or for amendments to or renewals of licenses or construction
permits or early site permits or other forms of permission;
* * * * *
(9) Issuance of an amendment to a permit or license for a reactor
under part 50 or part 52 of this chapter, which changes a requirement
with respect to installation or use of a facility component located
within the restricted area, as defined in part 20 of this chapter, or
which changes an inspection or a surveillance requirement, provided
that--
(i) The amendment involves no significant hazards consideration;
(ii) There is no significant change in the types or significant
increase in the amounts of any effluents that may be released offsite;
and
(iii) There is no significant increase in individual or cumulative
occupational radiation exposure.
(10) Issuance of an amendment to a permit or license under parts
30, 31, 32, 33, 34, 35, 36, 39, 40, 50, 52, 60, 61, 63, 70, or part 72
of this chapter which--
* * * * *
(12) Issuance of an amendment to a license under parts 50, 52, 60,
61, 63, 70, 72, or 75 of this chapter relating solely to safeguards
matters (i.e., protection against sabotage or loss or diversion of
special nuclear material) or issuance of an approval of a safeguards
plan submitted under parts 50, 52, 70, 72, and 73 of this chapter,
provided that the amendment or approval does not involve any
significant construction impacts. These amendments and approvals are
confined to--
* * * * *
(17) Issuance of an amendment to a permit or license under parts
30, 40, 50, 52, or part 70 of this chapter which deletes any limiting
condition of operation or monitoring requirement based on or applicable
to any matter subject to the provisions of the Federal Water Pollution
Control Act.
* * * * *
(22) Issuance of a standard design approval under part 52 of this
chapter.
(23) The Commission finding for a combined license under Sec.
52.103(g) of this chapter.
* * * * *
0
125. In Sec. 51.23 paragraphs (b) and (c) are revised to read as
follows:
Sec. 51.23 Temporary storage of spent fuel after cessation of reactor
operation--generic determination of no significant environmental
impact.
* * * * *
(b) Accordingly, as provided in Sec. Sec. 51.30(b), 51.53, 51.61,
51.80(b), 51.95, and 51.97(a), and within the scope of the generic
determination in paragraph (a) of this section, no discussion of any
environmental impact of spent fuel storage in reactor facility storage
pools or independent spent fuel storage installations (ISFSI) for the
period following the term of the reactor operating license or
amendment, reactor combined license or amendment, or initial ISFSI
license or amendment for which application is made, is required in any
environmental report, environmental impact statement, environmental
assessment, or other analysis prepared in connection with the issuance
or amendment of an
[[Page 49510]]
operating license for a nuclear power reactor under parts 50 and 54 of
this chapter, or issuance or amendment of a combined license for a
nuclear power reactor under parts 52 and 54 of this chapter, or the
issuance of an initial license for storage of spent fuel at an ISFSI,
or any amendment thereto.
(c) This section does not alter any requirements to consider the
environmental impacts of spent fuel storage during the term of a
reactor operating license or combined license, or a license for an
ISFSI in a licensing proceeding.
0
126. In Sec. 51.26, a new paragraph (d) is added to read as follows:
Sec. 51.26 Requirement to publish notice of intent and conduct
scoping process.
* * * * *
(d) Whenever the appropriate NRC staff director determines that a
supplement to an environmental impact statement will be prepared by the
NRC, a notice of intent will be prepared as provided in Sec. 51.27,
and will be published in the Federal Register as provided in Sec.
51.116. The NRC staff need not conduct a scoping process (see
Sec. Sec. 51.27, 51.28, and 51.29), provided, however, that if scoping
is conducted, then the scoping must be directed at matters to be
addressed in the supplement. If scoping is conducted in a proceeding
for a combined license referencing an early site permit under part 52,
then the scoping must be directed at matters to be addressed in the
supplement as described in Sec. 51.92(e).
0
127. In Sec. 51.27, the introductory text of paragraph (a) is revised,
and a new paragraph (b) is added to read as follows:
Sec. 51.27 Notice of intent.
(a) The notice of intent required by Sec. 51.26(a) shall:
* * * * *
(b) The notice of intent required by Sec. 51.26(d) shall:
(1) State that a supplement to a final environmental impact
statement will be prepared in accordance with Sec. 51.72 or Sec.
51.92. For a combined license application that references an early site
permit, the supplement to the early site permit environmental impact
statement will be prepared in accordance with Sec. 51.92(e);
(2) Describe the proposed action and, to the extent required,
possible alternatives. For the case of a combined license referencing
an early site permit, identify the proposed action as the issuance of a
combined license for the construction and operation of a nuclear power
plant as described in the combined license application at the site
described in the early site permit referenced in the combined license
application;
(3) Identify the environmental report prepared by the applicant and
information on where copies are available for public inspection;
(4) Describe the matters to be addressed in the supplement to the
final environmental impact statement;
(5) Describe any proposed scoping process that the NRC staff may
conduct, including the role of participants, whether written comments
will be accepted, the last date for submitting comments and where
comments should be sent, whether a public scoping meeting will be held,
the time and place of any scoping meeting or when the time and place of
the meeting will be announced; and
(6) State the name, address, and telephone number of an individual
in NRC who can provide information about the proposed action, the
scoping process, and the supplement to the environmental impact
statement.
0
128. In Sec. 51.29, the section heading and paragraph (a)(1) are
revised to read as follows:
Sec. 51.29 Scoping-environmental impact statement and supplement to
environmental impact statement.
(a) * * *
(1) Define the proposed action which is to be the subject of the
statement or supplement. For environmental impact statements other than
a supplement to an early site permit final environmental impact
statement prepared for a combined license application, the provisions
of 40 CFR 1502.4 will be used for this purpose. For a supplement to an
early site permit final environmental impact statement prepared for a
combined license application, the proposed action shall be as set forth
in the relevant provisions of Sec. 51.92(e).
* * * * *
0
129. In Sec. 51.30, the introductory text of paragraph (a) is revised,
and paragraphs (d) and (e) are added to read as follows:
Sec. 51.30 Environmental assessment.
(a) An environmental assessment for proposed actions, other than
those for a standard design certification under 10 CFR part 52 or a
manufacturing license under part 52, shall identify the proposed action
and include:
* * * * *
(d) An environmental assessment for a standard design certification
under subpart B of part 52 of this chapter must identify the proposed
action, and will be limited to the consideration of the costs and
benefits of severe accident mitigation design alternatives and the
bases for not incorporating severe accident mitigation design
alternatives in the design certification. An environmental assessment
for an amendment to a design certification will be limited to the
consideration of whether the design change which is the subject of the
proposed amendment renders a severe accident mitigation design
alternative previously rejected in the earlier environmental assessment
to become cost beneficial, or results in the identification of new
severe accident mitigation design alternatives, in which case the costs
and benefits of new severe accident mitigation design alternatives and
the bases for not incorporating new severe accident mitigation design
alternatives in the design certification must be addressed.
(e) An environmental assessment for a manufacturing license under
subpart F of part 52 of this chapter must identify the proposed action,
and will be limited to the consideration of the costs and benefits of
severe accident mitigation design alternatives and the bases for not
incorporating severe accident mitigation design alternatives in the
manufacturing license. An environmental assessment for an amendment to
a manufacturing license will be limited to consideration of whether the
design change which is the subject of the proposed amendment either
renders a severe accident mitigation design alternative previously
rejected in an environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation design
alternatives, in which case the costs and benefits of new severe
accident mitigation design alternatives and the bases for not
incorporating new severe accident mitigation design alternatives in the
manufacturing license must be addressed. In either case, the
environmental assessment will not address the environmental impacts
associated with manufacturing the reactor under the manufacturing
license.
0
130. Section 51.31 is revised to read as follows:
Sec. 51.31 Determinations based on environmental assessment.
(a) General. Upon completion of an environmental assessment for
proposed actions other than those involving a standard design
certification or a manufacturing license under part 52 of this chapter,
the appropriate NRC staff director will determine whether to prepare an
environmental impact statement or a finding of no significant
[[Page 49511]]
impact on the proposed action. As provided in Sec. 51.33, a
determination to prepare a draft finding of no significant impact may
be made.
(b) Standard design certification. (1) For actions involving the
issuance or amendment of a standard design certification, the
Commission shall prepare a draft environmental assessment for public
comment as part of the proposed rule. The proposed rule must state
that:
(i) The Commission has determined in Sec. 51.32 that there is no
significant environmental impact associated with the issuance of the
standard design certification or its amendment, as applicable; and
(ii) Comments on the environmental assessment will be limited to
the consideration of SAMDAs as required by Sec. 51.30(d).
(2) The Commission will prepare a final environmental assessment
following the close of the public comment period for the proposed
standard design certification.
(c) Manufacturing license. (1) Upon completion of the environmental
assessment for actions involving issuance or amendment of a
manufacturing license (manufacturing license environmental assessment),
the appropriate NRC staff director will determine the costs and
benefits of severe accident mitigation design alternatives and the
bases for not incorporating severe accident mitigation design
alternatives in the design of the reactor to be manufactured under the
manufacturing license. The NRC staff director may determine to prepare
a draft environmental assessment.
(2) The manufacturing license environmental assessment must state
that:
(i) The Commission has determined in Sec. 51.32 that there is no
significant environmental impact associated with the issuance of a
manufacturing license or an amendment to a manufacturing license, as
applicable;
(ii) The environmental assessment will not address the
environmental impacts associated with manufacturing the reactor under
the manufacturing license; and
(iii) Comments on the environmental assessment will be limited to
the consideration of severe accident mitigation design alternatives as
required by Sec. 51.30(e).
(3) If the NRC staff director makes a determination to prepare and
issue a draft environmental assessment for public review and comment
before making a final determination on the manufacturing license
application, the assessment will be marked, ``Draft.'' The NRC notice
of availability on the draft environmental assessment will include a
request for comments which specifies where comments should be submitted
and when the comment period expires. The notice will state that copies
of the environmental assessment and any related environmental documents
are available for public inspection and where inspections can be made.
A copy of the final environmental assessment will be sent to the U.S.
Environmental Protection Agency, the applicant, any party to a
proceeding, each commenter, and any other Federal, State, and local
agencies, and Indian tribes, State, regional, and metropolitan
clearinghouses expressing an interest in the action. Additional copies
will be made available in accordance with Sec. 51.123.
(4) When a hearing is held under the regulations in part 2 of this
chapter on the proposed issuance of the manufacturing license or
amendment, the NRC staff director will prepare a final environmental
assessment which may be subject to modification as a result of review
and decision as appropriate to the nature and scope of the proceeding.
(5) Only a party admitted into the proceeding with respect to a
contention on the environmental assessment, or an entity participating
in the proceeding pursuant to Sec. 2.315(c) of this chapter, may take
a position and offer evidence on the matters within the scope of the
environmental assessment.
0
131. In Sec. 51.32, paragraph (b) is added to read as follows:
Sec. 51.32 Finding of no significant impact.
* * * * *
(b) The Commission finds that there is no significant environmental
impact associated with the issuance of:
(1) A standard design certification under subpart B of part 52 of
this chapter;
(2) An amendment to a design certification;
(3) A manufacturing license under subpart F of part 52 of this
chapter; or
(4) An amendment to a manufacturing license.
0
132. In Sec. 51.45, paragraphs (a) and (c) are revised to read as
follows:
Sec. 51.45 Environmental report.
(a) General. As required by Sec. Sec. 51.50, 51.53, 51.54, 51.55,
51.60, 51.61, 51.62, or 51.68, as appropriate, each applicant or
petitioner for rulemaking shall submit with its application or petition
for rulemaking one signed original of a separate document entitled
``Applicant's'' or ``Petitioner's Environmental Report,'' as
appropriate. An applicant or petitioner for rulemaking may submit a
supplement to an environmental report at any time.
* * * * *
(c) Analysis. The environmental report shall include an analysis
that considers and balances the environmental effects of the proposed
action, the environmental impacts of alternatives to the proposed
action, and alternatives available for reducing or avoiding adverse
environmental effects. Except for environmental reports prepared at the
early site permit stage under Sec. 51.50(b), or environmental reports
prepared at the license renewal stage under Sec. 51.53(c), the
analysis in the environmental report should also include consideration
of the economic, technical, and other benefits and costs of the
proposed action and of alternatives. Environmental reports prepared at
the license renewal stage under Sec. 51.53(c) need not discuss the
economic or technical benefits and costs of either the proposed action
or alternatives except insofar as these benefits and costs are either
essential for a determination regarding the inclusion of an alternative
in the range of alternatives considered or relevant to mitigation. In
addition, environmental reports prepared under Sec. 51.53(c) need not
discuss issues not related to the environmental effects of the proposed
action and its alternatives. The analyses for environmental reports
shall, to the fullest extent practicable, quantify the various factors
considered. To the extent that there are important qualitative
considerations or factors that cannot be quantified, those
considerations or factors shall be discussed in qualitative terms. The
environmental report should contain sufficient data to aid the
Commission in its development of an independent analysis.
* * * * *
0
133. Section 51.50 is revised to read as follows:
Sec. 51.50 Environmental report--construction permit, early site
permit, or combined license stage.
(a) Construction permit stage. Each applicant for a permit to
construct a production or utilization facility covered by Sec. 51.20
shall submit with its application a separate document, entitled
``Applicant's Environmental Report--Construction Permit Stage,'' which
shall contain the information specified in Sec. Sec. 51.45, 51.51, and
51.52. Each environmental report shall identify procedures for
reporting and keeping records of environmental data, and any conditions
and monitoring requirements for protecting the non-aquatic
[[Page 49512]]
environment, proposed for possible inclusion in the license as
environmental conditions in accordance with Sec. 50.36b of this
chapter.
(b) Early site permit stage. Each applicant for an early site
permit shall submit with its application a separate document, entitled
``Applicant's Environmental Report--Early Site Permit Stage,'' which
shall contain the information specified in Sec. Sec. 51.45, 51.51, and
51.52, as modified in this paragraph.
(1) The environmental report must include an evaluation of
alternative sites to determine whether there is any obviously superior
alternative to the site proposed.
(2) The environmental report may address one or more of the
environmental effects of construction and operation of a reactor, or
reactors, which have design characteristics that fall within the site
characteristics and design parameters for the early site permit
application, provided however, that the environmental report must
address all environmental effects of construction and operation
necessary to determine whether there is any obviously superior
alternative to the site proposed. The environmental report need not
include an assessment of the economic, technical, or other benefits
(for example, need for power) and costs of the proposed action or an
evaluation of alternative energy sources.
(3) For other than light-water-cooled nuclear power reactors, the
environmental report must contain the basis for evaluating the
contribution of the environmental effects of fuel cycle activities for
the nuclear power reactor.
(4) Each environmental report must identify the procedures for
reporting and keeping records of environmental data, and any conditions
and monitoring requirements for protecting the non-aquatic environment,
proposed for possible inclusion in the license as environmental
conditions in accordance with Sec. 50.36b of this chapter.
(c) Combined license stage. Each applicant for a combined license
shall submit with its application a separate document, entitled
``Applicant's Environmental Report--Combined License Stage.'' Each
environmental report shall contain the information specified in
Sec. Sec. 51.45, 51.51, and 51.52, as modified in this paragraph. For
other than light-water-cooled nuclear power reactors, the environmental
report shall contain the basis for evaluating the contribution of the
environmental effects of fuel cycle activities for the nuclear power
reactor. Each environmental report shall identify procedures for
reporting and keeping records of environmental data, and any conditions
and monitoring requirements for protecting the non-aquatic environment,
proposed for possible inclusion in the license as environmental
conditions in accordance with Sec. 50.36b of this chapter. The
combined license environmental report may reference information
contained in a final environmental document previously prepared by the
NRC staff.
(1) Application referencing an early site permit. If the combined
license application references an early site permit, then the
``Applicant's Environmental Report--Combined License Stage'' need not
contain information or analyses submitted to the Commission in
``Applicant's Environmental Report--Early Site Permit Stage,'' or
resolved in the Commission's early site permit environmental impact
statement, but must contain, in addition to the environmental
information and analyses otherwise required:
(i) Information to demonstrate that the design of the facility
falls within the site characteristics and design parameters specified
in the early site permit;
(ii) Information to resolve any significant environmental issue
that was not resolved in the early site permit proceeding;
(iii) Any new and significant information for issues related to the
impacts of construction and operation of the facility that were
resolved in the early site permit proceeding;
(iv) A description of the process used to identify new and
significant information regarding the NRC's conclusions in the early
site permit environmental impact statement. The process must use a
reasonable methodology for identifying such new and significant
information; and
(v) A demonstration that all environmental terms and conditions
that have been included in the early site permit will be satisfied by
the date of issuance of the combined license. Any terms or conditions
of the early site permit that could not be met by the time of issuance
of the combined license, must be set forth as terms or conditions of
the combined license.
(2) Application referencing standard design certification. If the
combined license references a standard design certification, then the
combined license environmental report may incorporate by reference the
environmental assessment previously prepared by the NRC for the
referenced design certification. If the design certification
environmental assessment is referenced, then the combined license
environmental report must contain information to demonstrate that the
site characteristics for the combined license site fall within the site
parameters in the design certification environmental assessment.
(3) Application referencing a manufactured reactor. If the combined
license application proposes to use a manufactured reactor, then the
combined license environmental report may incorporate by reference the
environmental assessment previously prepared by the NRC for the
underlying manufacturing license. If the manufacturing license
environmental assessment is referenced, then the combined license
environmental report must contain information to demonstrate that the
site characteristics for the combined license site fall within the site
parameters in the manufacturing license environmental assessment. The
environmental report need not address the environmental impacts
associated with manufacturing the reactor under the manufacturing
license.
0
134. In Sec. 51.51 paragraph (a) is revised to read as follows:
Sec. 51.51 Uranium fuel cycle environmental data--Table S-3.
(a) Under Sec. 51.50, every environmental report prepared for the
construction permit stage or early site permit stage or combined
license stage of a light-water-cooled nuclear power reactor, and
submitted on or after September 4, 1979, shall take Table S-3, Table of
Uranium Fuel Cycle Environmental Data, as the basis for evaluating the
contribution of the environmental effects of uranium mining and
milling, the production of uranium hexafluoride, isotopic enrichment,
fuel fabrication, reprocessing of irradiated fuel, transportation of
radioactive materials and management of low-level wastes and high-level
wastes related to uranium fuel cycle activities to the environmental
costs of licensing the nuclear power reactor. Table S-3 shall be
included in the environmental report and may be supplemented by a
discussion of the environmental significance of the data set forth in
the table as weighed in the analysis for the proposed facility.
* * * * *
0
135. In Sec. 51.52, the introductory paragraph is revised to read as
follows:
Sec. 51.52 Environmental effects of transportation of fuel and
waste--Table S-4.
Under Sec. 51.50, every environmental report prepared for the
construction permit stage or early site permit stage or combined
license stage of a light-water-
[[Page 49513]]
cooled nuclear power reactor, and submitted after February 4, 1975,
shall contain a statement concerning transportation of fuel and
radioactive wastes to and from the reactor. That statement shall
indicate that the reactor and this transportation either meet all of
the conditions in paragraph (a) of this section or all of the
conditions of paragraph (b) of this section.
* * * * *
0
136. In Sec. 51.53, paragraph (a) and the introductory text of
paragraph (c)(3) are revised to read as follows:
Sec. 51.53 Postconstruction environmental reports.
(a) General. Any environmental report prepared under the provisions
of this section may incorporate by reference any information contained
in a prior environmental report or supplement thereto that relates to
the production or utilization facility or site, or any information
contained in a final environmental document previously prepared by the
NRC staff that relates to the production or utilization facility or
site. Documents that may be referenced include, but are not limited to,
the final environmental impact statement; supplements to the final
environmental impact statement, including supplements prepared at the
license renewal stage; NRC staff-prepared final generic environmental
impact statements; and environmental assessments and records of
decisions prepared in connection with the construction permit,
operating license, early site permit, combined license and any license
amendment for that facility.
* * * * *
(c) * * *
(3) For those applicants seeking an initial renewed license and
holding an operating license, construction permit, or combined license
as of June 30, 1995, the environmental report shall include the
information required in paragraph (c)(2) of this section subject to the
following conditions and considerations:
* * * * *
0
137. Section 51.54 is revised to read as follows:
Sec. 51.54 Environmental report--manufacturing license.
(a) Each applicant for a manufacturing license under subpart F of
part 52 of this chapter shall submit with its application a separate
document entitled, ``Applicant's Environmental Report--Manufacturing
License.'' The environmental report must address the costs and benefits
of severe accident mitigation design alternatives, and the bases for
not incorporating severe accident mitigation design alternatives into
the design of the reactor to be manufactured. The environmental report
need not address the environmental impacts associated with
manufacturing the reactor under the manufacturing license, the benefits
and impacts of utilizing the reactor in a nuclear power plant, or an
evaluation of alternative energy sources.
(b) Each applicant for an amendment to a manufacturing license
shall submit with its application a separate document entitled,
``Applicant's Supplemental Environmental Report--Amendment to
Manufacturing License.'' The environmental report must address whether
the design change which is the subject of the proposed amendment either
renders a severe accident mitigation design alternative previously
rejected in an environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation design
alternatives that may be reasonably incorporated into the design of the
manufactured reactor. The environmental report need not address the
environmental impacts associated with manufacturing the reactor under
the manufacturing license.
0
138. Section 51.55 is redesignated as Sec. 51.58, and is revised to
read as follows:
Sec. 51.58 Environmental report-number of copies; distribution.
(a) Each applicant for a license or permit to site, construct,
manufacture, or operate a production or utilization facility covered by
Sec. Sec. 51.20(b)(1), (b)(2), (b)(3), or (b)(4), each applicant for
renewal of an operating or combined license for a nuclear power plant,
each applicant for a license amendment authorizing the decommissioning
of a production or utilization facility covered by Sec. 51.20, and
each applicant for a license or license amendment to store spent fuel
at a nuclear power plant after expiration of the operating license or
combined license for the nuclear power plant shall submit a copy to the
Director of the Office of Nuclear Reactor Regulation, the Director of
the Office of New Reactors, the Director of the Office of Nuclear
Material Safety and Safeguards, as appropriate, of an environmental
report or any supplement to an environmental report. These reports must
be sent either by mail addressed: ATTN: Document Control Desk; U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; by hand
delivery to the NRC's offices at 11555 Rockville Pike, Rockville,
Maryland, between the hours of 7:30 a.m. and 4:15 p.m. eastern time;
or, where practicable, by electronic submission, for example, via
Electronic Information Exchange, or CD-ROM. Electronic submissions must
be made in a manner that enables the NRC to receive, read,
authenticate, distribute, and archive the submission, and process and
retrieve it a single page at a time. Detailed guidance on making
electronic submissions can be obtained by visiting the NRC's Web site
at http://www.nrc.gov/site-help/e-submittals.html, by calling (301)
415-0439, by e-mail to [email protected], or by writing the Office of
Information Services, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001. The guidance discusses, among other topics, the formats
the NRC can accept, the use of electronic signatures, and the treatment
of nonpublic information. If the communication is on paper, the signed
original must be sent. If a submission due date falls on a Saturday,
Sunday, or Federal holiday, the next Federal working day becomes the
official due date. The applicant shall maintain the capability to
generate additional copies of the environmental report or any
supplement to the environmental report for subsequent distribution to
parties and Boards in the NRC proceedings; Federal, State, and local
officials; and any affected Indian tribes, in accordance with written
instructions issued by the Director of the Office of New Reactors, the
Director of the Office of Nuclear Reactor Regulation, or the Director
of the Office of Nuclear Material Safety and Safeguards, as
appropriate.
(b) Each applicant for a license to manufacture a nuclear power
reactor, or for an amendment to a license to manufacture, seeking
approval of the final design of the nuclear power reactor under subpart
F of part 52 of this chapter, shall submit to the Commission an
environmental report or any supplement to an environmental report in
the manner specified in Sec. 50.3 of this chapter. The applicant shall
maintain the capability to generate additional copies of the
environmental report or any supplement to the environmental report for
subsequent distribution to parties and Boards in the NRC proceeding;
Federal, State, and local officials; and any affected Indian tribes, in
accordance with written instructions issued by the Director of the
Office of New Reactors or the Director of the Office of Nuclear Reactor
Regulation.
0
139. Section 51.55 is added to read as follows:
[[Page 49514]]
Sec. 51.55 Environmental report--standard design certification.
(a) Each applicant for a standard design certification under
subpart B of part 52 of this chapter shall submit with its application
a separate document entitled, ``Applicant's Environmental Report--
Standard Design Certification.'' The environmental report must address
the costs and benefits of severe accident mitigation design
alternatives, and the bases for not incorporating severe accident
mitigation design alternatives in the design to be certified.
(b) Each applicant for an amendment to a design certification shall
submit with its application a separate document entitled, ``Applicant's
Supplemental Environmental Report--Amendment to Standard Design
Certification.'' The environmental report must address whether the
design change which is the subject of the proposed amendment either
renders a severe accident mitigation design alternative previously
rejected in an environmental assessment to become cost beneficial, or
results in the identification of new severe accident mitigation design
alternatives that may be reasonably incorporated into the design
certification.
0
140. Section 51.66 is revised to read as follows:
Sec. 51.66 Environmental report--number of copies; distribution.
Each applicant for a license or other form of permission, or an
amendment to or renewal of a license or other form of permission issued
under parts 30, 32, 33, 34, 35, 36, 39, 40, 61, 70, and/or 72 of this
chapter, and covered by Sec. Sec. 51.60(b)(1) through (6); or by
Sec. Sec. 51.61 or 51.62 shall submit to the Director of Nuclear
Material Safety and Safeguards an environmental report or any
supplement to an environmental report in the manner specified in Sec.
51.58(a). The applicant shall maintain the capability to generate
additional copies of the environmental report or any supplement to the
environmental report for subsequent distribution to Federal, State, and
local officials, and any affected Indian tribes in accordance with
written instructions issued by the Director of Nuclear Material Safety
and Safeguards.
0
141. In Sec. 51.71 paragraph (d) and Footnote 3 are revised to read as
follows:
Sec. 51.71 Draft environmental impact statement--contents.
* * * * *
(d) Analysis. Unless excepted in this paragraph or Sec. 51.75, the
draft environmental impact statement will include a preliminary
analysis that considers and weighs the environmental effects of the
proposed action; the environmental impacts of alternatives to the
proposed action; and alternatives available for reducing or avoiding
adverse environmental effects and consideration of the economic,
technical, and other benefits and costs of the proposed action and
alternatives and indicate what other interests and considerations of
Federal policy, including factors not related to environmental quality
if applicable, are relevant to the consideration of environmental
effects of the proposed action identified under paragraph (a) of this
section. The draft supplemental environmental impact statement prepared
at the license renewal stage under Sec. 51.95(c) need not discuss the
economic or technical benefits and costs of either the proposed action
or alternatives except if benefits and costs are either essential for a
determination regarding the inclusion of an alternative in the range of
alternatives considered or relevant to mitigation. In addition, the
supplemental environmental impact statement prepared at the license
renewal stage need not discuss other issues not related to the
environmental effects of the proposed action and associated
alternatives. The draft supplemental environmental impact statement for
license renewal prepared under Sec. 51.95(c) will rely on conclusions
as amplified by the supporting information in the GEIS for issues
designated as Category 1 in appendix B to subpart A of this part. The
draft supplemental environmental impact statement must contain an
analysis of those issues identified as Category 2 in appendix B to
subpart A of this part that are open for the proposed action. The
analysis for all draft environmental impact statements will, to the
fullest extent practicable, quantify the various factors considered. To
the extent that there are important qualitative considerations or
factors that cannot be quantified, these considerations or factors will
be discussed in qualitative terms. Consideration will be given to
compliance with environmental quality standards and requirements that
have been imposed by Federal, State, regional, and local agencies
having responsibility for environmental protection, including
applicable zoning and land-use regulations and water pollution
limitations or requirements issued or imposed under the Federal Water
Pollution Control Act. The environmental impact of the proposed action
will be considered in the analysis with respect to matters covered by
environmental quality standards and requirements irrespective of
whether a certification or license from the appropriate authority has
been obtained.\3\ While satisfaction of Commission standards and
criteria pertaining to radiological effects will be necessary to meet
the licensing requirements of the Atomic Energy Act, the analysis will,
for the purposes of NEPA, consider the radiological effects of the
proposed action and alternatives.
---------------------------------------------------------------------------
\3\ Compliance with the environmental quality standards and
requirements of the Federal Water Pollution Control Act (imposed by
EPA or designated permitting states) is not a substitute for, and
does not negate the requirement for NRC to weigh all environmental
effects of the proposed action, including the degradation, if any,
of water quality, and to consider alternatives to the proposed
action that are available for reducing adverse effects. Where an
environmental assessment of aquatic impact from plant discharges is
available from the permitting authority, the NRC will consider the
assessment in its determination of the magnitude of environmental
impacts for striking an overall cost-benefit balance at the
construction permit and operating license and early site permit and
combined license stages, and in its determination of whether the
adverse environmental impacts of license renewal are so great that
preserving the option of license renewal for energy planning
decision-makers would be unreasonable at the license renewal stage.
When no such assessment of aquatic impacts is available from the
permitting authority, NRC will establish on its own, or in
conjunction with the permitting authority and other agencies having
relevant expertise, the magnitude of potential impacts for striking
an overall cost-benefit balance for the facility at the construction
permit and operating license and early site permit and combined
license stages, and in its determination of whether the adverse
environmental impacts of license renewal are so great that
preserving the option of license renewal for energy planning
decision-makers would be unreasonable at the license renewal stage.
---------------------------------------------------------------------------
* * * * *
0
142. Section 51.75 is revised to read as follows:
Sec. 51.75 Draft environmental impact statement--construction permit,
early site permit, or combined license.
(a) Construction permit stage. A draft environmental impact
statement relating to issuance of a construction permit for a
production or utilization facility will be prepared in accordance with
the procedures and measures described in Sec. Sec. 51.70, 51.71,
51.72, and 51.73. The contribution of the environmental effects of the
uranium fuel cycle activities specified in Sec. 51.51 shall be
evaluated on the basis of impact values set forth in Table S-3, Table
of Uranium Fuel Cycle Environmental Data, which shall be set out in the
draft environmental impact statement. With the exception of radon-222
and technetium-99 releases, no further discussion of fuel cycle release
values and other numerical data that appear explicitly in the table
shall be required.\5\
[[Page 49515]]
The impact statement shall take account of dose commitments and health
effects from fuel cycle effluents set forth in Table S-3 and shall in
addition take account of economic, socioeconomic, and possible
cumulative impacts and other fuel cycle impacts as may reasonably
appear significant.
(b) Early site permit stage. A draft environmental impact statement
relating to issuance of an early site permit for a production or
utilization facility will be prepared in accordance with the procedures
and measures described in Sec. Sec. 51.70, 51.71, 51.72, 51.73, and
this section. The contribution of the environmental effects of the
uranium fuel cycle activities specified in Sec. 51.51 shall be
evaluated on the basis of impact values set forth in Table S-3, Table
of Uranium Fuel Cycle Environmental Data, which shall be set out in the
draft environmental impact statement. With the exception of radon-222
and technetium-99 releases, no further discussion of fuel cycle release
values and other numerical data that appear explicitly in the table
shall be required.\5\ The impact statement shall take account of dose
commitments and health effects from fuel cycle effluents set forth in
Table S-3 and shall in addition take account of economic,
socioeconomic, and possible cumulative impacts and other fuel cycle
impacts as may reasonably appear significant. The draft environmental
impact statement must include an evaluation of alternative sites to
determine whether there is any obviously superior alternative to the
site proposed. The draft environmental impact statement must also
include an evaluation of the environmental effects of construction and
operation of a reactor, or reactors, which have design characteristics
that fall within the site characteristics and design parameters for the
early site permit application, but only to the extent addressed in the
early site permit environmental report or otherwise necessary to
determine whether there is any obviously superior alternative to the
site proposed. The draft environmental impact statement must not
include an assessment of the economic, technical, or other benefits
(for example, need for power) and costs of the proposed action or an
evaluation of alternative energy sources, unless these matters are
addressed in the early site permit environmental report.
---------------------------------------------------------------------------
\5\ Values for releases of Rn-222 and Tc-99 are not given in the
table. The amount and significance of Rn-222 releases from the fuel
cycle and Tc-99 releases from waste management or reprocessing
activities shall be considered in the draft environmental impact
statement and may be the subject of litigation in individual
licensing proceedings.
---------------------------------------------------------------------------
(c) Combined license stage. A draft environmental impact statement
relating to issuance of a combined license that does not reference an
early site permit will be prepared in accordance with the procedures
and measures described in Sec. Sec. 51.70, 51.71, 51.72, and 51.73.
The contribution of the environmental effects of the uranium fuel cycle
activities specified in Sec. 51.51 shall be evaluated on the basis of
impact values set forth in Table S-3, Table of Uranium Fuel Cycle
Environmental Data, which shall be set out in the draft environmental
impact statement. With the exception of radon-222 and technetium-99
releases, no further discussion of fuel cycle release values and other
numerical data that appear explicitly in the table shall be
required.\5\ The impact statement shall take account of dose
commitments and health effects from fuel cycle effluents set forth in
Table S-3 and shall in addition take account of economic,
socioeconomic, and possible cumulative impacts and other fuel cycle
impacts as may reasonably appear significant. The impact statement will
include a discussion of the storage of spent fuel for the nuclear power
plant within the scope of the generic determination in Sec. 51.23(a)
and in accordance with Sec. 51.23(b).
(1) Combined license application referencing an early site permit.
If the combined license application references an early site permit,
then the NRC staff shall prepare a draft supplement to the early site
permit environmental impact statement. The supplement must be prepared
in accordance with Sec. 51.92(e).
(2) Combined license application referencing a standard design
certification. If the combined license application references a
standard design certification and the site characteristics of the
combined license's site fall within the site parameters specified in
the design certification environmental assessment, then the draft
combined license environmental impact statement shall incorporate by
reference the design certification environmental assessment, and
summarize the findings and conclusions of the environmental assessment
with respect to severe accident mitigation design alternatives.
(3) Combined license application referencing a manufactured
reactor. If the combined license application proposes to use a
manufactured reactor and the site characteristics of the combined
license's site fall within the site parameters specified in the
manufacturing license environmental assessment, then the draft combined
license environmental impact statement shall incorporate by reference
the manufacturing license environmental assessment, and summarize the
findings and conclusions of the environmental assessment with respect
to severe accident mitigation design alternatives. The combined license
environmental impact statement report will not address the
environmental impacts associated with manufacturing the reactor under
the manufacturing license.
Sec. 51.76 [Removed]
0
143. Section 51.76 is removed and reserved.
0
144. Section 51.92 is revised to read as follows:
Sec. 51.92 Supplement to the final environmental impact statement.
(a) If the proposed action has not been taken, the NRC staff will
prepare a supplement to a final environmental impact statement for
which a notice of availability has been published in the Federal
Register as provided in Sec. 51.118, if:
(1) There are substantial changes in the proposed action that are
relevant to environmental concerns; or
(2) There are new and significant circumstances or information
relevant to environmental concerns and bearing on the proposed action
or its impacts.
(b) In a proceeding for a combined license application under 10 CFR
part 52 referencing an early site permit under part 52, the NRC staff
shall prepare a supplement to the final environmental impact statement
for the referenced early site permit in accordance with paragraph (e)
of this section.
(c) The NRC staff may prepare a supplement to a final environmental
impact statement when, in its opinion, preparation of a supplement will
further the purposes of NEPA.
(d) The supplement to a final environmental impact statement will
be prepared in the same manner as the final environmental impact
statement except that a scoping process need not be used.
(e) The supplement to an early site permit final environmental
impact statement which is prepared for a combined license application
in accordance with Sec. 51.75(c)(1) and paragraph (b) of this section
must:
(1) Identify the proposed action as the issuance of a combined
license for the construction and operation of a nuclear power plant as
described in the combined license application at the site described in
the early site permit referenced in the combined license application;
[[Page 49516]]
(2) Incorporate by reference the final environmental impact
statement prepared for the early site permit;
(3) Contain no separate discussion of alternative sites;
(4) Include an analysis of the economic, technical, and other
benefits and costs of the proposed action, to the extent that the final
environmental impact statement prepared for the early site permit did
not include an assessment of these benefits and costs;
(5) Include an analysis of other energy alternatives, to the extent
that the final environmental impact statement prepared for the early
site permit did not include an assessment of energy alternatives;
(6) Include an analysis of any environmental issue related to the
impacts of construction or operation of the facility that was not
resolved in the proceeding on the early site permit; and
(7) Include an analysis of the issues related to the impacts of
construction and operation of the facility that were resolved in the
early site permit proceeding for which new and significant information
has been identified, including, but not limited to, new and significant
information demonstrating that the design of the facility falls outside
the site characteristics and design parameters specified in the early
site permit.
(f)(1) A supplement to a final environmental impact statement will
be accompanied by or will include a request for comments as provided in
Sec. 51.73 and a notice of availability will be published in the
Federal Register as provided in Sec. 51.117 if paragraphs (a) or (b)
of this section applies.
(2) If comments are not requested, a notice of availability of a
supplement to a final environmental impact statement will be published
in the Federal Register as provided in Sec. 51.118.
0
145. In Sec. 51.95, paragraph (a), the introductory text of paragraph
(c), and paragraph (d) are revised to read as follows:
Sec. 51.95 Postconstruction environmental impact statements.
(a) General. Any supplement to a final environmental impact
statement or any environmental assessment prepared under the provisions
of this section may incorporate by reference any information contained
in a final environmental document previously prepared by the NRC staff
that relates to the same production or utilization facility. Documents
that may be referenced include, but are not limited to, the final
environmental impact statement; supplements to the final environmental
impact statement, including supplements prepared at the operating
license stage; NRC staff-prepared final generic environmental impact
statements; environmental assessments and records of decisions prepared
in connection with the construction permit, the operating license, the
early site permit, or the combined license and any license amendment
for that facility. A supplement to a final environmental impact
statement will include a request for comments as provided in Sec.
51.73.
* * * * *
(c) Operating license renewal stage. In connection with the renewal
of an operating license or combined license for a nuclear power plant
under parts 52 or 54 of this chapter, the Commission shall prepare an
environmental impact statement, which is a supplement to the
Commission's NUREG-1437, ``Generic Environmental Impact Statement for
License Renewal of Nuclear Plants'' (May 1996), which is available in
the NRC Public Document Room, 11555 Rockville Pike, Rockville,
Maryland.
* * * * *
(d) Postoperating license stage. In connection with the amendment
of an operating or combined license authorizing decommissioning
activities at a production or utilization facility covered by Sec.
51.20, either for unrestricted use or based on continuing use
restrictions applicable to the site, or with the issuance, amendment or
renewal of a license to store spent fuel at a nuclear power reactor
after expiration of the operating or combined license for the nuclear
power reactor, the NRC staff will prepare a supplemental environmental
impact statement for the post operating or post combined license stage
or an environmental assessment, as appropriate, which will update the
prior environmental documentation prepared by the NRC for compliance
with NEPA under the provisions of this part. The supplement or
assessment may incorporate by reference any information contained in
the final environmental impact statement--for the operating or combined
license stage, as appropriate, or in the records of decision prepared
in connection with the early site permit, construction permit,
operating license, or combined license for that facility. The
supplement will include a request for comments as provided in Sec.
51.73. Unless otherwise required by the Commission in accordance with
the generic determination in Sec. 51.23(a) and the provisions of Sec.
51.23(b), a supplemental environmental impact statement for the
postoperating or post combined license stage or an environmental
assessment, as appropriate, will address the environmental impacts of
spent fuel storage only for the term of the license, license amendment
or license renewal applied for.
0
146. Section 51.105 is revised to read as follows:
Sec. 51.105 Public hearings in proceedings for issuance of
construction permits or early site permits.
(a) In addition to complying with applicable requirements of Sec.
51.104, in a proceeding for the issuance of a construction permit or
early site permit for a nuclear power reactor, testing facility, fuel
reprocessing plant or isotopic enrichment plant, the presiding officer
will:
(1) Determine whether the requirements of Sections 102(2) (A), (C),
and (E) of NEPA and the regulations in this subpart have been met;
(2) Independently consider the final balance among conflicting
factors contained in the record of the proceeding with a view to
determining the appropriate action to be taken;
(3) Determine, after weighing the environmental, economic,
technical, and other benefits against environmental and other costs,
and considering reasonable alternatives, whether the construction
permit or early site permit should be issued, denied, or appropriately
conditioned to protect environmental values;
(4) Determine, in an uncontested proceeding, whether the NEPA
review conducted by the NRC staff has been adequate; and
(5) Determine, in a contested proceeding, whether in accordance
with the regulations in this subpart, the construction permit or early
site permit should be issued as proposed by the NRC's Director of New
Reactors or Director of Nuclear Reactor Regulation.
(b) The presiding officer in an early site permit hearing shall not
admit contentions proffered by any party concerning the benefits
assessment (e.g., need for power) or alternative energy sources if
those issues were not addressed by the applicant in the early site
permit application.
0
147. Section 51.105a is added to read as follows:
Sec. 51.105a Public hearings in proceedings for issuance of
manufacturing licenses.
In addition to complying with applicable requirements of Sec.
51.31(c), in a proceeding for the issuance of a manufacturing license,
the presiding officer will determine whether, in accordance with the
regulations in this subpart, the manufacturing license
[[Page 49517]]
should be issued as proposed by the NRC's Director of New Reactors or
Director of Nuclear Reactor Regulation.
0
148. Section 51.107 is added under the undesignated center heading
``Production and Utilization Facilities'' to read as follows:
Sec. 51.107 Public hearings in proceedings for issuance of combined
licenses.
(a) In addition to complying with the applicable requirements of
Sec. 51.104, in a proceeding for the issuance of a combined license
for a nuclear power reactor under part 52 of this chapter, the
presiding officer will:
(1) Determine whether the requirements of Sections 102(2) (A), (C),
and (E) of NEPA and the regulations in this subpart have been met;
(2) Independently consider the final balance among conflicting
factors contained in the record of the proceeding with a view to
determining the appropriate action to be taken;
(3) Determine, after weighing the environmental, economic,
technical, and other benefits against environmental and other costs,
and considering reasonable alternatives, whether the combined license
should be issued, denied, or appropriately conditioned to protect
environmental values;
(4) Determine, in an uncontested proceeding, whether the NEPA
review conducted by the NRC staff has been adequate; and
(5) Determine, in a contested proceeding, whether in accordance
with the regulations in this subpart, the combined license should be
issued as proposed by the NRC's Director of New Reactors or Director of
Nuclear Reactor Regulation.
(b) If a combined license application references an early site
permit, then the presiding officer in the combined license hearing
shall not admit any contention proffered by any party on environmental
issues which have been accorded finality under Sec. 52.39 of this
chapter, unless the contention:
(1) Demonstrates that the nuclear power reactor proposed to be
built does not fit within one or more of the site characteristics or
design parameters included in the early site permit;
(2) Raises any significant environmental issue that was not
resolved in the early site permit proceeding; or
(3) Raises any issue involving the impacts of construction and
operation of the facility that was resolved in the early site permit
proceeding for which new and significant information has been
identified.
(c) If the combined license application references a standard
design certification, or proposes to use a manufactured reactor, then
the presiding officer in a combined license hearing shall not admit
contentions proffered by any party concerning severe accident
mitigation design alternatives unless the contention demonstrates that
the site characteristics fall outside of the site parameters in the
standard design certification or underlying manufacturing license for
the manufactured reactor.
0
149. Section 51.108 is added under the undesignated center heading
``Production and Utilization Facilities,'' to read as follows:
Sec. 51.108 Public hearings on Commission findings that inspections,
tests, analyses, and acceptance criteria of combined licenses are met.
In any public hearing requested under 10 CFR 52.103(b), the
Commission will not admit any contentions on environmental issues, the
adequacy of the environmental impact statement for the combined license
issued under subpart C of part 52, or the adequacy of any other
environmental impact statement or environmental assessment referenced
in the combined license application. The Commission will not make any
environmental findings in connection with the finding under 10 CFR
52.103(g).
0
150. Part 52 is revised to read as follows:
PART 52--LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER
PLANTS
General Provisions
Sec.
52.0 Scope; applicability of 10 CFR Chapter I provisions.
52.1 Definitions.
52.2 Interpretations.
52.3 Written communications.
52.4 Deliberate misconduct.
52.5 Employee protection.
52.6 Completeness and accuracy of information.
52.7 Specific exemptions.
52.8 Combining licenses; elimination of repetition.
52.9 Jurisdictional limits.
52.10 Attacks and destructive acts.
52.11 Information collection requirements: OMB approval.
Subpart A--Early Site Permits
52.12 Scope of subpart.
52.13 Relationship to other subparts.
52.15 Filing of applications.
52.16 Contents of applications; general information.
52.17 Contents of applications; technical information.
52.18 Standards for review of applications.
52.21 Administrative review of applications; hearings.
52.23 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
52.24 Issuance of early site permit.
52.25 Extent of activities permitted.
52.27 Duration of permit.
52.28 Transfer of early site permit.
52.29 Application for renewal.
52.31 Criteria for renewal.
52.33 Duration of renewal.
52.35 Use of site for other purposes.
52.39 Finality of early site permit determinations.
Subpart B--Standard Design Certifications
52.41 Scope of subpart.
52.43 Relationship to other subparts.
52.45 Filing of applications.
52.46 Contents of applications; general information.
52.47 Contents of applications; technical information.
52.48 Standards for review of applications.
52.51 Administrative review of applications.
52.53 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
52.54 Issuance of standard design certification.
52.55 Duration of certification.
52.57 Application for renewal.
52.59 Criteria for renewal.
52.61 Duration of renewal.
52.63 Finality of standard design certifications.
Subpart C--Combined Licenses
52.71 Scope of subpart.
52.73 Relationship to other subparts.
52.75 Filing of applications.
52.77 Contents of applications; general information.
52.79 Contents of applications; technical information in final
safety analysis report.
52.80 Contents of applications; additional technical information.
52.81 Standards for review of applications.
52.83 Finality of referenced NRC approvals; partial initial decision
on site suitability.
52.85 Administrative review of applications; hearings.
52.87 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
52.89 Reserved.
52.91 Authorization to conduct site activities.
52.93 Exemptions and variances.
52.97 Issuance of combined licenses.
52.98 Finality of combined licenses; information requests.
52.99 Inspection during construction.
52.103 Operation under a combined license.
52.104 Duration of combined license.
52.105 Transfer of combined license.
52.107 Application for renewal.
52.109 Continuation of combined license.
52.110 Termination of license.
Subpart D--Reserved
Subpart E--Standard Design Approvals
52.131 Scope of subpart.
52.133 Relationship to other subparts.
52.135 Filing of applications.
[[Page 49518]]
52.136 Contents of applications; general information.
52.137 Contents of applications; technical information.
52.139 Standards for review of applications.
52.141 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
52.143 Staff approval of design.
52.145 Finality of standard design approvals; information requests.
52.147 Duration of design approval.
Subpart F--Manufacturing Licenses
52.151 Scope of subpart.
52.153 Relationship to other subparts.
52.155 Filing of applications.
52.156 Contents of applications; general information.
52.157 Contents of applications; technical information in final
safety analysis report.
52.158 Contents of application; additional technical information.
52.159 Standards for review of application.
52.161 Reserved.
52.163 Administrative review of applications; hearings.
52.165 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
52.167 Issuance of manufacturing license.
52.169 Reserved.
52.171 Finality of manufacturing licenses; information requests.
52.173 Duration of manufacturing license.
52.175 Transfer of manufacturing license.
52.177 Application for renewal.
52.179 Criteria for renewal.
52.181 Duration of renewal.
Subpart G--Reserved
Subpart H--Enforcement
52.301 Violations.
52.303 Criminal penalties.
Appendix A to Part 52--Design Certification Rule for the U.S.
Advanced Boiling Water Reactor
Appendix B to Part 52--Design Certification Rule for the System 80+
Design
Appendix C to Part 52--Design Certification Rule for the AP600
Design
Appendix D to Part 52--Design Certification Rule for the AP1000
Design
Appendixes E Through M to Part 52 [Reserved]
Appendix N to Part 52--Standardization of Nuclear Power Plant
Designs: Combined Licenses to Construct and Operate Nuclear Power
Reactors of Identical Design at Multiple Sites
Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat.
936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as
amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282); secs.
201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C.
5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
General Provisions
Sec. 52.0 Scope; applicability of 10 CFR Chapter I provisions.
(a) This part governs the issuance of early site permits, standard
design certifications, combined licenses, standard design approvals,
and manufacturing licenses for nuclear power facilities licensed under
Section 103 of the Atomic Energy Act of 1954, as amended (68 Stat.
919), and Title II of the Energy Reorganization Act of 1974 (88 Stat.
1242). This part also gives notice to all persons who knowingly provide
to any holder of or applicant for an approval, certification, permit,
or license, or to a contractor, subcontractor, or consultant of any of
them, components, equipment, materials, or other goods or services that
relate to the activities of a holder of or applicant for an approval,
certification, permit, or license, subject to this part, that they may
be individually subject to NRC enforcement action for violation of the
provisions in 10 CFR 52.4.
(b) Unless otherwise specifically provided for in this part, the
regulations in 10 CFR Chapter I apply to a holder of or applicant for
an approval, certification, permit, or license. A holder of or
applicant for an approval, certification, permit, or license issued
under this part shall comply with all requirements in 10 CFR Chapter I
that are applicable. A license, approval, certification, or permit
issued under this part is subject to all requirements in 10 CFR Chapter
I which, by their terms, are applicable to early site permits, design
certifications, combined licenses, design approvals, or manufacturing
licenses.
Sec. 52.1 Definitions.
(a) As used in this part--
Combined license means a combined construction permit and operating
license with conditions for a nuclear power facility issued under
subpart C of this part.
Decommission means to remove a facility or site safely from service
and reduce residual radioactivity to a level that permits--
(i) Release of the property for unrestricted use and termination of
the license; or
(ii) Release of the property under restricted conditions and
termination of the license.
Design characteristics are the actual features of a reactor or
reactors. Design characteristics are specified in a standard design
approval, a standard design certification, a combined license
application, or a manufacturing license.
Design parameters are the postulated features of a reactor or
reactors that could be built at a proposed site. Design parameters are
specified in an early site permit.
Early site permit means a Commission approval, issued under subpart
A of this part, for a site or sites for one or more nuclear power
facilities. An early site permit is a partial construction permit.
License means a license, including an early site permit, combined
license or manufacturing license under this part or a renewed license
issued by the Commission under this part or part 54 of this chapter.
Licensee means a person who is authorized to conduct activities
under a license issued by the Commission.
Major feature of the emergency plans means an aspect of those plans
necessary to:
(i) Address in whole or part one or more of the 16 standards in 10
CFR 50.47(b); or
(ii) Describe the emergency planning zones as required in 10 CFR
50.33(g).
Manufacturing license means a license, issued under subpart F of
this part, authorizing the manufacture of nuclear power reactors but
not their construction, installation, or operation at the sites on
which the reactors are to be operated.
Modular design means a nuclear power station that consists of two
or more essentially identical nuclear reactors (modules) and each
module is a separate nuclear reactor capable of being operated
independent of the state of completion or operating condition of any
other module co-located on the same site, even though the nuclear power
station may have some shared or common systems.
Prototype plant means a nuclear power plant that is used to test
new safety features, such as the testing required under 10 CFR
50.43(e). The prototype plant is similar to a first-of-a-kind or
standard plant design in all features and size, but may include
additional safety features to protect the public and the plant staff
from the possible consequences of accidents during the testing period.
Site characteristics are the actual physical, environmental and
demographic features of a site. Site characteristics are specified in
an early site permit or in a final safety analysis report for a
combined license.
Site parameters are the postulated physical, environmental and
demographic features of an assumed site. Site parameters are specified
in a standard design approval, standard design certification, or
manufacturing license.
Standard design means a design which is sufficiently detailed and
complete to support certification or approval in accordance with
subpart B or E of this part, and which is usable for a multiple number
of units or at a multiple number of sites without reopening or
repeating the review.
[[Page 49519]]
Standard design approval or design approval means an NRC staff
approval, issued under subpart E of this part, of a final standard
design for a nuclear power reactor of the type described in 10 CFR
50.22. The approval may be for either the final design for the entire
reactor facility or the final design of major portions thereof.
Standard design certification or design certification means a
Commission approval, issued under subpart B of this part, of a final
standard design for a nuclear power facility. This design may be
referred to as a certified standard design.
(b) All other terms in this part have the meaning set out in 10 CFR
50.2, or Section 11 of the Atomic Energy Act, as applicable.
Sec. 52.2 Interpretations.
Except as specifically authorized by the Commission in writing, no
interpretation of the meaning of the regulations in this part by any
officer or employee of the Commission other than a written
interpretation by the General Counsel will be recognized to be binding
upon the Commission.
Sec. 52.3 Written communications.
(a) General requirements. All correspondence, reports,
applications, and other written communications from an applicant,
licensee, or holder of a standard design approval to the Nuclear
Regulatory Commission concerning the regulations in this part,
individual license conditions, or the terms and conditions of an early
site permit or standard design approval, must be sent either by mail
addressed: ATTN: Document Control Desk, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001; by hand delivery to the NRC's
offices at 11555 Rockville Pike, Rockville, Maryland, between the hours
of 7:30 a.m. and 4:15 p.m. eastern time; or, where practicable, by
electronic submission, for example, via Electronic Information
Exchange, e-mail, or CD-ROM. Electronic submissions must be made in a
manner that enables the NRC to receive, read, authenticate, distribute,
and archive the submission, and process and retrieve it a single page
at a time. Detailed guidance on making electronic submissions can be
obtained by visiting the NRC's Web site at http://www.nrc.gov/site-help/eie.html, by calling (301) 415-6030, by e-mail at [email protected], or
by writing the Office of Information Services, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001. The guidance discusses, among
other topics, the formats the NRC can accept, the use of electronic
signatures, and the treatment of nonpublic information. If the
communication is on paper, the signed original must be sent. If a
submission due date falls on a Saturday, Sunday, or Federal holiday,
the next Federal working day becomes the official due date.
(b) Distribution requirements. Copies of all correspondence,
reports, and other written communications concerning the regulations in
this part or individual license conditions, or the terms and conditions
of an early site permit or standard design approval, must be submitted
to the persons listed in paragraph (b)(1) of this section (addresses
for the NRC Regional Offices are listed in appendix D to part 20 of
this chapter).
(1) Applications for amendment of permits and licenses; reports;
and other communications. All written communications (including
responses to: generic letters, bulletins, information notices,
regulatory information summaries, inspection reports, and miscellaneous
requests for additional information) that are required of holders of
early site permits, standard design approvals, combined licenses, or
manufacturing licenses issued under this part must be submitted as
follows, except as otherwise specified in paragraphs (b)(2) through
(b)(7) of this section: to the NRC's Document Control Desk (if on
paper, the signed original), with a copy to the appropriate Regional
Office, and a copy to the appropriate NRC Resident Inspector, if one
has been assigned to the site of the facility or the place of
manufacture of a reactor licensed under subpart F of this part.
(2) Applications and amendments to applications. Applications for
early site permits, standard design approvals, combined licenses,
manufacturing licenses and amendments to any of these types of
applications must be submitted to the NRC's Document Control Desk, with
a copy to the appropriate Regional Office, and a copy to the
appropriate NRC Resident Inspector, if one has been assigned to the
site of the facility or the place of manufacture of a reactor licensed
under subpart F of this part, except as otherwise specified in
paragraphs (b)(3) through (b)(7) of this section. If the application or
amendment is on paper, the submission to the Document Control Desk must
be the signed original.
(3) Acceptance review application. Written communications required
for an application for determination of suitability for docketing must
be submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
(4) Security plan and related submissions. Written communications,
as defined in paragraphs (b)(4)(i) through (iv) of this section, must
be submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office. If the communication is on paper, the
submission to the Document Control Desk must be the signed original.
(i) Physical security plan under Sec. 52.79 of this chapter;
(ii) Safeguards contingency plan under Sec. 52.79 of this chapter;
(iii) Change to security plan, guard training and qualification
plan, or safeguards contingency plan made without prior Commission
approval under Sec. 50.54(p) of this chapter;
(iv) Application for amendment of physical security plan, guard
training and qualification plan, or safeguards contingency plan under
Sec. 50.90 of this chapter.
(5) Emergency plan and related submissions. Written communications
as defined in paragraphs (b)(5)(i) through (iii) of this section must
be submitted to the NRC's Document Control Desk, with a copy to the
appropriate Regional Office, and a copy to the appropriate NRC Resident
Inspector if one has been assigned to the site of the facility. If the
communication is on paper, the submission to the Document Control Desk
must be the signed original.
(i) Emergency plan under Sec. 52.17(b) or Sec. 52.79(a);
(ii) Change to an emergency plan under Sec. 50.54(q) of this
chapter;
(iii) Emergency implementing procedures under appendix E, Section V
of part 50 of this chapter.
(6) Updated FSAR. An updated final safety analysis report (FSAR) or
replacement pages under Sec. 50.71(e) of this chapter, or the
regulations in this part must be submitted to the NRC's Document
Control Desk, with a copy to the appropriate Regional Office, and a
copy to the appropriate NRC Resident Inspector if one has been assigned
to the site of the facility or the place of manufacture of a reactor
licensed under subpart F of this part. Paper copy submissions may be
made using replacement pages; however, if a licensee chooses to use
electronic submission, all subsequent updates or submissions must be
performed electronically on a total replacement basis. If the
communication is on paper, the submission to the Document Control Desk
must be the signed original. If the communications are submitted
electronically, see Guidance for
[[Page 49520]]
Electronic Submissions to the Commission.
(7) Quality assurance related submissions. (i) A change to the
safety analysis report quality assurance program description under
Sec. 50.54(a)(3) or Sec. 50.55(f)(4) of this chapter, or a change to
a licensee's NRC-accepted quality assurance topical report under Sec.
50.54(a)(3) or Sec. 50.55(f)(4) of this chapter, must be submitted to
the NRC's Document Control Desk, with a copy to the appropriate
Regional Office, and a copy to the appropriate NRC Resident Inspector
if one has been assigned to the site of the facility. If the
communication is on paper, the submission to the Document Control Desk
must be the signed original.
(ii) A change to an NRC-accepted quality assurance topical report
from nonlicensees (i.e., architect/engineers, NSSS suppliers, fuel
suppliers, constructors, etc.) must be submitted to the NRC's Document
Control Desk. If the communication is on paper, the signed original
must be sent.
(8) Certification of permanent cessation of operations. The
licensee's certification of permanent cessation of operations under
Sec. 52.110(a)(1), must state the date on which operations have ceased
or will cease, and must be submitted to the NRC's Document Control
Desk. This submission must be under oath or affirmation.
(9) Certification of permanent fuel removal. The licensee's
certification of permanent fuel removal under Sec. 52.110(a)(1), must
state the date on which the fuel was removed from the reactor vessel
and the disposition of the fuel, and must be submitted to the NRC's
Document Control Desk. This submission must be under oath or
affirmation.
(c) Form of communications. All paper copies submitted to meet the
requirements set forth in paragraph (b) of this section must be
typewritten, printed or otherwise reproduced in permanent form on
unglazed paper. Exceptions to these requirements imposed on paper
submissions may be granted for the submission of micrographic,
photographic, or similar forms.
(d) Regulation governing submission. Applicants, licensees, and
holders of standard design approvals submitting correspondence,
reports, and other written communications under the regulations of this
part are requested but not required to cite whenever practical, in the
upper right corner of the first page of the submission, the specific
regulation or other basis requiring submission.
Sec. 52.4 Deliberate misconduct.
(a) Applicability. This section applies to any:
(1) Licensee;
(2) Holder of a standard design approval;
(3) Applicant for a standard design certification;
(4) Applicant for a license or permit;
(5) Applicant for a standard design approval;
(6) Employee of a licensee;
(7) Employee of an applicant for a license, a standard design
certification, or a standard design approval;
(8) Any contractor (including a supplier or consultant),
subcontractor, or employee of a contractor or subcontractor of any
licensee; or
(9) Any contractor (including a supplier or consultant),
subcontractor, or employee of a contractor or subcontractor of any
applicant for a license, a standard design certification, or a standard
design approval.
(b) Definitions. For purposes of this section:
Deliberate misconduct means an intentional act or omission that a
person or entity knows:
(i) Would cause a licensee or an applicant for a license, standard
design certification, or standard design approval to be in violation of
any rule, regulation, or order; or any term, condition, or limitation,
of any license, standard design certification, or standard design
approval; or
(ii) Constitutes a violation of a requirement, procedure,
instruction, contract, purchase order, or policy of a licensee, holder
of a standard design approval, applicant for a license, standard design
certification, or standard design approval, or contractor, or
subcontractor.
(c) Prohibition against deliberate misconduct. Any person or entity
subject to this section, who knowingly provides to any licensee, any
applicant for a license, standard design certification or standard
design approval, or a contractor, or subcontractor of a person or
entity subject to this section, any components, equipment, materials,
or other goods or services that relate to a licensee's or applicant's
activities under this part, may not:
(1) Engage in deliberate misconduct that causes or would have
caused, if not detected, a licensee, holder of a standard design
approval, or applicant to be in violation of any rule, regulation, or
order; or any term, condition, or limitation of any license issued by
the Commission, any standard design approval, or standard design
certification; or
(2) Deliberately submit to the NRC; a licensee, an applicant for a
license, standard design certification or standard design approval; or
a licensee's, standard design approval holder's, or applicant's
contractor or subcontractor, information that the person submitting the
information knows to be incomplete or inaccurate in some respect
material to the NRC.
(d) A person or entity who violates paragraph (c)(1) or (c)(2) of
this section may be subject to enforcement action in accordance with
the procedures in 10 CFR part 2, subpart B.
Sec. 52.5 Employee protection.
(a) Discrimination by a Commission licensee, holder of a standard
design approval, an applicant for a license, standard design
certification, or standard design approval, a contractor or
subcontractor of a Commission licensee, holder of a standard design
approval, applicant for a license, standard design certification, or
standard design approval, against an employee for engaging in certain
protected activities is prohibited. Discrimination includes discharge
and other actions that relate to compensation, terms, conditions, or
privileges of employment. The protected activities are established in
Section 211 of the Energy Reorganization Act of 1974, as amended, and
in general are related to the administration or enforcement of a
requirement imposed under the Atomic Energy Act or the Energy
Reorganization Act.
(1) The protected activities include but are not limited to:
(i) Providing the Commission or his or her employer information
about alleged violations of either of the statutes named in the
introductory text of paragraph (a) of this section or possible
violations of requirements imposed under either of those statutes;
(ii) Refusing to engage in any practice made unlawful under either
of the statutes named in the introductory text of paragraph (a) of this
section or under these requirements if the employee has identified the
alleged illegality to the employer;
(iii) Requesting the Commission to institute action against his or
her employer for the administration or enforcement of these
requirements;
(iv) Testifying in any Commission proceeding, or before Congress,
or at any Federal or State proceeding regarding any provision (or
proposed provision) of either of the statutes named in the introductory
text of paragraph (a) of this section; and
[[Page 49521]]
(v) Assisting or participating in, or is about to assist or
participate in, these activities.
(2) These activities are protected even if no formal proceeding is
actually initiated as a result of the employee assistance or
participation.
(3) This section has no application to any employee alleging
discrimination prohibited by this section who, acting without direction
from his or her employer (or the employer's agent), deliberately causes
a violation of any requirement of the Energy Reorganization Act of
1974, as amended, or the Atomic Energy Act of 1954, as amended.
(b) Any employee who believes that he or she has been discharged or
otherwise discriminated against by any person for engaging in protected
activities specified in paragraph (a)(1) of this section may seek a
remedy for the discharge or discrimination through an administrative
proceeding in the Department of Labor. The administrative proceeding
must be initiated within 180 days after an alleged violation occurs.
The employee may do this by filing a complaint alleging the violation
with the Department of Labor, Employment Standards Administration, Wage
and Hour Division. The Department of Labor may order reinstatement,
back pay, and compensatory damages.
(c) A violation of paragraph (a), (e), or (f) of this section by a
Commission licensee, a holder of a standard design approval, an
applicant for a Commission license, standard design certification, or a
standard design approval, or a contractor or subcontractor of a
Commission licensee, holder of a standard design approval, or any
applicant may be grounds for--
(1) Denial, revocation, or suspension of the license or standard
design approval;
(2) Withdrawal or revocation of a proposed or final standard design
certification;
(3) Imposition of a civil penalty on the licensee, holder of a
standard design approval, or applicant (including an applicant for a
standard design certification under this part following Commission
adoption of final design certification rule).
(4) Other enforcement action.
(d) Actions taken by an employer, or others, which adversely affect
an employee may be predicated upon nondiscriminatory grounds. The
prohibition applies when the adverse action occurs because the employee
has engaged in protected activities. An employee's engagement in
protected activities does not automatically render him or her immune
from discharge or discipline for legitimate reasons or from adverse
action dictated by nonprohibited considerations.
(e)(1) Each licensee, each holder of a standard design approval,
and each applicant for a license, standard design certification, or
standard design approval, shall prominently post the revision of NRC
Form 3, ``Notice to Employees,'' referenced in 10 CFR 19.11(e). This
form must be posted at locations sufficient to permit employees
protected by this section to observe a copy on the way to or from their
place of work. Premises must be posted not later than thirty (30) days
after an application is docketed and remain posted while the
application is pending before the Commission, during the term of the
license, standard design certification, or standard design approval
under 10 CFR part 52, and for 30 days following license termination or
the expiration or termination of the standard design certification or
standard design approval under 10 CFR part 52.
(2) Copies of NRC Form 3 may be obtained by writing to the Regional
Administrator of the appropriate U.S. Nuclear Regulatory Commission
Regional Office listed in appendix D to part 20 of this chapter, by
calling (301) 415-7232, via e-mail to [email protected], or by visiting the
NRC's Web site at http://www.nrc.gov and selecting forms from the index
found on the NRC's home page.
(f) No agreement affecting the compensation, terms, conditions, or
privileges of employment, including an agreement to settle a complaint
filed by an employee with the Department of Labor under Section 211 of
the Energy Reorganization Act of 1974, as amended, may contain any
provision which would prohibit, restrict, or otherwise discourage an
employee from participating in protected activity as defined in
paragraph (a)(1) of this section including, but not limited to,
providing information to the NRC or to his or her employer on potential
violations or other matters within NRC's regulatory responsibilities.
(g) Part 19 of this chapter sets forth requirements and regulatory
provisions applicable to licensees, holders of a standard design
approval, applicants for a license, standard design certification, or
standard design approval, and contractors or subcontractors of a
Commission licensee, or holder of a standard design approval, and are
in addition to the requirements in this section.
Sec. 52.6 Completeness and accuracy of information.
(a) Information provided to the Commission by a licensee (including
an early site permit holder, a combined license holder, and a
manufacturing license holder), a holder of a standard design approval
under this part, and an applicant for a license or an applicant for a
standard design certification or a standard design approval under this
part, and information required by statute or by the Commission's
regulations, orders, license conditions, or terms and conditions of a
standard design approval to be maintained by the licensee, the holder
of a standard design approval under this part, the applicant for a
standard design certification under this part following Commission
adoption of a final design certification rule, and an applicant for a
license, a standard design certification, or a standard design approval
under this part shall be complete and accurate in all material
respects.
(b) Each applicant or licensee, each holder of a standard design
approval under this part, and each applicant for a standard design
certification under this part following Commission adoption of a final
design certification regulation, shall notify the Commission of
information identified by the applicant or the licensee as having for
the regulated activity a significant implication for public health and
safety or common defense and security. An applicant, licensee, or
holder violates this paragraph only if the applicant, licensee, or
holder fails to notify the Commission of information that the
applicant, licensee, or holder has been identified as having a
significant implication for public health and safety or common defense
and security. Notification shall be provided to the Administrator of
the appropriate Regional Office within 2 working days of identifying
the information. This requirement is not applicable to information
which is already required to be provided to the Commission by other
reporting or updating requirements.
Sec. 52.7 Specific exemptions.
The Commission may, upon application by any interested person or
upon its own initiative, grant exemptions from the requirements of the
regulations of this part. The Commission's consideration will be
governed by Sec. 50.12 of this chapter, unless other criteria are
provided for in this part, in which case the Commission's consideration
will be governed by the criteria in this part. Only if those criteria
are not met will the Commission's consideration be
[[Page 49522]]
governed by Sec. 50.12 of this chapter. The Commission's consideration
of requests for exemptions from requirements of the regulations of
other parts in this chapter, which are applicable by virtue of this
part, shall be governed by the exemption requirements of those parts.
Sec. 52.8 Combining licenses; elimination of repetition.
(a) An applicant for a license under this part may combine in its
application several applications for different kinds of licenses under
the regulations of this chapter.
(b) An applicant may incorporate by reference in its application
information contained in previous applications, statements or reports
filed with the Commission, provided, however, that such references are
clear and specific.
(c) The Commission may combine in a single license the activities
of an applicant which would otherwise be licensed separately.
Sec. 52.9 Jurisdictional limits.
No permit, license, standard design approval, or standard design
certification under this part shall be deemed to have been issued for
activities which are not under or within the jurisdiction of the United
States.
Sec. 52.10 Attacks and destructive acts.
Neither an applicant for a license to manufacture, construct, and
operate a utilization facility under this part, nor for an amendment to
this license, or an applicant for an early site permit, a standard
design certification, or standard design approval under this part, or
for an amendment to the early site permit, standard design
certification, or standard design approval, is required to provide for
design features or other measures for the specific purpose of
protection against the effects of--
(a) Attacks and destructive acts, including sabotage, directed
against the facility by an enemy of the United States, whether a
foreign government or other person; or
(b) Use or deployment of weapons incident to U.S. defense
activities.
Sec. 52.11 Information collection requirements: OMB approval.
(a) The Nuclear Regulatory Commission has submitted the information
collection requirements contained in this part to the Office of
Management and Budget (OMB) for approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or
sponsor, and a person is not required to respond to, a collection of
information unless it displays a currently valid OMB control number.
OMB has approved the information collection requirements contained in
this part under Control Number 3150-0151.
(b) The approved information collection requirements contained in
this part appear in Sec. Sec. 52.7, 52.15, 52.16, 52.17, 52.29, 52.35,
52.39, 52.45, 52.46, 52.47, 52.57, 52.63, 52.75, 52.77, 52.79, 52.80,
52.93, 52.99, 52.110, 52.135, 52.136, 52.137, 52.155, 52.156, 52.157,
52.158, 52.171, 52.177, and appendices A, B, C, D, and N of part 52.
Subpart A--Early Site Permits
Sec. 52.12 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of an early site permit for approval of a site for
one or more nuclear power facilities separate from the filing of an
application for a construction permit or combined license for the
facility.
Sec. 52.13 Relationship to other subparts.
This subpart applies when any person who may apply for a
construction permit under 10 CFR part 50, or for a combined license
under this part seeks an early site permit from the Commission
separately from an application for a construction permit or a combined
license.
Sec. 52.15 Filing of applications.
(a) Any person who may apply for a construction permit under 10 CFR
part 50, or for a combined license under this part, may file an
application for an early site permit with the Director, Office of New
Reactors, or the Director, Office of Nuclear Reactor Regulation, as
appropriate. An application for an early site permit may be filed
notwithstanding the fact that an application for a construction permit
or a combined license has not been filed in connection with the site
for which a permit is sought.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and 50.30 of this chapter.
(c) The fees associated with the filing and review of an
application for the initial issuance or renewal of an early site permit
are set forth in 10 CFR part 170.
Sec. 52.16 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (d) and (j) of this chapter.
Sec. 52.17 Contents of applications; technical information.
(a) For applications submitted before September 27, 2007, the rule
provisions in effect at the date of docketing apply unless otherwise
requested by the applicant in writing. The application must contain:
(1) A site safety analysis report. The site safety analysis report
shall include the following:
(i) The specific number, type, and thermal power level of the
facilities, or range of possible facilities, for which the site may be
used;
(ii) The anticipated maximum levels of radiological and thermal
effluents each facility will produce;
(iii) The type of cooling systems, intakes, and outflows that may
be associated with each facility;
(iv) The boundaries of the site;
(v) The proposed general location of each facility on the site;
(vi) The seismic, meteorological, hydrologic, and geologic
characteristics of the proposed site with appropriate consideration of
the most severe of the natural phenomena that have been historically
reported for the site and surrounding area and with sufficient margin
for the limited accuracy, quantity, and period of time in which the
historical data have been accumulated;
(vii) The location and description of any nearby industrial,
military, or transportation facilities and routes;
(viii) The existing and projected future population profile of the
area surrounding the site;
(ix) A description and safety assessment of the site on which a
facility is to be located. The assessment must contain an analysis and
evaluation of the major structures, systems, and components of the
facility that bear significantly on the acceptability of the site under
the radiological consequence evaluation factors identified in
paragraphs (a)(1)(ix)(A) and (a)(1)(ix)(B) of this section. In
performing this assessment, an applicant shall assume a fission product
release \1\ from the core into the containment assuming that the
facility is operated at the ultimate power level contemplated. The
applicant shall perform an evaluation and analysis of the postulated
fission product release, using the expected demonstrable
[[Page 49523]]
containment leak rate and any fission product cleanup systems intended
to mitigate the consequences of the accidents, together with applicable
site characteristics, including site meteorology, to evaluate the
offsite radiological consequences. Site characteristics must comply
with part 100 of this chapter. The evaluation must determine that:
---------------------------------------------------------------------------
\1\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events. Such accidents have generally been assumed to
result in substantial meltdown of the core with subsequent release
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(A) An individual located at any point on the boundary of the
exclusion area for any 2 hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \2\ total effective dose equivalent (TEDE).
---------------------------------------------------------------------------
\2\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose
for radiation workers which, according to NCRP recommendations at
the time could be disregarded in the determination of their
radiation exposure status (see NBS Handbook 69 dated June 5, 1959).
However, its use is not intended to imply that this number
constitutes an acceptable limit for an emergency dose to the public
under accident conditions. Rather, this dose value has been set
forth in this section as a reference value, which can be used in the
evaluation of plant design features with respect to postulated
reactor accidents, to assure that these designs provide assurance of
low risk of public exposure to radiation, in the event of an
accidents.
---------------------------------------------------------------------------
(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
TEDE;
(x) Information demonstrating that site characteristics are such
that adequate security plans and measures can be developed;
(xi) For applications submitted after September 27, 2007, a
description of the quality assurance program applied to site-related
activities for the future design, fabrication, construction, and
testing of the structures, systems, and components of a facility or
facilities that may be constructed on the site. Appendix B to 10 CFR
part 50 sets forth the requirements for quality assurance programs for
nuclear power plants. The description of the quality assurance program
for a nuclear power plant site shall include a discussion of how the
applicable requirements of appendix B to part 50 of this chapter will
be satisfied; and
(xii) An evaluation of the site against applicable sections of the
Standard Review Plan (SRP) revision in effect 6 months before the
docket date of the application. The evaluation required by this section
shall include an identification and description of all differences in
analytical techniques and procedural measures proposed for a site and
those corresponding techniques and measures given in the SRP acceptance
criteria. Where such a difference exists, the evaluation shall discuss
how the proposed alternative provides an acceptable method of complying
with the Commission's regulations, or portions thereof, that underlie
the corresponding SRP acceptance criteria. The SRP is not a substitute
for the regulations, and compliance is not a requirement.
(2) A complete environmental report as required by 10 CFR 51.50(b).
(b)(1) The site safety analysis report must identify physical
characteristics of the proposed site, such as egress limitations from
the area surrounding the site, that could pose a significant impediment
to the development of emergency plans. If physical characteristics are
identified that could pose a significant impediment to the development
of emergency plans, the application must identify measures that would,
when implemented, mitigate or eliminate the significant impediment.
(2) The site safety analysis report may also:
(i) Propose major features of the emergency plans, in accordance
with the pertinent standards of 10 CFR 50.47, and the requirements of
appendix E to 10 CFR part 50, such as the exact size and configuration
of the emergency planning zones, for review and approval by NRC, in
consultation with the Department of Homeland Security (DHS) in the
absence of complete and integrated emergency plans; or
(ii) Propose complete and integrated emergency plans for review and
approval by the NRC, in consultation with DHS, in accordance with the
applicable standards of 10 CFR 50.47, and the requirements of appendix
E to 10 CFR part 50. To the extent approval of emergency plans is
sought, the application must contain the information required by
Sec. Sec. 50.33(g) and (j) of this chapter.
(3) Emergency plans submitted under paragraph (b)(2)(ii) of this
section must include the proposed inspections, tests, and analyses that
the holder of a combined license referencing the early site permit
shall perform, and the acceptance criteria that are necessary and
sufficient to provide reasonable assurance that, if the inspections,
tests, and analyses are performed and the acceptance criteria met, the
facility has been constructed and will be operated in conformity with
the emergency plans, the provisions of the Act, and the Commission's
rules and regulations. Major features of an emergency plan submitted
under paragraph (b)(2)(i) of this section may include proposed
inspections, tests, analyses, and acceptance criteria.
(4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the site
safety analysis report must include a description of contacts and
arrangements made with Federal, State, and local governmental agencies
with emergency planning responsibilities. The site safety analysis
report must contain any certifications that have been obtained. If
these certifications cannot be obtained, the site safety analysis
report must contain information, including a utility plan, sufficient
to show that the proposed plans provide reasonable assurance that
adequate protective measures can and will be taken in the event of a
radiological emergency at the site. Under the option set forth in
paragraph (b)(2)(ii) of this section, the applicant shall make good
faith efforts to obtain from the same governmental agencies
certifications that:
(i) The proposed emergency plans are practicable;
(ii) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations,
and
(iii) That these agencies are committed to executing their
responsibilities under the plans in the event of an emergency.
(c) If the applicant requests authorization to perform activities
at the site, which are identified in 10 CFR 50.10(e)(1), after issuance
of the early site permit and without a separate authorization under 10
CFR 50.10(e)(1), the applicant must identify the activities that are
requested, and propose a plan for redress of the site in the event that
the activities are performed and the early site permit expires before
it is referenced in an application for a construction permit or a
combined license. The application must demonstrate that there is
reasonable assurance that redress carried out under the plan will
achieve an environmentally stable and aesthetically acceptable site
suitable for whatever non-nuclear use may conform with local zoning
laws.
Sec. 52.18 Standards for review of applications.
Applications filed under this subpart will be reviewed according to
the applicable standards set out in 10 CFR part 50 and its appendices
and 10 CFR part 100. In addition, the Commission shall prepare an
environmental impact statement during review of the application, in
accordance with the applicable provisions of 10 CFR part 51. The
Commission shall determine, after
[[Page 49524]]
consultation with DHS, whether the information required of the
applicant by Sec. 52.17(b)(1) shows that there is no significant
impediment to the development of emergency plans that cannot be
mitigated or eliminated by measures proposed by the applicant, whether
any major features of emergency plans submitted by the applicant under
Sec. 52.17(b)(2)(i) are acceptable in accordance with the applicable
standards of 10 CFR 50.47 and the requirements of appendix E to 10 CFR
part 50, and whether any emergency plans submitted by the applicant
under Sec. 52.17(b)(2)(ii) provide reasonable assurance that adequate
protective measures can and will be taken in the event of a
radiological emergency.
Sec. 52.21 Administrative review of applications; hearings.
An early site permit is subject to all procedural requirements in
10 CFR part 2, including the requirements for docketing in Sec.
2.101(a)(1) through (4) of this chapter, and the requirements for
issuance of a notice of hearing in Sec. Sec. 2.104(a) and (d) of this
chapter, provided that the designated sections may not be construed to
require that the environmental report, or draft or final environmental
impact statement include an assessment of the benefits of construction
and operation of the reactor or reactors, or an analysis of alternative
energy sources. The presiding officer in an early site permit hearing
shall not admit contentions proffered by any party concerning an
assessment of the benefits of construction and operation of the reactor
or reactors, or an analysis of alternative energy sources if those
issues were not addressed by the applicant in the early site permit
application. All hearings conducted on applications for early site
permits filed under this part are governed by the procedures contained
in subparts C, G, L, and N of 10 CFR part 2, as applicable.
Sec. 52.23 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
The Commission shall refer a copy of the application for an early
site permit to the ACRS. The ACRS shall report on those portions of the
application which concern safety.
Sec. 52.24 Issuance of early site permit.
(a) After conducting a hearing under Sec. 52.21 and receiving the
report to be submitted by the ACRS under Sec. 52.23, the Commission
may issue an early site permit, in the form the Commission deems
appropriate, if the Commission finds that:
(1) An application for an early site permit meets the applicable
standards and requirements of the Act and the Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the site is in conformity
with the provisions of the Act, and the Commission's regulations;
(4) The applicant is technically qualified to engage in any
activities authorized;
(5) The proposed inspections, tests, analyses and acceptance
criteria, including any on emergency planning, are necessary and
sufficient, within the scope of the early site permit, to provide
reasonable assurance that the facility has been constructed and will be
operated in conformity with the license, the provisions of the Act, and
the Commission's regulations;
(6) Issuance of the permit will not be inimical to the common
defense and security or to the health and safety of the public;
(7) Any significant adverse environmental impact resulting from
activities requested under Sec. 52.17(c) can be redressed; and
(8) The findings required by subpart A of 10 CFR part 51 have been
made.
(b) The early site permit must specify the site characteristics,
design parameters, and terms and conditions of the early site permit
the Commission deems appropriate. Before issuance of either a
construction permit or combined license referencing an early site
permit, the Commission shall find that any relevant terms and
conditions of the early site permit have been met. Any terms or
conditions of the early site permit that could not be met by the time
of issuance of the construction permit or combined license, must be set
forth as terms or conditions of the construction permit or combined
license.
(c) The early site permit shall specify the activities under Sec.
52.17(c) that the permit holder is authorized to perform.
Sec. 52.25 Extent of activities permitted.
If the activities authorized by Sec. 52.24(c) are performed and
the site is not referenced in an application for a construction permit
or a combined license issued under subpart C of this part while the
permit remains valid, then the early site permit remains in effect
solely for the purpose of site redress, and the holder of the permit
shall redress the site in accordance with the terms of the site redress
plan required by Sec. 52.17(c). If, before redress is complete, a use
not envisaged in the redress plan is found for the site or parts
thereof, the holder of the permit shall carry out the redress plan to
the greatest extent possible consistent with the alternate use.
Sec. 52.27 Duration of permit.
(a) Except as provided in paragraph (b) of this section, an early
site permit issued under this subpart may be valid for not less than
10, nor more than 20 years from the date of issuance.
(b) An early site permit continues to be valid beyond the date of
expiration in any proceeding on a construction permit application or a
combined license application that references the early site permit and
is docketed before the date of expiration of the early site permit, or,
if a timely application for renewal of the permit has been docketed,
before the Commission has determined whether to renew the permit.
(c) An applicant for a construction permit or combined license may,
at its own risk, reference in its application a site for which an early
site permit application has been docketed but not granted.
(d) Upon issuance of a construction permit or combined license, a
referenced early site permit is subsumed, to the extent referenced,
into the construction permit or combined license.
Sec. 52.28 Transfer of early site permit.
An application to transfer an early site permit will be processed
under 10 CFR 50.80.
Sec. 52.29 Application for renewal.
(a) Not less than 12, nor more than 36 months before the expiration
date stated in the early site permit, or any later renewal period, the
permit holder may apply for a renewal of the permit. An application for
renewal must contain all information necessary to bring up to date the
information and data contained in the previous application.
(b) Any person whose interests may be affected by renewal of the
permit may request a hearing on the application for renewal. The
request for a hearing must comply with 10 CFR 2.309. If a hearing is
granted, notice of the hearing will be published in accordance with 10
CFR 2.309.
(c) An early site permit, either original or renewed, for which a
timely application for renewal has been filed, remains in effect until
the Commission has determined whether to renew the permit. If the
permit is not renewed, it continues to be valid in certain proceedings
in accordance with the provisions of Sec. 52.27(b).
(d) The Commission shall refer a copy of the application for
renewal to the
[[Page 49525]]
ACRS. The ACRS shall report on those portions of the application which
concern safety and shall apply the criteria set forth in Sec. 52.31.
Sec. 52.31 Criteria for renewal.
(a) The Commission shall grant the renewal if it determines that:
(1) The site complies with the Act, the Commission's regulations,
and orders applicable and in effect at the time the site permit was
originally issued; and
(2) Any new requirements the Commission may wish to impose are:
(i) Necessary for adequate protection to public health and safety
or common defense and security;
(ii) Necessary for compliance with the Commission's regulations,
and orders applicable and in effect at the time the site permit was
originally issued; or
(iii) A substantial increase in overall protection of the public
health and safety or the common defense and security to be derived from
the new requirements, and the direct and indirect costs of
implementation of those requirements are justified in view of this
increased protection.
(b) A denial of renewal for failure to comply with the provisions
of Sec. 52.31(a) does not bar the permit holder or another applicant
from filing a new application for the site which proposes changes to
the site or the way that it is used to correct the deficiencies cited
in the denial of the renewal.
Sec. 52.33 Duration of renewal.
Each renewal of an early site permit may be for not less than 10,
nor more than 20 years, plus any remaining years on the early site
permit then in effect before renewal.
Sec. 52.35 Use of site for other purposes.
A site for which an early site permit has been issued under this
subpart may be used for purposes other than those described in the
permit, including the location of other types of energy facilities. The
permit holder shall inform the Director of New Reactors or the Director
of Nuclear Reactor Regulation, as appropriate, (Director) of any
significant uses for the site which have not been approved in the early
site permit. The information about the activities must be given to the
Director at least 30 days in advance of any actual construction or site
modification for the activities. The information provided could be the
basis for imposing new requirements on the permit, in accordance with
the provisions of Sec. 52.39. If the permit holder informs the
Director that the holder no longer intends to use the site for a
nuclear power plant, the Director may terminate the permit.
Sec. 52.39 Finality of early site permit determinations.
(a) Commission finality. (1) Notwithstanding any provision in 10
CFR 50.109, while an early site permit is in effect under Sec. Sec.
52.27 or 52.33, the Commission may not change or impose new site
characteristics, design parameters, or terms and conditions, including
emergency planning requirements, on the early site permit unless the
Commission:
(i) Determines that a modification is necessary to bring the permit
or the site into compliance with the Commission's regulations and
orders applicable and in effect at the time the permit was issued;
(ii) Determines the modification is necessary to assure adequate
protection of the public health and safety or the common defense and
security;
(iii) Determines that a modification is necessary based on an
update under paragraph (b) of this section; or
(iv) Issues a variance requested under paragraph (d) of this
section.
(2) In making the findings required for issuance of a construction
permit or combined license, or the findings required by Sec. 52.103,
or in any enforcement hearing other than one initiated by the
Commission under paragraph (a)(1) of this section, if the application
for the construction permit or combined license references an early
site permit, the Commission shall treat as resolved those matters
resolved in the proceeding on the application for issuance or renewal
of the early site permit, except as provided for in paragraphs (b),
(c), and (d) of this section.
(i) If the early site permit approved an emergency plan (or major
features thereof) that is in use by a licensee of a nuclear power
plant, the Commission shall treat as resolved changes to the early site
permit emergency plan (or major features thereof) that are identical to
changes made to the licensee's emergency plans in compliance with Sec.
50.54(q) of this chapter occurring after issuance of the early site
permit.
(ii) If the early site permit approved an emergency plan (or major
features thereof) that is not in use by a licensee of a nuclear power
plant, the Commission shall treat as resolved changes that are
equivalent to those that could be made under Sec. 50.54(q) of this
chapter without prior NRC approval had the emergency plan been in use
by a licensee.
(b) Updating of early site permit-emergency preparedness. An
applicant for a construction permit, operating license, or combined
license who has filed an application referencing an early site permit
issued under this subpart shall update the emergency preparedness
information that was provided under Sec. 52.17(b), and discuss whether
the updated information materially changes the bases for compliance
with applicable NRC requirements.
(c) Hearings and petitions. (1) In any proceeding for the issuance
of a construction permit, operating license, or combined license
referencing an early site permit, contentions on the following matters
may be litigated in the same manner as other issues material to the
proceeding:
(i) The nuclear power reactor proposed to be built does not fit
within one or more of the site characteristics or design parameters
included in the early site permit;
(ii) One or more of the terms and conditions of the early site
permit have not been met;
(iii) A variance requested under paragraph (d) of this section is
unwarranted or should be modified;
(iv) New or additional information is provided in the application
that substantially alters the bases for a previous NRC conclusion or
constitutes a sufficient basis for the Commission to modify or impose
new terms and conditions related to emergency preparedness; or
(v) Any significant environmental issue that was not resolved in
the early site permit proceeding, or any issue involving the impacts of
construction and operation of the facility that was resolved in the
early site permit proceeding for which significant new information has
been identified.
(2) Any person may file a petition requesting that the site
characteristics, design parameters, or terms and conditions of the
early site permit should be modified, or that the permit should be
suspended or revoked. The petition will be considered in accordance
with Sec. 2.206 of this chapter. Before construction commences, the
Commission shall consider the petition and determine whether any
immediate action is required. If the petition is granted, an
appropriate order will be issued. Construction under the construction
permit or combined license will not be affected by the granting of the
petition unless the order is made immediately effective. Any change
required by the Commission in response to the petition must meet the
requirements of paragraph (a)(1) of this section.
(d) Variances. An applicant for a construction permit, operating
license, or combined license referencing an early site permit may
include in its application a request for a variance from
[[Page 49526]]
one or more site characteristics, design parameters, or terms and
conditions of the early site permit, or from the site safety analysis
report. In determining whether to grant the variance, the Commission
shall apply the same technically relevant criteria applicable to the
application for the original or renewed early site permit. Once a
construction permit or combined license referencing an early site
permit is issued, variances from the early site permit will not be
granted for that construction permit or combined license.
(e) Early site permit amendment. The holder of an early site permit
may not make changes to the early site permit, including the site
safety analysis report, without prior Commission approval. The request
for a change to the early site permit must be in the form of an
application for a license amendment, and must meet the requirements of
10 CFR 50.90 and 50.92.
(f) Information requests. Except for information requests seeking
to verify compliance with the current licensing basis of the early site
permit, information requests to the holder of an early site permit must
be evaluated before issuance to ensure that the burden to be imposed on
respondents is justified in view of the potential safety significance
of the issue to be addressed in the requested information. Each
evaluation performed by the NRC staff must be in accordance with 10 CFR
50.54(f), and must be approved by the Executive Director for Operations
or his or her designee before issuance of the request.
Subpart B--Standard Design Certifications
Sec. 52.41 Scope of subpart.
(a) This subpart sets forth the requirements and procedures
applicable to Commission issuance of rules granting standard design
certifications for nuclear power facilities separate from the filing of
an application for a construction permit or combined license for such a
facility.
(b)(1) Any person may seek a standard design certification for an
essentially complete nuclear power plant design which is an
evolutionary change from light water reactor designs of plants which
have been licensed and in commercial operation before April 18, 1989.
(2) Any person may also seek a standard design certification for a
nuclear power plant design which differs significantly from the light
water reactor designs described in paragraph (b)(1) of this section or
uses simplified, inherent, passive, or other innovative means to
accomplish its safety functions.
Sec. 52.43 Relationship to other subparts.
(a) This subpart applies to a person that requests a standard
design certification from the NRC separately from an application for a
combined license filed under subpart C of this part for a nuclear power
facility. An applicant for a combined license may reference a standard
design certification.
(b) Subpart E of this part governs the NRC staff review and
approval of a final standard design. Subpart E may be used
independently of the provisions in this subpart.
(c) Subpart F of this part governs the issuance of licenses to
manufacture nuclear power reactors to be installed and operated at
sites not identified in the manufacturing license application. Subpart
F may be used independently of the provisions in this subpart. However,
an applicant for a manufacturing license under subpart F may reference
a design certification.
Sec. 52.45 Filing of applications.
(a) An application for design certification may be filed
notwithstanding the fact that an application for a construction permit,
combined license, or manufacturing license for such a facility has not
been filed.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and Sec. Sec. 2.811 through 2.819 of
this chapter.
(c) The fees associated with the review of an application for the
initial issuance or renewal of a standard design certification are set
forth in 10 CFR part 170.
Sec. 52.46 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (c) and (j).
Sec. 52.47 Contents of applications; technical information.
The application must contain a level of design information
sufficient to enable the Commission to judge the applicant's proposed
means of assuring that construction conforms to the design and to reach
a final conclusion on all safety questions associated with the design
before the certification is granted. The information submitted for a
design certification must include performance requirements and design
information sufficiently detailed to permit the preparation of
acceptance and inspection requirements by the NRC, and procurement
specifications and construction and installation specifications by an
applicant. The Commission will require, before design certification,
that information normally contained in certain procurement
specifications and construction and installation specifications be
completed and available for audit if the information is necessary for
the Commission to make its safety determination.
(a) The application must contain a final safety analysis report
(FSAR) that describes the facility, presents the design bases and the
limits on its operation, and presents a safety analysis of the
structures, systems, and components and of the facility as a whole, and
must include the following information:
(1) The site parameters postulated for the design, and an analysis
and evaluation of the design in terms of those site parameters;
(2) A description and analysis of the structures, systems, and
components (SSCs) of the facility, with emphasis upon performance
requirements, the bases, with technical justification therefor, upon
which these requirements have been established, and the evaluations
required to show that safety functions will be accomplished. It is
expected that the standard plant will reflect through its design,
construction, and operation an extremely low probability for accidents
that could result in the release of significant quantities of
radioactive fission products. The description shall be sufficient to
permit understanding of the system designs and their relationship to
the safety evaluations. Such items as the reactor core, reactor coolant
system, instrumentation and control systems, electrical systems,
containment system, other engineered safety features, auxiliary and
emergency systems, power conversion systems, radioactive waste handling
systems, and fuel handling systems shall be discussed insofar as they
are pertinent. The following power reactor design characteristics will
be taken into consideration by the Commission:
(i) Intended use of the reactor including the proposed maximum
power level and the nature and inventory of contained radioactive
materials;
(ii) The extent to which generally accepted engineering standards
are applied to the design of the reactor;
(iii) The extent to which the reactor incorporates unique, unusual
or enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials; and
[[Page 49527]]
(iv) The safety features that are to be engineered into the
facility and those barriers that must be breached as a result of an
accident before a release of radioactive material to the environment
can occur. Special attention must be directed to plant design features
intended to mitigate the radiological consequences of accidents. In
performing this assessment, an applicant shall assume a fission product
release \3\ from the core into the containment assuming that the
facility is operated at the ultimate power level contemplated. The
applicant shall perform an evaluation and analysis of the postulated
fission product release, using the expected demonstrable containment
leak rate and any fission product cleanup systems intended to mitigate
the consequences of the accidents, together with applicable postulated
site parameters, including site meteorology, to evaluate the offsite
radiological consequences. The evaluation must determine that:
---------------------------------------------------------------------------
\3\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events. These accidents have generally been assumed to
result in substantial meltdown of the core with subsequent release
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(A) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \4\ total effective dose equivalent (TEDE);
---------------------------------------------------------------------------
\4\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose
for radiation workers which, according to NCRP recommendations at
the time could be disregarded in the determination of their
radiation exposure status (see NBS Handbook 69 dated June 5, 1959).
However, its use is not intended to imply that this number
constitutes an acceptable limit for an emergency dose to the public
under accident conditions. This dose value has been set forth in
this section as a reference value, which can be used in the
evaluation of plant design features with respect to postulated
reactor accidents, to assure that these designs provide assurance of
low risk of public exposure to radiation, in the event of an
accident.
---------------------------------------------------------------------------
(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
TEDE;
(3) The design of the facility including:
(i) The principal design criteria for the facility. Appendix A to
10 CFR part 50, general design criteria (GDC), establishes minimum
requirements for the principal design criteria for water-cooled nuclear
power plants similar in design and location to plants for which
construction permits have previously been issued by the Commission and
provides guidance to applicants in establishing principal design
criteria for other types of nuclear power units;
(ii) The design bases and the relation of the design bases to the
principal design criteria;
(iii) Information relative to materials of construction, general
arrangement, and approximate dimensions, sufficient to provide
reasonable assurance that the design will conform to the design bases
with an adequate margin for safety;
(4) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and
components provided for the prevention of accidents and the mitigation
of the consequences of accidents. Analysis and evaluation of emergency
core cooling system (ECCS) cooling performance and the need for high-
point vents following postulated loss-of-coolant accidents shall be
performed in accordance with the requirements of Sec. Sec. 50.46 and
50.46a of this chapter;
(5) The kinds and quantities of radioactive materials expected to
be produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter;
(6) The information required by Sec. 20.1406 of this chapter;
(7) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(8) The information necessary to demonstrate compliance with any
technically relevant portions of the Three Mile Island requirements set
forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix),
and (f)(3)(v);
(9) For applications for light-water-cooled nuclear power plants,
an evaluation of the standard plant design against the Standard Review
Plan (SRP) revision in effect 6 months before the docket date of the
application. The evaluation required by this section shall include an
identification and description of all differences in design features,
analytical techniques, and procedural measures proposed for the design
and those corresponding features, techniques, and measures given in the
SRP acceptance criteria. Where a difference exists, the evaluation
shall discuss how the proposed alternative provides an acceptable
method of complying with the Commission's regulations, or portions
thereof, that underlie the corresponding SRP acceptance criteria. The
SRP is not a substitute for the regulations, and compliance is not a
requirement.
(10) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations described in 10 CFR
50.34a(e);
(11) Proposed technical specifications prepared in accordance with
the requirements of Sec. Sec. 50.36 and 50.36a of this chapter;
(12) An analysis and description of the equipment and systems for
combustible gas control as required by 10 CFR 50.44;
(13) The list of electric equipment important to safety that is
required by 10 CFR 50.49(d);
(14) A description of protection provided against pressurized
thermal shock events, including projected values of the reference
temperature for reactor vessel beltline materials as defined in 10 CFR
50.60 and 50.61;
(15) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram events in Sec. 50.62;
(16) A coping analysis, and any design features necessary to
address station blackout, as required by 10 CFR 50.63;
(17) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68(b)(2)-(b)(4);
(18) A description and analysis of the fire protection design
features for the standard plant necessary to comply with 10 CFR part
50, appendix A, GDC 3, and Sec. 50.48 of this chapter;
(19) A description of the quality assurance program applied to the
design of the structures, systems, and components of the facility.
Appendix B to 10 CFR part 50, ``Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants,'' sets forth the
requirements for quality assurance programs for nuclear power plants.
The description of the quality assurance program for a nuclear power
plant shall include a discussion of how the applicable requirements of
appendix B to 10 CFR part 50 were satisfied;
(20) The information necessary to demonstrate that the standard
plant complies with the earthquake
[[Page 49528]]
engineering criteria in 10 CFR part 50, appendix S;
(21) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(22) The information necessary to demonstrate how operating
experience insights have been incorporated into the plant design;
(23) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection,
hydrogen combustion, and containment bypass;
(24) A representative conceptual design for those portions of the
plant for which the application does not seek certification, to aid the
NRC in its review of the FSAR and to permit assessment of the adequacy
of the interface requirements in paragraph (a)(25) of this section;
(25) The interface requirements to be met by those portions of the
plant for which the application does not seek certification. These
requirements must be sufficiently detailed to allow completion of the
FSAR;
(26) Justification that compliance with the interface requirements
of paragraph (a)(25) of this section is verifiable through inspections,
tests, or analyses. The method to be used for verification of interface
requirements must be included as part of the proposed ITAAC required by
paragraph (b)(1) of this section; and
(27) A description of the design-specific probabilistic risk
assessment (PRA) and its results.
(b) The application must also contain:
(1) The proposed inspections, tests, analyses, and acceptance
criteria that are necessary and sufficient to provide reasonable
assurance that, if the inspections, tests, and analyses are performed
and the acceptance criteria met, a facility that incorporates the
design certification has been constructed and will be operated in
conformity with the design certification, the provisions of the Act,
and the Commission's rules and regulations; and
(2) An environmental report as required by 10 CFR 51.55.
(c) This paragraph applies, according to its provisions, to
particular applications:
(1) An application for certification of a nuclear power reactor
design that is an evolutionary change from light-water reactor designs
of plants that have been licensed and in commercial operation before
April 18, 1989, must provide an essentially complete nuclear power
plant design except for site-specific elements such as the service
water intake structure and the ultimate heat sink;
(2) An application for certification of a nuclear power reactor
design that differs significantly from the light-water reactor designs
described in paragraph (c)(1) of this section or uses simplified,
inherent, passive, or other innovative means to accomplish its safety
functions must provide an essentially complete nuclear power reactor
design except for site-specific elements such as the service water
intake structure and the ultimate heat sink, and must meet the
requirements of 10 CFR 50.43(e); and
(3) An application for certification of a modular nuclear power
reactor design must describe and analyze the possible operating
configurations of the reactor modules with common systems, interface
requirements, and system interactions. The final safety analysis must
also account for differences among the configurations, including any
restrictions that will be necessary during the construction and startup
of a given module to ensure the safe operation of any module already
operating.
Sec. 52.48 Standards for review of applications.
Applications filed under this subpart will be reviewed for
compliance with the standards set out in 10 CFR parts 20, 50 and its
appendices, 51, 73, and 100.
Sec. 52.51 Administrative review of applications.
(a) A standard design certification is a rule that will be issued
in accordance with the provisions of subpart H of 10 CFR part 2, as
supplemented by the provisions of this section. The Commission shall
initiate the rulemaking after an application has been filed under Sec.
52.45 and shall specify the procedures to be used for the rulemaking.
The notice of proposed rulemaking published in the Federal Register
must provide an opportunity for the submission of comments on the
proposed design certification rule. If, at the time a proposed design
certification rule is published in the Federal Register under this
paragraph (a), the Commission decides that a legislative hearing should
be held, the information required by 10 CFR 2.1502(c) must be included
in the Federal Register document for the proposed design certification.
(b) Following the submission of comments on the proposed design
certification rule, the Commission may, at its discretion, hold a
legislative hearing under the procedures in subpart O of part 2 of this
chapter. The Commission shall publish a document in the Federal
Register of its decision to hold a legislative hearing. The document
shall contain the information specified in paragraph (c) of this
section, and specify whether the Commission or a presiding officer will
conduct the legislative hearing.
(c) Notwithstanding anything in 10 CFR 2.390 to the contrary,
proprietary information will be protected in the same manner and to the
same extent as proprietary information submitted in connection with
applications for licenses, provided that the design certification shall
be published in Chapter I of this title.
Sec. 52.53 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application which
concern safety.
Sec. 52.54 Issuance of standard design certification.
(a) After conducting a rulemaking proceeding under Sec. 52.51 on
an application for a standard design certification and receiving the
report to be submitted by the Advisory Committee on Reactor Safeguards
under Sec. 52.53, the Commission may issue a standard design
certification in the form of a rule for the design which is the subject
of the application, if the Commission determines that:
(1) The application meets the applicable standards and requirements
of the Atomic Energy Act and the Commission's regulations;
(2) Notifications, if any, to other agencies or bodies have been
duly made;
(3) There is reasonable assurance that the standard design conforms
with the provisions of the Act, and the Commission's regulations;
(4) The applicant is technically qualified;
(5) The proposed inspections, tests, analyses, and acceptance
criteria are necessary and sufficient, within the scope of the standard
design, to provide reasonable assurance that, if the inspections,
tests, and analyses are performed and the acceptance criteria met, the
facility has been constructed and will be operated in accordance with
the design certification, the provisions of the Act, and the
Commission's regulations;
[[Page 49529]]
(6) Issuance of the standard design certification will not be
inimical to the common defense and security or to the health and safety
of the public;
(7) The findings required by subpart A of part 51 of this chapter
have been made; and
(8) The applicant has implemented the quality assurance program
described or referenced in the safety analysis report.
(b) The design certification rule must specify the site parameters,
design characteristics, and any additional requirements and
restrictions of the design certification rule.
(c) After the Commission has adopted a final design certification
rule, the applicant shall not permit any individual to have access to
or any facility to possess restricted data or classified National
Security Information until the individual and/or facility has been
approved for access under the provisions of 10 CFR parts 25 and/or 95,
as applicable.
Sec. 52.55 Duration of certification.
(a) Except as provided in paragraph (b) of this section, a standard
design certification issued under this subpart is valid for 15 years
from the date of issuance.
(b) A standard design certification continues to be valid beyond
the date of expiration in any proceeding on an application for a
combined license or an operating license that references the standard
design certification and is docketed either before the date of
expiration of the certification, or, if a timely application for
renewal of the certification has been filed, before the Commission has
determined whether to renew the certification. A design certification
also continues to be valid beyond the date of expiration in any hearing
held under Sec. 52.103 before operation begins under a combined
license that references the design certification.
(c) An applicant for a construction permit or a combined license
may, at its own risk, reference in its application a design for which a
design certification application has been docketed but not granted.
Sec. 52.57 Application for renewal.
(a) Not less than 12 nor more than 36 months before the expiration
of the initial 15-year period, or any later renewal period, any person
may apply for renewal of the certification. An application for renewal
must contain all information necessary to bring up to date the
information and data contained in the previous application. The
Commission will require, before renewal of certification, that
information normally contained in certain procurement specifications
and construction and installation specifications be completed and
available for audit if this information is necessary for the Commission
to make its safety determination. Notice and comment procedures must be
used for a rulemaking proceeding on the application for renewal. The
Commission, in its discretion, may require the use of additional
procedures in individual renewal proceedings.
(b) A design certification, either original or renewed, for which a
timely application for renewal has been filed remains in effect until
the Commission has determined whether to renew the certification. If
the certification is not renewed, it continues to be valid in certain
proceedings, in accordance with the provisions of Sec. 52.55.
(c) The Commission shall refer a copy of the application for
renewal to the Advisory Committee on Reactor Safeguards (ACRS). The
ACRS shall report on those portions of the application which concern
safety and shall apply the criteria set forth in Sec. 52.59.
Sec. 52.59 Criteria for renewal.
(a) The Commission shall issue a rule granting the renewal if the
design, either as originally certified or as modified during the
rulemaking on the renewal, complies with the Atomic Energy Act and the
Commission's regulations applicable and in effect at the time the
certification was issued.
(b) The Commission may impose other requirements if it determines
that:
(1) They are necessary for adequate protection to public health and
safety or common defense and security;
(2) They are necessary for compliance with the Commission's
regulations and orders applicable and in effect at the time the design
certification was issued; or
(3) There is a substantial increase in overall protection of the
public health and safety or the common defense and security to be
derived from the new requirements, and the direct and indirect costs of
implementing those requirements are justified in view of this increased
protection.
(c) In addition, the applicant for renewal may request an amendment
to the design certification. The Commission shall grant the amendment
request if it determines that the amendment will comply with the Atomic
Energy Act and the Commission's regulations in effect at the time of
renewal. If the amendment request entails such an extensive change to
the design certification that an essentially new standard design is
being proposed, an application for a design certification must be filed
in accordance with this subpart.
(d) Denial of renewal does not bar the applicant, or another
applicant, from filing a new application for certification of the
design, which proposes design changes that correct the deficiencies
cited in the denial of the renewal.
Sec. 52.61 Duration of renewal.
Each renewal of certification for a standard design will be for not
less than 10, nor more than 15 years.
Sec. 52.63 Finality of standard design certifications.
(a)(1) Notwithstanding any provision in 10 CFR 50.109, while a
standard design certification rule is in effect under Sec. Sec. 52.55
or 52.61, the Commission may not modify, rescind, or impose new
requirements on the certification information, whether on its own
motion, or in response to a petition from any person, unless the
Commission determines in a rulemaking that the change:
(i) Is necessary either to bring the certification information or
the referencing plants into compliance with the Commission's
regulations applicable and in effect at the time the certification was
issued;
(ii) Is necessary to provide adequate protection of the public
health and safety or the common defense and security;
(iii) Reduces unnecessary regulatory burden and maintains
protection to public health and safety and the common defense and
security;
(iv) Provides the detailed design information to be verified under
those inspections, tests, analyses, and acceptance criteria (ITAAC)
which are directed at certification information (i.e., design
acceptance criteria);
(v) Is necessary to correct material errors in the certification
information;
(vi) Substantially increases overall safety, reliability, or
security of facility design, construction, or operation, and the direct
and indirect costs of implementation of the rule change are justified
in view of this increased safety, reliability, or security; or
(vii) Contributes to increased standardization of the certification
information.
(2)(i) In a rulemaking under Sec. 52.63(a)(1), except for Sec.
52.63(a)(1)(ii), the Commission will give consideration to whether the
benefits justify the costs for plants that are already licensed or for
which an application for a permit or license is under consideration.
[[Page 49530]]
(ii) The rulemaking procedures for changes under Sec. 52.63(a)(1)
must provide for notice and opportunity for public comment.
(3) Any modification the NRC imposes on a design certification rule
under paragraph (a)(1) of this section will be applied to all plants
referencing the certified design, except those to which the
modification has been rendered technically irrelevant by action taken
under paragraphs (a)(4) or (b)(1) of this section.
(4) The Commission may not impose new requirements by plant-
specific order on any part of the design of a specific plant
referencing the design certification rule if that part was approved in
the design certification while a design certification rule is in effect
under Sec. 52.55 or Sec. 52.61, unless:
(i) A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time the
certification was issued, or to assure adequate protection of the
public health and safety or the common defense and security; and
(ii) Special circumstances as defined in 10 CFR 52.7 are present.
In addition to the factors listed in Sec. 52.7, the Commission shall
consider whether the special circumstances which Sec. 52.7 requires to
be present outweigh any decrease in safety that may result from the
reduction in standardization caused by the plant-specific order.
(5) Except as provided in 10 CFR 2.335, in making the findings
required for issuance of a combined license, construction permit,
operating license, or manufacturing license, or for any hearing under
Sec. 52.103, the Commission shall treat as resolved those matters
resolved in connection with the issuance or renewal of a design
certification rule.
(b)(1) An applicant or licensee who references a design
certification rule may request an exemption from one or more elements
of the certification information. The Commission may grant such a
request only if it determines that the exemption will comply with the
requirements of Sec. 52.7. In addition to the factors listed in Sec.
52.7, the Commission shall consider whether the special circumstances
that Sec. 52.7 requires to be present outweigh any decrease in safety
that may result from the reduction in standardization caused by the
exemption. The granting of an exemption on request of an applicant is
subject to litigation in the same manner as other issues in the
operating license or combined license hearing.
(2) Subject to Sec. 50.59 of this chapter, a licensee who
references a design certification rule may make departures from the
design of the nuclear power facility, without prior Commission
approval, unless the proposed departure involves a change to the design
as described in the rule certifying the design. The licensee shall
maintain records of all departures from the facility and these records
must be maintained and available for audit until the date of
termination of the license.
(c) The Commission will require, before granting a construction
permit, combined license, operating license, or manufacturing license
which references a design certification rule, that information normally
contained in certain procurement specifications and construction and
installation specifications be completed and available for audit if the
information is necessary for the Commission to make its safety
determinations, including the determination that the application is
consistent with the certification information. This information may be
acquired by appropriate arrangements with the design certification
applicant.
Subpart C--Combined Licenses
Sec. 52.71 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of combined licenses for nuclear power facilities.
Sec. 52.73 Relationship to other subparts.
(a) An application for a combined license under this subpart may,
but need not, reference a standard design certification, standard
design approval, or manufacturing license issued under subparts B, E,
or F of this part, respectively, or an early site permit issued under
subpart A of this part. In the absence of a demonstration that an
entity other than the one originally sponsoring and obtaining a design
certification is qualified to supply a design, the Commission will
entertain an application for a combined license that references a
standard design certification issued under subpart B of this part only
if the entity that sponsored and obtained the certification supplies
the design for the applicant's use.
(b) The Commission will require, before granting a combined license
that references a standard design certification, that information
normally contained in certain procurement specifications and
construction and installation specifications be completed and available
for audit if the information is necessary for the Commission to make
its safety determinations, including the determination that the
application is consistent with the certification information.
Sec. 52.75 Filing of applications.
(a) Any person except one excluded by 10 CFR 50.38 may file an
application for a combined license for a nuclear power facility with
the Director of New Reactors or the Director of Nuclear Reactor
Regulation, as appropriate.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and 50.30 of this chapter.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Sec. 52.77 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33.
Sec. 52.79 Contents of applications; technical information in final
safety analysis report.
(a) The application must contain a final safety analysis report
that describes the facility, presents the design bases and the limits
on its operation, and presents a safety analysis of the structures,
systems, and components of the facility as a whole. The final safety
analysis report shall include the following information, at a level of
information sufficient to enable the Commission to reach a final
conclusion on all safety matters that must be resolved by the
Commission before issuance of a combined license:
(1)(i) The boundaries of the site;
(ii) The proposed general location of each facility on the site;
(iii) The seismic, meteorological, hydrologic, and geologic
characteristics of the proposed site with appropriate consideration of
the most severe of the natural phenomena that have been historically
reported for the site and surrounding area and with sufficient margin
for the limited accuracy, quantity, and time in which the historical
data have been accumulated;
(iv) The location and description of any nearby industrial,
military, or transportation facilities and routes;
(v) The existing and projected future population profile of the
area surrounding the site;
(vi) A description and safety assessment of the site on which the
facility is to be located. The assessment must contain an analysis and
evaluation of the major structures, systems, and components of the
facility that bear significantly on the acceptability of the site under
the radiological consequence evaluation factors identified in
paragraphs (a)(1)(vi)(A) and (a)(1)(vi)(B)
[[Page 49531]]
of this section. In performing this assessment, an applicant shall
assume a fission product release \5\ from the core into the containment
assuming that the facility is operated at the ultimate power level
contemplated. The applicant shall perform an evaluation and analysis of
the postulated fission product release, using the expected demonstrable
containment leak rate and any fission product cleanup systems intended
to mitigate the consequences of the accidents, together with applicable
site characteristics, including site meteorology, to evaluate the
offsite radiological consequences. Site characteristics must comply
with part 100 of this chapter. The evaluation must determine that:
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\5\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events. These accidents have generally been assumed to
result in substantial meltdown of the core with subsequent release
into the containment of appreciable quantities of fission products.
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(A) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \6\ total effective dose equivalent (TEDE).
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\6\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose
for radiation workers which, according to NCRP recommendations at
the time could be disregarded in the determination of their
radiation exposure status (see NBS Handbook 69 dated June 5, 1959).
However, its use is not intended to imply that this number
constitutes an acceptable limit for an emergency dose to the public
under accident conditions. Rather, this dose value has been set
forth in this section as a reference value, which can be used in the
evaluation of plant design features with respect to postulated
reactor accidents, to assure that these designs provide assurance of
low risk of public exposure to radiation, in the event of an
accident.
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(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
TEDE; and
(2) A description and analysis of the structures, systems, and
components of the facility with emphasis upon performance requirements,
the bases, with technical justification therefor, upon which these
requirements have been established, and the evaluations required to
show that safety functions will be accomplished. It is expected that
reactors will reflect through their design, construction, and operation
an extremely low probability for accidents that could result in the
release of significant quantities of radioactive fission products. The
descriptions shall be sufficient to permit understanding of the system
designs and their relationship to safety evaluations. Items such as the
reactor core, reactor coolant system, instrumentation and control
systems, electrical systems, containment system, other engineered
safety features, auxiliary and emergency systems, power conversion
systems, radioactive waste handling systems, and fuel handling systems
shall be discussed insofar as they are pertinent. The following power
reactor design characteristics and proposed operation will be taken
into consideration by the Commission:
(i) Intended use of the reactor including the proposed maximum
power level and the nature and inventory of contained radioactive
materials;
(ii) The extent to which generally accepted engineering standards
are applied to the design of the reactor;
(iii) The extent to which the reactor incorporates unique, unusual
or enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials;
(iv) The safety features that are to be engineered into the
facility and those barriers that must be breached as a result of an
accident before a release of radioactive material to the environment
can occur. Special attention must be directed to plant design features
intended to mitigate the radiological consequences of accidents. In
performing this assessment, an applicant shall assume a fission product
release \7\ from the core into the containment assuming that the
facility is operated at the ultimate power level contemplated;
---------------------------------------------------------------------------
\7\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events. These accidents have generally been assumed to
result in substantial meltdown of the core with subsequent release
into the containment of appreciable quantities of fission products.
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(3) The kinds and quantities of radioactive materials expected to
be produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter;
(4) The design of the facility including:
(i) The principal design criteria for the facility. Appendix A to
part 50 of this chapter, ``General Design Criteria for Nuclear Power
Plants,'' establishes minimum requirements for the principal design
criteria for water-cooled nuclear power plants similar in design and
location to plants for which construction permits have previously been
issued by the Commission and provides guidance to applicants in
establishing principal design criteria for other types of nuclear power
units;
(ii) The design bases and the relation of the design bases to the
principal design criteria;
(iii) Information relative to materials of construction,
arrangement, and dimensions, sufficient to provide reasonable assurance
that the design will conform to the design bases with adequate margin
for safety.
(5) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and
components provided for the prevention of accidents and the mitigation
of the consequences of accidents. Analysis and evaluation of ECCS
cooling performance and the need for high-point vents following
postulated loss-of-coolant accidents shall be performed in accordance
with the requirements of Sec. Sec. 50.46 and 50.46a of this chapter;
(6) A description and analysis of the fire protection design
features for the reactor necessary to comply with 10 CFR part 50,
appendix A, GDC 3, and Sec. 50.48 of this chapter;
(7) A description of protection provided against pressurized
thermal shock events, including projected values of the reference
temperature for reactor vessel beltline materials as defined in
Sec. Sec. 50.60 and 50.61(b)(1) and (b)(2) of this chapter;
(8) An analysis and description of the equipment and systems for
combustible gas control as required by Sec. 50.44 of this chapter;
(9) The coping analyses, and any design features necessary to
address station blackout, as described in Sec. 50.63 of this chapter;
(10) A description of the program, and its implementation, required
by Sec. 50.49(a) of this chapter for the environmental qualification
of electric equipment important to safety and the list of electric
equipment important to safety that is required by 10 CFR 50.49(d);
(11) A description of the program(s), and their implementation,
necessary to ensure that the systems and components meet the
requirements of the ASME
[[Page 49532]]
Boiler and Pressure Vessel Code and the ASME Code for Operation and
Maintenance of Nuclear Power Plants in accordance with 50.55a of this
chapter;
(12) A description of the primary containment leakage rate testing
program, and its implementation, necessary to ensure that the
containment meets the requirements of appendix J to 10 CFR part 50;
(13) A description of the reactor vessel material surveillance
program required by appendix H to 10 CFR part 50 and its
implementation;
(14) A description of the operator training program, and its
implementation, necessary to meet the requirements of 10 CFR part 55;
(15) A description of the program, and its implementation, for
monitoring the effectiveness of maintenance necessary to meet the
requirements of Sec. 50.65 of this chapter;
(16)(i) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations, as described in
Sec. 50.34a(d) of this chapter;
(ii) A description of the process and effluent monitoring and
sampling program required by appendix I to 10 CFR part 50 and its
implementation.
(17) The information with respect to compliance with technically
relevant positions of the Three Mile Island requirements in Sec.
50.34(f) of this chapter, with the exception of Sec. Sec.
50.34(f)(1)(xii), (f)(2)(ix), and (f)(3)(v);
(18) If the applicant seeks to use risk-informed treatment of SSCs
in accordance with Sec. 50.69 of this chapter, the information
required by Sec. 50.69(b)(2) of this chapter;
(19) Information necessary to demonstrate that the plant complies
with the earthquake engineering criteria in 10 CFR part 50, appendix S;
(20) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(21) Emergency plans complying with the requirements of Sec. 50.47
of this chapter, and 10 CFR part 50, appendix E;
(22)(i) All emergency plan certifications that have been obtained
from the State and local governmental agencies with emergency planning
responsibilities must state that:
(A) The proposed emergency plans are practicable;
(B) These agencies are committed to participating in any further
development of the plans, including any required field demonstrations;
and
(C) These agencies are committed to executing their
responsibilities under the plans in the event of an emergency;
(ii) If certifications cannot be obtained after sustained, good
faith efforts by the applicant, then the application must contain
information, including a utility plan, sufficient to show that the
proposed plans provide reasonable assurance that adequate protective
measures can and will be taken in the event of a radiological emergency
at the site.
(23) [Reserved]
(24) If the application is for a nuclear power reactor design which
differs significantly from light-water reactor designs that were
licensed before 1997 or use simplified, inherent, passive, or other
innovative means to accomplish their safety functions, the application
must describe how the design meets the requirements in Sec. 50.43(e)
of this chapter;
(25) A description of the quality assurance program, applied to the
design, and to be applied to the fabrication, construction, and
testing, of the structures, systems, and components of the facility.
Appendix B to 10 CFR part 50 sets forth the requirements for quality
assurance programs for nuclear power plants. The description of the
quality assurance program for a nuclear power plant must include a
discussion of how the applicable requirements of appendix B to 10 CFR
part 50 have been and will be satisfied, including a discussion of how
the quality assurance program will be implemented;
(26) The applicant's organizational structure, allocations or
responsibilities and authorities, and personnel qualifications
requirements for operation;
(27) Managerial and administrative controls to be used to assure
safe operation. Appendix B to 10 CFR part 50 sets forth the
requirements for these controls for nuclear power plants. The
information on the controls to be used for a nuclear power plant shall
include a discussion of how the applicable requirements of appendix B
to 10 CFR part 50 will be satisfied;
(28) Plans for preoperational testing and initial operations;
(29)(i) Plans for conduct of normal operations, including
maintenance, surveillance, and periodic testing of structures, systems,
and components;
(ii) Plans for coping with emergencies, other than the plans
required by Sec. 52.79(a)(21);
(30) Proposed technical specifications prepared in accordance with
the requirements of Sec. Sec. 50.36 and 50.36a of this chapter;
(31) For nuclear power plants to be operated on multi-unit sites,
an evaluation of the potential hazards to the structures, systems, and
components important to safety of operating units resulting from
construction activities, as well as a description of the managerial and
administrative controls to be used to provide assurance that the
limiting conditions for operation are not exceeded as a result of
construction activities at the multi-unit sites;
(32) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(33) A description of the training program required by Sec. 50.120
of this chapter and its implementation;
(34) A description and plans for implementation of an operator
requalification program. The operator requalification program must as a
minimum, meet the requirements for those programs contained in Sec.
55.59 of this chapter;
(35)(i) A physical security plan, describing how the applicant will
meet the requirements of 10 CFR part 73 (and 10 CFR part 11, if
applicable, including the identification and description of jobs as
required by Sec. 11.11(a) of this chapter, at the proposed facility).
The plan must list tests, inspections, audits, and other means to be
used to demonstrate compliance with the requirements of 10 CFR parts 11
and 73, if applicable;
(ii) A description of the implementation of the physical security
plan;
(36)(i) A safeguards contingency plan in accordance with the
criteria set forth in appendix C to 10 CFR part 73. The safeguards
contingency plan shall include plans for dealing with threats, thefts,
and radiological sabotage, as defined in part 73 of this chapter,
relating to the special nuclear material and nuclear facilities
licensed under this chapter and in the applicant's possession and
control. Each application for this type of license shall include the
information contained in the applicant's safeguards contingency
plan.\8\ (Implementing procedures required for this plan need not be
submitted for approval.)
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\8\ A physical security plan that contains all the information
required in both Sec. 73.55 of this chapter and appendix C to 10
CFR part 73 satisfies the requirement for a contingency plan.
---------------------------------------------------------------------------
(ii) A training and qualification plan in accordance with the
criteria set forth in appendix B to 10 CFR part 73.
(iii) A description of the implementation of the safeguards
[[Page 49533]]
contingency plan and the training and qualification plan;
(iv) Each applicant who prepares a physical security plan, a
safeguards contingency plan, or a guard qualification and training
plan, shall protect the plans and other related Safeguards Information
against unauthorized disclosure in accordance with the requirements of
Sec. 73.21 of this chapter, as appropriate.
(37) The information necessary to demonstrate how operating
experience insights have been incorporated into the plant design;
(38) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection,
hydrogen combustion, and containment bypass;
(39) A description of the radiation protection program required by
Sec. 20.1101 of this chapter and its implementation.
(40) A description of the fire protection program required by Sec.
50.48 of this chapter and its implementation.
(41) For applications for light-water-cooled nuclear power plant
combined licenses, an evaluation of the facility against the Standard
Review Plan (SRP) revision in effect 6 months before the docket date of
the application. The evaluation required by this section shall include
an identification and description of all differences in design
features, analytical techniques, and procedural measures proposed for a
facility and those corresponding features, techniques, and measures
given in the SRP acceptance criteria. Where a difference exists, the
evaluation shall discuss how the proposed alternative provides an
acceptable method of complying with the Commission's regulations, or
portions thereof, that underlie the corresponding SRP acceptance
criteria. The SRP is not a substitute for the regulations, and
compliance is not a requirement;
(42) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram (ATWS) events in Sec. 50.62 of this chapter;
(43) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68 of this chapter;
(44) A description of the fitness-for-duty program required by 10
CFR part 26 and its implementation.
(45) The information required by Sec. 20.1406 of this chapter.
(46) A description of the plant-specific probabilistic risk
assessment (PRA) and its results.
(b) If the combined license application references an early site
permit, then the following requirements apply:
(1) The final safety analysis report need not contain information
or analyses submitted to the Commission in connection with the early
site permit, provided, however, that the final safety analysis report
must either include or incorporate by reference the early site permit
site safety analysis report and must contain, in addition to the
information and analyses otherwise required, information sufficient to
demonstrate that the design of the facility falls within the site
characteristics and design parameters specified in the early site
permit.
(2) If the final safety analysis report does not demonstrate that
design of the facility falls within the site characteristics and design
parameters, the application shall include a request for a variance that
complies with the requirements of Sec. Sec. 52.39 and 52.93.
(3) The final safety analysis report must demonstrate that all
terms and conditions that have been included in the early site permit,
other than those imposed under Sec. 50.36b, will be satisfied by the
date of issuance of the combined license. Any terms or conditions of
the early site permit that could not be met by the time of issuance of
the combined license, must be set forth as terms or conditions of the
combined license.
(4) If the early site permit approves complete and integrated
emergency plans, or major features of emergency plans, then the final
safety analysis report must include any new or additional information
that updates and corrects the information that was provided under Sec.
52.17(b), and discuss whether the new or additional information
materially changes the bases for compliance with the applicable
requirements. The application must identify changes to the emergency
plans or major features of emergency plans that have been incorporated
into the proposed facility emergency plans and that constitute or would
constitute a decrease in effectiveness under Sec. 50.54(q) of this
chapter.
(5) If complete and integrated emergency plans are approved as part
of the early site permit, new certifications meeting the requirements
of paragraph (a)(22) of this section are not required.
(c) If the combined license application references a standard
design approval, then the following requirements apply:
(1) The final safety analysis report need not contain information
or analyses submitted to the Commission in connection with the design
approval, provided, however, that the final safety analysis report must
either include or incorporate by reference the standard design approval
final safety analysis report and must contain, in addition to the
information and analyses otherwise required, information sufficient to
demonstrate that the characteristics of the site fall within the site
parameters specified in the design approval. In addition, the plant-
specific PRA information must use the PRA information for the design
approval and must be updated to account for site-specific design
information and any design changes or departures.
(2) The final safety analysis report must demonstrate that all
terms and conditions that have been included in the final design
approval will be satisfied by the date of issuance of the combined
license.
(d) If the combined license application references a standard
design certification, then the following requirements apply:
(1) The final safety analysis report need not contain information
or analyses submitted to the Commission in connection with the design
certification, provided, however, that the final safety analysis report
must either include or incorporate by reference the standard design
certification final safety analysis report and must contain, in
addition to the information and analyses otherwise required,
information sufficient to demonstrate that the site characteristics
fall within the site parameters specified in the design certification.
In addition, the plant-specific PRA information must use the PRA
information for the design certification and must be updated to account
for site-specific design information and any design changes or
departures.
(2) The final safety analysis report must demonstrate that the
interface requirements established for the design under Sec. 52.47
have been met.
(3) The final safety analysis report must demonstrate that all
requirements and restrictions set forth in the referenced design
certification rule, other than those imposed under Sec. 50.36b, must
be satisfied by the date of issuance of the combined license. Any
requirements and restrictions set forth in the referenced design
certification rule that could not be satisfied by the time of issuance
of the combined license, must be set forth as terms or conditions of
the combined license.
[[Page 49534]]
(e) If the combined license application references the use of one
or more manufactured nuclear power reactors licensed under subpart F of
this part, then the following requirements apply:
(1) The final safety analysis report need not contain information
or analyses submitted to the Commission in connection with the
manufacturing license, provided, however, that the final safety
analysis report must either include or incorporate by reference the
manufacturing license final safety analysis report and must contain, in
addition to the information and analyses otherwise required,
information sufficient to demonstrate that the site characteristics
fall within the site parameters specified in the manufacturing license.
In addition, the plant-specific PRA information must use the PRA
information for the manufactured reactor and must be updated to account
for site-specific design information and any design changes or
departures.
(2) The final safety analysis report must demonstrate that the
interface requirements established for the design have been met.
(3) The final safety analysis report must demonstrate that all
terms and conditions that have been included in the manufacturing
license, other than those imposed under Sec. 50.36b, will be satisfied
by the date of issuance of the combined license. Any terms or
conditions of the manufacturing license that could not be met by the
time of issuance of the combined license, must be set forth as terms or
conditions of the combined license.
Sec. 52.80 Contents of applications; additional technical
information.
The application must contain:
(a) The proposed inspections, tests, and analyses, including those
applicable to emergency planning, that the licensee shall perform, and
the acceptance criteria that are necessary and sufficient to provide
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met, the facility has been
constructed and will be operated in conformity with the combined
license, the provisions of the Act, and the Commission's rules and
regulations.
(1) If the application references an early site permit with ITAAC,
the early site permit ITAAC must apply to those aspects of the combined
license which are approved in the early site permit.
(2) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those
portions of the facility design which are approved in the design
certification.
(3) If the application references an early site permit with ITAAC
or a standard design certification or both, the application may include
a notification that a required inspection, test, or analysis in the
ITAAC has been successfully completed and that the corresponding
acceptance criterion has been met. The Federal Register notification
required by Sec. 52.85 must indicate that the application includes
this notification.
(b) A complete environmental report as required by 10 CFR 51.50(c).
(c) If the applicant wishes to be able to perform the activities at
the site allowed by 10 CFR 50.10(e) before issuance of the combined
license, the applicant must identify and describe the activities that
are requested and propose a plan for redress of the site in the event
that the activities are performed and either construction is abandoned
or the combined license is revoked. The application must demonstrate
that there is reasonable assurance that redress carried out under the
plan will achieve an environmentally stable and aesthetically
acceptable site suitable for whatever non-nuclear use may conform with
local zoning laws.
Sec. 52.81 Standards for review of applications.
Applications filed under this subpart will be reviewed according to
the standards set out in 10 CFR parts 20, 50, 51, 54, 55, 73, 100, and
140.
Sec. 52.83 Finality of referenced NRC approvals; partial initial
decision on site suitability.
(a) If the application for a combined license under this subpart
references an early site permit, design certification rule, standard
design approval, or manufacturing license, the scope and nature of
matters resolved for the application and any combined license issued
are governed by the relevant provisions addressing finality, including
Sec. Sec. 52.39, 52.63, 52.98, 52.145, and 52.171.
(b) While a partial decision on site suitability is in effect under
10 CFR 2.617(b)(2), the scope and nature of matters resolved in the
proceeding are governed by the finality provisions in 10 CFR 2.629.
Sec. 52.85 Administrative review of applications; hearings.
A proceeding on a combined license is subject to all applicable
procedural requirements contained in 10 CFR part 2, including the
requirements for docketing (Sec. 2.101 of this chapter) and issuance
of a notice of hearing (Sec. 2.104 of this chapter). If an applicant
requests a Commission finding on certain ITAAC with the issuance of the
combined license, then those ITAAC will be identified in the notice of
hearing. All hearings on combined licenses are governed by the
procedures contained in 10 CFR part 2.
Sec. 52.87 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application that concern
safety and shall apply the standards referenced in Sec. 52.81, in
accordance with the finality provisions in Sec. 52.83.
Sec. 52.89 [Reserved].
Sec. 52.91 Authorization to conduct site activities.
(a) If the application does not reference an early site permit
which authorizes the applicant to perform site preparation activities,
the applicant may not perform the site preparation activities allowed
by 10 CFR 50.10(e)(1) without obtaining the separate authorization
required by 10 CFR 50.10(e)(1). Authorization may be granted only after
the presiding officer in the proceeding on the application has made the
findings and determination required by 10 CFR 50.10(e)(2) and has
determined that there is reasonable assurance that redress carried out
under the site redress plan will achieve an environmentally stable and
aesthetically acceptable site suitable for whatever non-nuclear use may
conform with local zoning laws.
(b) Authorization to conduct the activities described in 10 CFR
50.10(e)(3)(i) may be granted only after the presiding officer in the
combined license proceeding makes the additional finding required by 10
CFR 50.10(e)(3)(ii).
(c) If, after an applicant for a combined license has performed the
activities permitted by paragraph (a) or (b) of this section, and the
application for the license is withdrawn or denied, then the applicant
shall redress the site in accord with the terms of the site redress
plan. If a use not envisaged in the redress plan is found for the site
or parts before redress is complete, the applicant shall carry out the
redress plan to the greatest extent possible consistent with the
alternate use.
Sec. 52.93 Exemptions and variances.
(a) Applicants for a combined license under this subpart, or any
amendment to a combined license, may include in the application a
request for an
[[Page 49535]]
exemption from one or more of the Commission's regulations.
(1) If the request is for an exemption from any part of a
referenced design certification rule, the Commission may grant the
request if it determines that the exemption complies with any exemption
provisions of the referenced design certification rule, or with Sec.
52.63 if there are no applicable exemption provisions in the referenced
design certification rule.
(2) For all other requests for exemptions, the Commission may grant
a request if it determines that the exemption complies with Sec. 52.7.
(b) An applicant for a combined license who has filed an
application referencing an early site permit issued under subpart A of
this part may include in the application a request for a variance from
one or more site characteristics, design parameters, or terms and
conditions of the permit, or from the site safety analysis report. In
determining whether to grant the variance, the Commission shall apply
the same technically relevant criteria as were applicable to the
application for the original or renewed site permit. Once a
construction permit or combined license referencing an early site
permit is issued, variances from the early site permit will not be
granted for that construction permit or combined license.
(c) An applicant for a combined license who has filed an
application referencing a nuclear power reactor manufactured under a
manufacturing license issued under subpart F of this part may include
in the application a request for a departure from one or more design
characteristics, site parameters, terms and conditions, or approved
design of the manufactured reactor. The Commission may grant a request
only if it determines that the departure will comply with the
requirements of 10 CFR 52.7, and that the special circumstances
outweigh any decrease in safety that may result from the reduction in
standardization caused by the departure.
(d) Issuance of a variance under paragraph (b) or a departure under
paragraph (c) of this section is subject to litigation during the
combined license proceeding in the same manner as other issues material
to that proceeding.
Sec. 52.97 Issuance of combined licenses.
(a)(1) After conducting a hearing in accordance with Sec. 52.85
and receiving the report submitted by the ACRS, the Commission may
issue a combined license if the Commission finds that:
(i) The applicable standards and requirements of the Act and the
Commission's regulations have been met;
(ii) Any required notifications to other agencies or bodies have
been duly made;
(iii) There is reasonable assurance that the facility will be
constructed and will operate in conformity with the license, the
provisions of the Act, and the Commission's regulations.
(iv) The applicant is technically and financially qualified to
engage in the activities authorized; and
(v) Issuance of the license will not be inimical to the common
defense and security or to the health and safety of the public; and
(vi) The findings required by subpart A of part 51 of this chapter
have been made.
(2) The Commission may also find, at the time it issues the
combined license, that certain acceptance criteria in one or more of
the inspections, tests, analyses, and acceptance criteria (ITAAC) in a
referenced early site permit or standard design certification have been
met. This finding will finally resolve that those acceptance criteria
have been met, those acceptance criteria will be deemed to be excluded
from the combined license, and findings under Sec. 52.103(g) with
respect to those acceptance criteria are unnecessary.
(b) The Commission shall identify within the combined license the
inspections, tests, and analyses, including those applicable to
emergency planning, that the licensee shall perform, and the acceptance
criteria that, if met, are necessary and sufficient to provide
reasonable assurance that the facility has been constructed and will be
operated in conformity with the license, the provisions of the Act, and
the Commission's rules and regulations.
(c) A combined license shall contain the terms and conditions,
including technical specifications, as the Commission deems necessary
and appropriate.
Sec. 52.98 Finality of combined licenses; information requests.
(a) After issuance of a combined license, the Commission may not
modify, add, or delete any term or condition of the combined license,
the design of the facility, the inspections, tests, analyses, and
acceptance criteria contained in the license which are not derived from
a referenced standard design certification or manufacturing license,
except in accordance with the provisions of Sec. 52.103 or Sec.
50.109 of this chapter, as applicable.
(b) If the combined license does not reference a design
certification or a reactor manufactured under a subpart F of this part
manufacturing license, then a licensee may make changes in the facility
as described in the final safety analysis report (as updated), make
changes in the procedures as described in the final safety analysis
report (as updated), and conduct tests or experiments not described in
the final safety analysis report (as updated) under the applicable
change processes in 10 CFR part 50 (e.g., Sec. Sec. 50.54, 50.59, or
50.90 of this chapter).
(c) If the combined license references a certified design, then--
(1) Changes to or departures from information within the scope of
the referenced design certification rule are subject to the applicable
change processes in that rule; and
(2) Changes that are not within the scope of the referenced design
certification rule are subject to the applicable change processes in 10
CFR part 50, unless they also involve changes to or noncompliance with
information within the scope of the referenced design certification
rule. In these cases, the applicable provisions of this section and the
design certification rule apply.
(d) If the combined license references a reactor manufactured under
a subpart F of this part manufacturing license, then--
(1) Changes to or departures from information within the scope of
the manufactured reactor's design are subject to the change processes
in Sec. 52.171; and
(2) Changes that are not within the scope of the manufactured
reactor's design are subject to the applicable change processes in 10
CFR part 50.
(e) The Commission may issue and make immediately effective any
amendment to a combined license upon a determination by the Commission
that the amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person. The amendment may be issued and made
immediately effective in advance of the holding and completion of any
required hearing. The amendment will be processed in accordance with
the procedures specified in 10 CFR 50.91.
(f) Any modification to, addition to, or deletion from the terms
and conditions of a combined license, including any modification to,
addition to, or deletion from the inspections, tests, analyses, or
related acceptance criteria contained in the license is a proposed
amendment to the license. There must be an opportunity for a hearing on
the amendment.
(g) Except for information sought to verify licensee compliance
with the
[[Page 49536]]
current licensing basis for that facility, information requests to the
holder of a combined license must be evaluated before issuance to
ensure that the burden to be imposed on the licensee is justified in
view of the potential safety significance of the issue to be addressed
in the requested information. Each evaluation performed by the NRC
staff must be in accordance with 10 CFR 50.54(f) and must be approved
by the Executive Director for Operations or his or her designee before
issuance of the request.
Sec. 52.99 Inspection during construction.
(a) The licensee shall submit to the NRC, no later than 1 year
after issuance of the combined license or at the start of construction
as defined in 10 CFR 50.10(b), whichever is later, its schedule for
completing the inspections, tests, or analyses in the ITAAC. The
licensee shall submit updates to the ITAAC schedule every 6 months
thereafter and, within 1 year of its scheduled date for initial loading
of fuel, the licensee shall submit updates to the ITAAC schedule every
30 days until the final notification is provided to the NRC under
paragraph (c)(1) of this section.
(b) With respect to activities subject to an ITAAC, an applicant
for a combined license may proceed at its own risk with design and
procurement activities, and a licensee may proceed at its own risk with
design, procurement, construction, and pre-operational activities, even
though the NRC may not have found that any one of the prescribed
acceptance criteria have been met.
(c)(1) The licensee shall notify the NRC that the prescribed
inspections, tests, and analyses have been performed and that the
prescribed acceptance criteria have been met. The notification must
contain sufficient information to demonstrate that the prescribed
inspections, tests, and analyses have been performed and that the
prescribed acceptance criteria have been met.
(2) If the licensee has not provided, by the date 225 days before
the scheduled date for initial loading of fuel, the notification
required by paragraph (c)(1) of this section for all ITAAC, then the
licensee shall notify the NRC that the prescribed inspections, tests,
or analyses for all uncompleted ITAAC will be performed and that the
prescribed acceptance criteria will be met prior to operation. The
notification must be provided no later than the date 225 days before
the scheduled date for initial loading of fuel, and must provide
sufficient information to demonstrate that the prescribed inspections,
tests, or analyses will be performed and the prescribed acceptance
criteria for the uncompleted ITAAC will be met, including, but not
limited to, a description of the specific procedures and analytical
methods to be used for performing the prescribed inspections, tests,
and analyses and determining that the prescribed acceptance criteria
have been met.
(d)(1) In the event that an activity is subject to an ITAAC derived
from a referenced standard design certification and the licensee has
not demonstrated that the ITAAC has been met, the licensee may take
corrective actions to successfully complete that ITAAC or request an
exemption from the standard design certification ITAAC, as applicable.
A request for an exemption must also be accompanied by a request for a
license amendment under Sec. 52.98(f).
(2) In the event that an activity is subject to an ITAAC not
derived from a referenced standard design certification and the
licensee has not demonstrated that the ITAAC has been met, the licensee
may take corrective actions to successfully complete that ITAAC or
request a license amendment under Sec. 52.98(f).
(e) The NRC shall ensure that the prescribed inspections, tests,
and analyses in the ITAAC are performed.
(1) At appropriate intervals until the last date for submission of
requests for hearing under Sec. 52.103(a), the NRC shall publish
notices in the Federal Register of the NRC staff's determination of the
successful completion of inspections, tests, and analyses.
(2) The NRC shall make publicly available the licensee
notifications under paragraph (c)(1), and, no later than the date of
publication of the notice of intended operation required by Sec.
52.103(a), make available all licensee notifications under paragraphs
(c)(1) and (c)(2) of this section.
Sec. 52.103 Operation under a combined license.
(a) The licensee shall notify the NRC of its scheduled date for
initial loading of fuel no later than 270 days before the scheduled
date and shall notify the NRC of updates to its schedule every 30 days
thereafter. Not less than 180 days before the date scheduled for
initial loading of fuel into a plant by a licensee that has been issued
a combined license under this part, the Commission shall publish notice
of intended operation in the Federal Register. The notice must provide
that any person whose interest may be affected by operation of the
plant may, within 60 days, request that the Commission hold a hearing
on whether the facility as constructed complies, or on completion will
comply, with the acceptance criteria in the combined license, except
that a hearing shall not be granted for those ITAAC which the
Commission found were met under Sec. 52.97(a)(2).
(b) A request for hearing under paragraph (a) of this section must
show, prima facie, that--
(1) One or more of the acceptance criteria of the ITAAC in the
combined license have not been, or will not be, met; and
(2) The specific operational consequences of nonconformance that
would be contrary to providing reasonable assurance of adequate
protection of the public health and safety.
(c) The Commission, acting as the presiding officer, shall
determine whether to grant or deny the request for hearing in
accordance with the applicable requirements of 10 CFR 2.309. If the
Commission grants the request, the Commission, acting as the presiding
officer, shall determine whether during a period of interim operation
there will be reasonable assurance of adequate protection to the public
health and safety. The Commission's determination must consider the
petitioner's prima facie showing and any answers thereto. If the
Commission determines there is such reasonable assurance, it shall
allow operation during an interim period under the combined license.
(d) The Commission, in its discretion, shall determine appropriate
hearing procedures, whether informal or formal adjudicatory, for any
hearing under paragraph (a) of this section, and shall state its
reasons therefore.
(e) The Commission shall, to the maximum possible extent, render a
decision on issues raised by the hearing request within 180 days of the
publication of the notice provided by paragraph (a) of this section or
by the anticipated date for initial loading of fuel into the reactor,
whichever is later.
(f) A petition to modify the terms and conditions of the combined
license will be processed as a request for action in accordance with 10
CFR 2.206. The petitioner shall file the petition with the Secretary of
the Commission. Before the licensed activity allegedly affected by the
petition (fuel loading, low power testing, etc.) commences, the
Commission shall determine whether any immediate action is required. If
the petition is granted, then an appropriate order will be issued. Fuel
loading and operation under the combined license will not be affected
by the granting of the petition unless the order is made immediately
effective.
[[Page 49537]]
(g) The licensee shall not operate the facility until the
Commission makes a finding that the acceptance criteria in the combined
license are met, except for those acceptance criteria that the
Commission found were met under Sec. 52.97(a)(2). If the combined
license is for a modular design, each reactor module may require a
separate finding as construction proceeds.
(h) After the Commission has made the finding in paragraph (g) of
this section, the ITAAC do not, by virtue of their inclusion in the
combined license, constitute regulatory requirements either for
licensees or for renewal of the license; except for the specific ITAAC
for which the Commission has granted a hearing under paragraph (a) of
this section, all ITAAC expire upon final Commission action in the
proceeding. However, subsequent changes to the facility or procedures
described in the final safety analysis report (as updated) must comply
with the requirements in Sec. Sec. 52.98(e) or (f), as applicable.
Sec. 52.104 Duration of combined license.
A combined license is issued for a specified period not to exceed
40 years from the date on which the Commission makes a finding that
acceptance criteria are met under Sec. 52.103(g) or allowing operation
during an interim period under the combined license under Sec.
52.103(c).
Sec. 52.105 Transfer of combined license.
A combined license may be transferred in accordance with Sec.
50.80 of this chapter.
Sec. 52.107 Application for renewal.
The filing of an application for a renewed license must be in
accordance with 10 CFR part 54.
Sec. 52.109 Continuation of combined license.
Each combined license for a facility that has permanently ceased
operations, continues in effect beyond the expiration date to authorize
ownership and possession of the production or utilization facility,
until the Commission notifies the licensee in writing that the license
is terminated. During this period of continued effectiveness the
licensee shall--
(1) Take actions necessary to decommission and decontaminate the
facility and continue to maintain the facility, including, where
applicable, the storage, control and maintenance of the spent fuel, in
a safe condition; and
(2) Conduct activities in accordance with all other restrictions
applicable to the facility in accordance with the NRC's regulations and
the provisions of the combined license for the facility.
Sec. 52.110 Termination of license.
(a)(1) When a licensee has determined to permanently cease
operations the licensee shall, within 30 days, submit a written
certification to the NRC, consistent with the requirements of Sec.
52.3(b)(8);
(2) Once fuel has been permanently removed from the reactor vessel,
the licensee shall submit a written certification to the NRC that meets
the requirements of Sec. 52.3(b)(9); and
(3) For licensees whose licenses have been permanently modified to
allow possession but not operation of the facility, before September
27, 2007, the certification required in paragraph (a)(1) of this
section shall be deemed to have been submitted.
(b) Upon docketing of the certifications for permanent cessation of
operations and permanent removal of fuel from the reactor vessel, or
when a final legally effective order to permanently cease operations
has come into effect, the 10 CFR part 52 license no longer authorizes
operation of the reactor or emplacement or retention of fuel into the
reactor vessel.
(c) Decommissioning will be completed within 60 years of permanent
cessation of operations. Completion of decommissioning beyond 60 years
will be approved by the Commission only when necessary to protect
public health and safety. Factors that will be considered by the
Commission in evaluating an alternative that provides for completion of
decommissioning beyond 60 years of permanent cessation of operations
include unavailability of waste disposal capacity and other site-
specific factors affecting the licensee's capability to carry out
decommissioning, including presence of other nuclear facilities at the
site.
(d)(1) Before or within 2 years following permanent cessation of
operations, the licensee shall submit a post-shutdown decommissioning
activities report (PSDAR) to the NRC, and a copy to the affected
State(s). The report must include a description of the planned
decommissioning activities along with a schedule for their
accomplishment, an estimate of expected costs, and a discussion that
provides the reasons for concluding that the environmental impacts
associated with site-specific decommissioning activities will be
bounded by appropriate previously issued environmental impact
statements.
(2) The NRC shall notice receipt of the PSDAR and make the PSDAR
available for public comment. The NRC shall also schedule a public
meeting in the vicinity of the licensee's facility upon receipt of the
PSDAR. The NRC shall publish a document in the Federal Register and in
a forum, such as local newspapers, that is readily accessible to
individuals in the vicinity of the site, announcing the date, time and
location of the meeting, along with a brief description of the purpose
of the meeting.
(e) Licensees shall not perform any major decommissioning
activities, as defined in Sec. 50.2 of this chapter, until 90 days
after the NRC has received the licensee's PSDAR submittal and until
certifications of permanent cessation of operations and permanent
removal of fuel from the reactor vessel, as required under Sec.
52.110(a)(1), have been submitted.
(f) Licensees shall not perform any decommissioning activities, as
defined in Sec. 52.1, that--
(1) Foreclose release of the site for possible unrestricted use;
(2) Result in significant environmental impacts not previously
reviewed; or
(3) Result in there no longer being reasonable assurance that
adequate funds will be available for decommissioning.
(g) In taking actions permitted under Sec. 50.59 of this chapter
following submittal of the PSDAR, the licensee shall notify the NRC in
writing and send a copy to the affected State(s), before performing any
decommissioning activity inconsistent with, or making any significant
schedule change from, those actions and schedules described in the
PSDAR, including changes that significantly increase the
decommissioning cost.
(h)(1) Decommissioning trust funds may be used by licensees if--
(i) The withdrawals are for expenses for legitimate decommissioning
activities consistent with the definition of decommissioning in Sec.
52.1;
(ii) The expenditure would not reduce the value of the
decommissioning trust below an amount necessary to place and maintain
the reactor in a safe storage condition if unforeseen conditions or
expenses arise and;
(iii) The withdrawals would not inhibit the ability of the licensee
to complete funding of any shortfalls in the decommissioning trust
needed to ensure the availability of funds to ultimately release the
site and terminate the license.
(2) Initially, 3 percent of the generic amount specified in Sec.
50.75 of this chapter may be used for decommissioning planning. For
licensees that have submitted the certifications required under Sec.
52.110(a) and commencing 90 days after the NRC has received the PSDAR,
an additional
[[Page 49538]]
20 percent may be used. A site-specific decommissioning cost estimate
must be submitted to the NRC before the licensee may use any funding in
excess of these amounts.
(3) Within 2 years following permanent cessation of operations, if
not already submitted, the licensee shall submit a site-specific
decommissioning cost estimate.
(4) For decommissioning activities that delay completion of
decommissioning by including a period of storage or surveillance, the
licensee shall provide a means of adjusting cost estimates and
associated funding levels over the storage or surveillance period.
(i) All power reactor licensees must submit an application for
termination of license. The application for termination of license must
be accompanied or preceded by a license termination plan to be
submitted for NRC approval.
(1) The license termination plan must be a supplement to the FSAR
or equivalent and must be submitted at least 2 years before termination
of the license date.
(2) The license termination plan must include--
(i) A site characterization;
(ii) Identification of remaining dismantlement activities;
(iii) Plans for site remediation;
(iv) Detailed plans for the final radiation survey;
(v) A description of the end use of the site, if restricted;
(vi) An updated site-specific estimate of remaining decommissioning
costs;
(vii) A supplement to the environmental report, under Sec. 51.53
of this chapter, describing any new information or significant
environmental change associated with the licensee's proposed
termination activities; and
(viii) Identification of parts, if any, of the facility or site
that were released for use before approval of the license termination
plan.
(3) The NRC shall notice receipt of the license termination plan
and make the license termination plan available for public comment. The
NRC shall also schedule a public meeting in the vicinity of the
licensee's facility upon receipt of the license termination plan. The
NRC shall publish a document in the Federal Register and in a forum,
such as local newspapers, which is readily accessible to individuals in
the vicinity of the site, announcing the date, time and location of the
meeting, along with a brief description of the purpose of the meeting.
(j) If the license termination plan demonstrates that the remainder
of decommissioning activities will be performed in accordance with the
regulations in this chapter, will not be inimical to the common defense
and security or to the health and safety of the public, and will not
have a significant effect on the quality of the environment and after
notice to interested persons, the Commission shall approve the plan, by
license amendment, subject to terms and conditions as it deems
appropriate and necessary and authorize implementation of the license
termination plan.
(k) The Commission shall terminate the license if it determines
that--
(1) The remaining dismantlement has been performed in accordance
with the approved license termination plan; and
(2) The final radiation survey and associated documentation,
including an assessment of dose contributions associated with parts
released for use before approval of the license termination plan,
demonstrate that the facility and site have met the criteria for
decommissioning in subpart E to 10 CFR part 20.
(l) For a facility that has permanently ceased operation before the
expiration of its license, the collection period for any shortfall of
funds will be determined, upon application by the licensee, on a case-
by-case basis taking into account the specific financial situation of
each licensee.
Subpart D--Reserved
Subpart E--Standard Design Approvals
Sec. 52.131 Scope of subpart.
This subpart sets out procedures for the filing, NRC staff review,
and referral to the Advisory Committee on Reactor Safeguards of
standard designs for a nuclear power reactor of the type described in
Sec. 50.22 of this chapter or major portions thereof.
Sec. 52.133 Relationship to other subparts.
(a) This subpart applies to a person that requests a standard
design approval from the NRC staff separately from an application for a
construction permit filed under 10 CFR part 50 or a combined license
filed under subpart C of this part. An applicant for a construction
permit or combined license may reference a standard design approval.
(b) Subpart B of this part governs the certification by rulemaking
of the design of a nuclear power plant. Subpart B may be used
independently of the provisions in this subpart.
(c) Subpart F of this part governs the issuance of licenses to
manufacture nuclear power reactors to be installed and operated at
sites not identified in the manufacturing license application. Subpart
F of this part may be used independently of the provisions in this
subpart.
Sec. 52.135 Filing of applications.
(a) Any person may submit a proposed standard design for a nuclear
power reactor of the type described in 10 CFR 50.22 to the NRC staff
for its review. The submittal may consist of either the final design
for the entire facility or the final design of major portions thereof.
(b) The submittal for review of the proposed standard design must
be made in the same manner and in the same number of copies as provided
in 10 CFR 50.30 and 52.3 for license applications.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Sec. 52.136 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (d) and (j).
Sec. 52.137 Contents of applications; technical information.
If the applicant seeks review of a major portion of a standard
design, the application need only contain the information required by
this section to the extent the requirements are applicable to the major
portion of the standard design for which NRC staff approval is sought.
(a) The application must contain a final safety analysis report
that describes the facility, presents the design bases and the limits
on its operation, and presents a safety analysis of the structures,
systems, and components and of the facility, or major portion thereof,
and must include the following information:
(1) The site parameters postulated for the design, and an analysis
and evaluation of the design in terms of those site parameters;
(2) A description and analysis of the SSCs of the facility, with
emphasis upon performance requirements, the bases, with technical
justification, upon which the requirements have been established, and
the evaluations required to show that safety functions will be
accomplished. It is expected that the standard plant will reflect
through its design, construction, and operation an extremely low
probability for accidents that could result in the release of
significant quantities of radioactive fission products. The description
shall be sufficient to permit understanding of the system designs and
their
[[Page 49539]]
relationship to the safety evaluations. Items such as the reactor core,
reactor coolant system, instrumentation and control systems, electrical
systems, containment system, other engineered safety features,
auxiliary and emergency systems, power conversion systems, radioactive
waste handling systems, and fuel handling systems shall be discussed
insofar as they are pertinent. The following power reactor design
characteristics will be taken into consideration by the Commission:
(i) Intended use of the reactor including the proposed maximum
power level and the nature and inventory of contained radioactive
materials;
(ii) The extent to which generally accepted engineering standards
are applied to the design of the reactor;
(iii) The extent to which the reactor incorporates unique, unusual
or enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials; and
(iv) The safety features that are to be engineered into the
facility and those barriers that must be breached as a result of an
accident before a release of radioactive material to the environment
can occur. Special attention must be directed to plant design features
intended to mitigate the radiological consequences of accidents. In
performing this assessment, an applicant shall assume a fission product
release \9\ from the core into the containment assuming that the
facility is operated at the ultimate power level contemplated. The
applicant shall perform an evaluation and analysis of the postulated
fission product release, using the expected demonstrable containment
leak rate and any fission product cleanup systems intended to mitigate
the consequences of the accidents, together with applicable postulated
site parameters, including site meteorology, to evaluate the offsite
radiological consequences. The evaluation must determine that:
---------------------------------------------------------------------------
\9\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events. These accidents have generally been assumed to
result in substantial meltdown of the core with subsequent release
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(A) An individual located at any point on the boundary of the
exclusion area for any 2-hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \10\ total effective dose equivalent (TEDE); and
---------------------------------------------------------------------------
\10\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose
for radiation workers which, according to NCRP recommendations at
the time could be disregarded in the determination of their
radiation exposure status (see NBS Handbook 69 dated June 5, 1959).
However, its use is not intended to imply that this number
constitutes an acceptable limit for an emergency dose to the public
under accident conditions. Rather, this dose value has been set
forth in this section as a reference value, which can be used in the
evaluation of plant design features with respect to postulated
reactor accidents, to assure that these designs provide assurance of
low risk of public exposure to radiation, in the event of an
accident.
---------------------------------------------------------------------------
(B) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
TEDE;
(3) The design of the facility including:
(i) The principal design criteria for the facility. Appendix A to
10 CFR part 50, general design criteria (GDC), establishes minimum
requirements for the principal design criteria for water-cooled nuclear
power plants similar in design and location to plants for which
construction permits have previously been issued by the Commission and
provides guidance to applicants in establishing principal design
criteria for other types of nuclear power units;
(ii) The design bases and the relation of the design bases to the
principal design criteria; and
(iii) Information relative to materials of construction, general
arrangement, and approximate dimensions, sufficient to provide
reasonable assurance that the design will conform to the design bases
with adequate margin for safety;
(4) An analysis and evaluation of the design and performance of SSC
with the objective of assessing the risk to public health and safety
resulting from operation of the facility and including determination of
the margins of safety during normal operations and transient conditions
anticipated during the life of the facility, and the adequacy of SSCs
provided for the prevention of accidents and the mitigation of the
consequences of accidents. Analysis and evaluation of ECCS cooling
performance and the need for high-point vents following postulated
loss-of-coolant accidents shall be performed in accordance with the
requirements of 10 CFR 50.46 and 50.46a;
(5) The kinds and quantities of radioactive materials expected to
be produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter;
(6) The information required by Sec. 20.1406 of this chapter;
(7) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(8) The information necessary to demonstrate compliance with any
technically relevant portions of the Three Mile Island requirements set
forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix),
and (f)(3)(v) of 10 CFR 50.34(f);
(9) For applications for light-water-cooled nuclear power plants,
an evaluation of the standard plant design against the Standard Review
Plan (SRP) revision in effect 6 months before the docket date of the
application. The evaluation required by this section shall include an
identification and description of all differences in design features,
analytical techniques, and procedural measures proposed for the design
and those corresponding features, techniques, and measures given in the
SRP acceptance criteria. Where a difference exists, the evaluation
shall discuss how the proposed alternative provides an acceptable
method of complying with the Commission's regulations, or portions
thereof, that underlie the corresponding SRP acceptance criteria. The
SRP is not a substitute for the regulations, and compliance is not a
requirement;
(10) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations described in 10 CFR
50.34a(e);
(11) The information pertaining to design features that affect
plans for coping with emergencies in the operation of the reactor
facility or a major portion thereof;
(12) An analysis and description of the equipment and systems for
combustible gas control as required by Sec. 50.44 of this chapter;
(13) The list of electric equipment important to safety that is
required by 10 CFR 50.49(d);
(14) A description of protection provided against pressurized
thermal shock events, including projected values of the reference
temperature for reactor vessel beltline materials as defined in 10 CFR
50.60 and 50.61;
(15) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram (ATWS) events in Sec. 50.62;
[[Page 49540]]
(16) The coping analysis, and any design features necessary to
address station blackout, as described in Sec. 50.63 of this chapter;
(17) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68(b)(2)-(b)(4);
(18) A description and analysis of the fire protection design
features for the standard plant necessary to comply with part 50,
appendix A, GDC 3, and Sec. 50.48 of this chapter;
(19) A description of the quality assurance program applied to the
design of the SSCs of the facility. Appendix B to 10 CFR part 50,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' sets forth the requirements for quality
assurance programs for nuclear power plants. The description of the
quality assurance program for a nuclear power plant shall include a
discussion of how the applicable requirements of appendix B to 10 CFR
part 50 were satisfied;
(20) The information necessary to demonstrate that the standard
plant complies with the earthquake engineering criteria in 10 CFR part
50, appendix S;
(21) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(22) The information necessary to demonstrate how operating
experience insights have been incorporated into the plant design;
(23) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection,
hydrogen combustion, and containment bypass;
(24) A description, analysis, and evaluation of the interfaces
between the standard design and the balance of the nuclear power plant;
and
(25) A description of the design-specific probabilistic risk
assessment and its results.
(b) An application for approval of a standard design, which differs
significantly from the light-water reactor designs of plants that have
been licensed and in commercial operation before April 18, 1989, or
uses simplified, inherent, passive, or other innovative means to
accomplish its safety functions, must meet the requirements of 10 CFR
50.43(e).
Sec. 52.139 Standards for review of applications.
Applications filed under this subpart will be reviewed for
compliance with the standards set out in 10 CFR parts 20, 50 and its
appendices, and 10 CFR parts 73 and 100.
Sec. 52.141 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application which
concern safety.
Sec. 52.143 Staff approval of design.
Upon completion of its review of a submittal under this subpart and
receipt of a report by the Advisory Committee on Reactor Safeguards
under Sec. 52.141 of this subpart, the NRC staff shall publish a
determination in the Federal Register as to whether or not the design
is acceptable, subject to appropriate terms and conditions, and make an
analysis of the design in the form of a report available at the NRC Web
site, http://www.nrc.gov.
Sec. 52.145 Finality of standard design approvals; information
requests.
(a) An approved design must be used by and relied upon by the NRC
staff and the ACRS in their review of any individual facility license
application that incorporates by reference a standard design approved
in accordance with this paragraph unless there exists significant new
information that substantially affects the earlier determination or
other good cause.
(b) The determination and report by the NRC staff do not constitute
a commitment to issue a permit or license, or in any way affect the
authority of the Commission, Atomic Safety and Licensing Board Panel,
or presiding officers in any proceeding under part 2 of this chapter.
(c) Except for information requests seeking to verify compliance
with the current licensing basis of the standard design approval,
information requests to the holder of a standard design approval must
be evaluated before issuance to ensure that the burden to be imposed on
respondents is justified in view of the potential safety significance
of the issue to be addressed in the requested information. Each
evaluation performed by the NRC staff must be in accordance with 10 CFR
50.54(f) and must be approved by the Executive Director for Operations
or his or her designee before issuance of the request.
Sec. 52.147 Duration of design approval.
A standard design approval issued under this subpart is valid for
15 years from the date of issuance and may not be renewed. A design
approval continues to be valid beyond the date of expiration in any
proceeding on an application for a construction permit or an operating
license under part 50 or a combined license or manufacturing license
under part 52 that references the final design approval and is docketed
before the date of expiration of the design approval.
Subpart F--Manufacturing Licenses
Sec. 52.151 Scope of subpart.
This subpart sets out the requirements and procedures applicable to
Commission issuance of a license authorizing manufacture of nuclear
power reactors to be installed at sites not identified in the
manufacturing license application.
Sec. 52.153 Relationship to other subparts.
(a) A nuclear power reactor manufactured under a manufacturing
license issued under this subpart may only be transported to and
installed at a site for which either a construction permit under part
50 of this chapter or a combined license under subpart C of this part
has been issued.
(b) Subpart B of this part governs the certification by rulemaking
of the design of standard nuclear power facilities. Subpart E of this
part governs the NRC staff review and approval of standard designs for
a nuclear power facility. A manufacturing license applicant may
reference a standard design certification or a standard design approval
in its application. These subparts may also be used independently of
the provisions in this subpart.
Sec. 52.155 Filing of applications.
(a) Any person, except one excluded by 10 CFR 50.38, may file an
application for a manufacturing license under this subpart with the
Director of New Reactors or the Director of Nuclear Reactor Regulation,
as appropriate.
(b) The application must comply with the applicable filing
requirements of Sec. Sec. 52.3 and 50.30 of this chapter.
(c) The fees associated with the filing and review of the
application are set forth in 10 CFR part 170.
Sec. 52.156 Contents of applications; general information.
The application must contain all of the information required by 10
CFR 50.33(a) through (d), and (j).
Sec. 52.157 Contents of applications; technical information in final
safety analysis report.
The application must contain a final safety analysis report
containing the information set forth below, with a level
[[Page 49541]]
of design information sufficient to enable the Commission to judge the
applicant's proposed means of assuring that the manufacturing conforms
to the design and to reach a final conclusion on all safety questions
associated with the design, permit the preparation of construction and
installation specifications by an applicant who seeks to use the
manufactured reactor, and permit the preparation of acceptance and
inspection requirements by the NRC:
(a) The principal design criteria for the reactor to be
manufactured. Appendix A of 10 CFR part 50, ``General Design Criteria
for Nuclear Power Plants,'' establishes minimum requirements for the
principal design criteria for water-cooled nuclear power plants similar
in design and location to plants for which construction permits have
previously been issued by the Commission and provides guidance to
applicants in establishing principal design criteria for other types of
nuclear power units;
(b) The design bases and the relation of the design bases to the
principal design criteria;
(c) A description and analysis of the structures, systems, and
components of the reactor to be manufactured, with emphasis upon the
materials of manufacture, performance requirements, the bases, with
technical justification therefor, upon which the performance
requirements have been established, and the evaluations required to
show that safety functions will be accomplished. The description shall
be sufficient to permit understanding of the system designs and their
relationship to safety evaluations. Items such as the reactor core,
reactor coolant system, instrumentation and control systems, electrical
systems, containment system, other engineered safety features,
auxiliary and emergency systems, power conversion systems, radioactive
waste handling systems, and fuel handling systems shall be discussed
insofar as they are pertinent. The following power reactor design
characteristics will be taken into consideration by the Commission:
(1) Intended use of the manufactured reactor including the proposed
maximum power level and the nature and inventory of contained
radioactive materials;
(2) The extent to which generally accepted engineering standards
are applied to the design of the reactor; and
(3) The extent to which the reactor incorporates unique, unusual or
enhanced safety features having a significant bearing on the
probability or consequences of accidental release of radioactive
materials;
(d) The safety features that are engineered into the reactor and
those barriers that must be breached as a result of an accident before
a release of radioactive material to the environment can occur. Special
attention must be directed to reactor design features intended to
mitigate the radiological consequences of accidents. In performing this
assessment, an applicant shall assume a fission product release \11\
from the core into the containment assuming that the facility is
operated at the ultimate power level contemplated. The applicant shall
perform an evaluation and analysis of the postulated fission product
release, using the expected demonstrable containment leak rate and any
fission product cleanup systems intended to mitigate the consequences
of the accidents, together with applicable postulated site parameters,
including site meteorology, to evaluate the offsite radiological
consequences. The evaluation must determine that:
---------------------------------------------------------------------------
\11\ The fission product release assumed for this evaluation
should be based upon a major accident, hypothesized for purposes of
site analysis or postulated from considerations of possible
accidental events. These accidents have generally been assumed to
result in substantial meltdown of the core with subsequent release
into the containment of appreciable quantities of fission products.
---------------------------------------------------------------------------
(1) An individual located at any point on the boundary of the
exclusion area for any 2 hour period following the onset of the
postulated fission product release, would not receive a radiation dose
in excess of 25 rem \12\ total effective dose equivalent (TEDE);
---------------------------------------------------------------------------
\12\ A whole body dose of 25 rem has been stated to correspond
numerically to the once in a lifetime accidental or emergency dose
for radiation workers which, according to NCRP recommendations at
the time could be disregarded in the determination of their
radiation exposure status (see NBS Handbook 69 dated June 5, 1959).
However, its use is not intended to imply that this number
constitutes an acceptable limit for an emergency dose to the public
under accident conditions. Rather, this dose value has been set
forth in this section as a reference value, which can be used in the
evaluation of plant design features with respect to postulated
reactor accidents, to assure that these designs provide assurance of
low risk of public exposure to radiation, in the event of an
accident.
---------------------------------------------------------------------------
(2) An individual located at any point on the outer boundary of the
low population zone, who is exposed to the radioactive cloud resulting
from the postulated fission product release (during the entire period
of its passage) would not receive a radiation dose in excess of 25 rem
TEDE; and
(e) The kinds and quantities of radioactive materials expected to
be produced in the operation and the means for controlling and limiting
radioactive effluents and radiation exposures within the limits set
forth in part 20 of this chapter.
(f) Information necessary to establish that the design of the
reactor to be manufactured complies with the technical requirements in
10 CFR Chapter I, including:
(1) An analysis and evaluation of the design and performance of
structures, systems, and components with the objective of assessing the
risk to public health and safety resulting from operation of the
facility and including determination of the margins of safety during
normal operations and transient conditions anticipated during the life
of the facility, and the adequacy of structures, systems, and
components provided for the prevention of accidents and the mitigation
of the consequences of accidents. Analysis and evaluation of ECCS
cooling performance and the need for high-point vents following
postulated loss-of-coolant accidents shall be performed in accordance
with the requirements of Sec. Sec. 50.46 and 50.46a of this chapter;
(2) A description and analysis of the fire protection design
features for the reactor necessary to comply with 10 CFR part 50,
appendix A, GDC 3 and Sec. 50.48 of this chapter;
(3) A description of protection provided against pressurized
thermal shock events, including projected values of the reference
temperature for reactor vessel beltline materials as defined in
Sec. Sec. 50.60 and 50.61 of this chapter;
(4) An analysis and description of the equipment and systems for
combustible gas control as required by Sec. 50.44 of this chapter;
(5) The coping analysis, and any design features necessary to
address station blackout, as described in Sec. 50.63 of this chapter;
(6) The list of electric equipment important to safety that is
required by 10 CFR 50.49(d);
(7) Information demonstrating how the applicant will comply with
requirements for reduction of risk from anticipated transients without
scram (ATWS) events in Sec. 50.62;
(8) Information demonstrating how the applicant will comply with
requirements for criticality accidents in Sec. 50.68(b)(2)-(b)(4);
(9) The information required by Sec. 20.1406 of this chapter;
(10) [Reserved];
(11) The information with respect to the design of equipment to
maintain control over radioactive materials in gaseous and liquid
effluents produced during normal reactor operations, as described in
Sec. 50.34a(e) of this chapter;
[[Page 49542]]
(12) The information necessary to demonstrate compliance with any
technically relevant portions of the Three Mile Island requirements set
forth in Sec. 50.34(f) of this chapter, except paragraphs (f)(1)(xii),
(f)(2)(ix), and (f)(3)(v);
(13) If the applicant seeks to use risk-informed treatment of SSCs
in accordance with Sec. 50.69 of this chapter, the information
required by Sec. 50.69(b)(2) of this chapter;
(14) The information necessary to demonstrate that the manufactured
reactor complies with the earthquake engineering criteria in appendix S
to 10 CFR part 50;
(15) Information sufficient to demonstrate compliance with the
applicable requirements regarding testing, analysis, and prototypes as
set forth in Sec. 50.43(e) of this chapter;
(16) The technical qualifications of the applicant to engage in the
proposed activities in accordance with the regulations in this chapter;
(17) A description of the quality assurance program applied to the
design, and to be applied to the manufacture of, the structures,
systems, and components of the reactor. Appendix B to 10 CFR part 50,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants,'' sets forth the requirements for quality
assurance programs for nuclear power plants. The description of the
quality assurance program must include a discussion of how the
applicable requirements of appendix B to 10 CFR part 50 have been and
will be satisfied; and
(18) Proposed technical specifications applicable to the reactor
being manufactured, prepared in accordance with the requirements of
Sec. Sec. 50.36 and 50.36a of this chapter;
(19) The site parameters postulated for the design, and an analysis
and evaluation of the reactor design in terms of those site parameters;
(20) The interface requirements between the manufactured reactor
and the remaining portions of the nuclear power plant. These
requirements must be sufficiently detailed to allow for completion of
the final safety analysis;
(21) Justification that compliance with the interface requirements
of paragraph (f)(20) of this section is verifiable through inspections,
testing, or analysis. The method to be used for verification of
interface requirements must be included as part of the proposed ITAAC
required by Sec. 52.158(a);
(22) A representative conceptual design for a nuclear power
facility using the manufactured reactor, to aid the NRC in its review
of the final safety analysis required by this section and to permit
assessment of the adequacy of the interface requirements in paragraph
(f)(20) of this section;
(23) For light-water reactor designs, a description and analysis of
design features for the prevention and mitigation of severe accidents,
e.g., challenges to containment integrity caused by core-concrete
interaction, steam explosion, high-pressure core melt ejection,
hydrogen combustion, and containment bypass;
(24) [Reserved];
(25) If the reactor is to be used in modular plant design, a
description of the possible operating configurations of the reactor
modules with common systems, interface requirements, and system
interactions. The final safety analysis must also account for
differences among the configurations, including any restrictions that
will be necessary during the construction and startup of a given module
to ensure the safe operation of any module already operating;
(26) A description of the management plan for design and
manufacturing activities, including:
(i) The organizational and management structure singularly
responsible for direction of design and manufacture of the reactor;
(ii) Technical resources directed by the applicant, and the
qualifications requirements;
(iii) Details of the interaction of design and manufacture within
the applicant's organization and the manner by which the applicant will
ensure close integration of the architect engineer and the nuclear
steam supply vendor, as applicable;
(iv) Proposed procedures governing the preparation of the
manufactured reactor for shipping to the site where it is to be
operated, the conduct of shipping, and verifying the condition of the
manufactured reactor upon receipt at the site; and
(v) The degree of top level management oversight and technical
control to be exercised by the applicant during design and manufacture,
including the preparation and implementation of procedures necessary to
guide the effort;
(27) Necessary parameters to be used in developing plans for
preoperational testing and initial operation;
(28) Proposed technical resolutions of those Unresolved Safety
Issues and medium- and high-priority generic safety issues which are
identified in the version of NUREG-0933 current on the date up to 6
months before the docket date of the application and which are
technically relevant to the design;
(29) The information necessary to demonstrate how operating
experience insights have been incorporated into the manufactured
reactor design;
(30) For applications for light-water-cooled nuclear power plants,
an evaluation of the design to be manufactured against the Standard
Review Plan (SRP) revision in effect 6 months before the docket date of
the application. The evaluation required by this section shall include
an identification and description of all differences in design
features, analytical techniques, and procedural measures proposed for
the design and those corresponding features, techniques, and measures
given in the SRP acceptance criteria. Where a difference exists, the
evaluation shall discuss how the proposed alternative provides an
acceptable method of complying with the Commission's regulations, or
portions thereof, that underlie the corresponding SRP acceptance
criteria. The SRP is not a substitute for the regulations, and
compliance is not a requirement; and
(31) A description of the design-specific probabilistic risk
assessment and its results.
Sec. 52.158 Contents of application; additional technical
information.
The application must contain:
(a)(1) Inspections, tests, analyses, and acceptance criteria
(ITAAC). The proposed inspections, tests, and analyses that the
licensee who will be operating the reactor shall perform, and the
acceptance criteria that are necessary and sufficient to provide
reasonable assurance that, if the inspections, tests, and analyses are
performed and the acceptance criteria met:
(i) The reactor has been manufactured in conformity with the
manufacturing license; the provisions of the Act, and the Commission's
rules and regulations; and
(ii) The manufactured reactor will be operated in conformity with
the approved design and any license authorizing operation of the
manufactured reactor.
(2) If the application references a standard design certification,
the ITAAC contained in the certified design must apply to those
portions of the facility design which are covered by the design
certification.
(3) If the application references a standard design certification,
the application may include a notification that a required inspection,
test, or analysis in the design certification ITAAC has been
successfully completed and that the corresponding acceptance
[[Page 49543]]
criterion has been met. The Federal Register notification required by
Sec. 52.163 must indicate that the application includes this
notification.
(b)(1) An environmental report as required by 10 CFR 51.54.
(2) If the manufacturing license application references a standard
design certification, the environmental report need not contain a
discussion of severe accident mitigation design alternatives for the
reactor.
Sec. 52.159 Standards for review of application.
Applications filed under this subpart will be reviewed according to
the applicable standards set out in 10 CFR parts 20, 50 and its
appendices, 51, 73, and 100 and its appendices.
Sec. 52.161 Reserved.
Sec. 52.163 Administrative review of applications; hearings.
A proceeding on a manufacturing license is subject to all
applicable procedural requirements contained in 10 CFR part 2,
including the requirements for docketing in Sec. 2.101(a)(1) through
(4) of this chapter, and the requirements for issuance of a notice of
proposed action in Sec. 2.105 of this chapter, provided, however, that
the designated sections may not be construed to require that the
environmental report or draft or final environmental impact statement
include an assessment of the benefits of constructing and/or operating
the manufactured reactor or an evaluation of alternative energy
sources. All hearings on manufacturing licenses are governed by the
hearing procedures contained in 10 CFR part 2, subparts C, G, L, and N.
Sec. 52.165 Referral to the Advisory Committee on Reactor Safeguards
(ACRS).
The Commission shall refer a copy of the application to the ACRS.
The ACRS shall report on those portions of the application which
concern safety.
Sec. 52.167 Issuance of manufacturing license.
(a) After completing any hearing under Sec. 52.163, and receiving
the report submitted by the ACRS, the Commission may issue a
manufacturing license if the Commission finds that:
(1) Applicable standards and requirements of the Act and the
Commission's regulations have been met;
(2) There is reasonable assurance that the reactor(s) will be
manufactured, and can be transported, incorporated into a nuclear power
plant, and operated in conformity with the manufacturing license, the
provision of the Act, and the Commission's regulations;
(3) The proposed reactor(s) can be incorporated into a nuclear
power plant and operated at sites having characteristics that fall
within the site parameters postulated for the design of the
manufactured reactor(s) without undue risk to the health and safety of
the public;
(4) The applicant is technically qualified to design and
manufacture the proposed nuclear power reactor(s);
(5) The proposed inspections, tests, analyses and acceptance
criteria are necessary and sufficient, within the scope of the
manufacturing license, to provide reasonable assurance that the
manufactured reactor has been manufactured and will be operated in
conformity with the license, the provisions of the Act, and the
Commission's regulations;
(6) The issuance of a license to the applicant will not be inimical
to the common defense and security or to the health and safety of the
public; and
(7) The findings required by subpart A of part 51 of this chapter
have been made.
(b) Each manufacturing license issued under this subpart shall
specify:
(1) Terms and conditions as the Commission deems necessary and
appropriate;
(2) Technical specifications for operation of the manufactured
reactor, as the Commission deems necessary and appropriate;
(3) Site parameters and design characteristics for the manufactured
reactor; and
(4) The interface requirements to be met by the site-specific
elements of the facility, such as the service water intake structure
and the ultimate heat sink, not within the scope of the manufactured
reactor.
(c)(1) A holder of a manufacturing license may not transport or
allow to be removed from the place of manufacture the manufactured
reactor except to the site of a licensee with either a construction
permit under part 50 of this chapter or a combined license under
subpart C of this part. The construction permit or combined license
must authorize the construction of a nuclear power facility using the
manufactured reactor(s).
(2) A holder of a manufacturing license shall include, in any
contract governing the transport of a manufactured reactor from the
place of manufacture to any other location, a provision requiring that
the person or entity transporting the manufactured reactor to comply
with all NRC-approved shipping requirements in the manufacturing
license.
Sec. 52.169 [Reserved].
Sec. 52.171 Finality of manufacturing licenses; information requests.
(a)(1) Notwithstanding any provision in 10 CFR 50.109, during the
term of a manufacturing license the Commission may not modify, rescind,
or impose new requirements on the design of the nuclear power reactor
being manufactured, or the requirements for the manufacture of the
nuclear power reactor, unless the Commission determines that a
modification is necessary to bring the design of the reactor or its
manufacture into compliance with the Commission's requirements
applicable and in effect at the time the manufacturing license was
issued, or to provide reasonable assurance of adequate protection to
public health and safety or common defense and security.
(2) Any modification to the design of a manufactured nuclear power
reactor which is imposed by the Commission under paragraph (a)(1) of
this section will be applied to all reactors manufactured under the
license, including those that have already been transported and sited,
except those reactors to which the modification has been rendered
technically irrelevant by action taken under paragraph (b) of this
section.
(3) In making the findings required for issuance of a construction
permit, operating license, combined license, in any hearing under Sec.
52.103, or in any enforcement hearing other than one initiated by the
Commission under paragraph (a)(1) of this section, for which a nuclear
power reactor manufactured under this subpart is referenced or used,
the Commission shall treat as resolved those matters resolved in the
proceeding on the application for issuance or renewal of the
manufacturing license, including the adequacy of design of the
manufactured reactor, the costs and benefits of severe accident
mitigation design alternatives, and the bases for not incorporating
severe accident mitigation design alternatives into the design of the
reactor to be manufactured.
(b)(1) The holder of a manufacturing license may not make changes
to the design of the nuclear power reactor authorized to be
manufactured without prior Commission approval. The request for a
change to the design must be in the form of an application for a
license amendment, and must meet the requirements of 10 CFR 50.90 and
50.92.
[[Page 49544]]
(2) An applicant or licensee who references or uses a nuclear power
reactor manufactured under a manufacturing license under this subpart
may request a departure from the design characteristics, site
parameters, terms and conditions, or approved design of the
manufactured reactor. The Commission may grant a request only if it
determines that the departure will comply with the requirements of 10
CFR 52.7, and that the special circumstances outweigh any decrease in
safety that may result from the reduction in standardization caused by
the departure. The granting of a departure on request of an applicant
is subject to litigation in the same manner as other issues in the
construction permit or combined license hearing.
(c) Except for information requests seeking to verify compliance
with the current licensing basis of either the manufacturing license or
the manufactured reactor, information requests to the holder of a
manufacturing license or an applicant or licensee using a manufactured
reactor must be evaluated before issuance to ensure that the burden to
be imposed on respondents is justified in view of the potential safety
significance of the issue to be addressed in the requested information.
Each evaluation performed by the NRC staff must be in accordance with
10 CFR 50.54(f) and must be approved by the Executive Director for
Operations or his or her designee before issuance of the request.
Sec. 52.173 Duration of manufacturing license.
A manufacturing license issued under this subpart may be valid for
not less than 5, nor more than 15 years from the date of issuance. A
holder of a manufacturing license may not initiate the manufacture of a
reactor less than 3 years before the expiration of the license even
though a timely application for renewal has been docketed with the NRC.
Upon expiration of the manufacturing license, the manufacture of any
uncompleted reactors must cease unless a timely application for renewal
has been docketed with the NRC.
Sec. 52.175 Transfer of manufacturing license.
A manufacturing license may be transferred in accordance with Sec.
50.80 of this chapter.
Sec. 52.177 Application for renewal.
(a) Not less than 12 months, nor more than 5 years before the
expiration of the manufacturing license, or any later renewal period,
the holder of the manufacturing license may apply for a renewal of the
license. An application for renewal must contain all information
necessary to bring up to date the information and data contained in the
previous application.
(b) The filing of an application for a renewed license must be in
accordance with subpart A of 10 CFR part 2 and 10 CFR 52.3 and 50.30.
(c) A manufacturing license, either original or renewed, for which
a timely application for renewal has been filed, remains in effect
until the Commission has made a final determination on the renewal
application, provided, however, that in accordance with Sec. 52.173,
the holder of a manufacturing license may not begin manufacture of a
reactor less than 3 years before the expiration of the license.
(d) Any person whose interest may be affected by renewal of the
permit may request a hearing on the application for renewal. The
request for a hearing must comply with 10 CFR 2.309. If a hearing is
granted, notice of the hearing will be published in accordance with 10
CFR 2.104.
(e) The Commission shall refer a copy of the application for
renewal to the Advisory Committee on Reactor Safeguards (ACRS). The
ACRS shall report on those portions of the application which concern
safety and shall apply the criteria set forth in Sec. 52.159.
Sec. 52.179 Criteria for renewal.
The Commission may grant the renewal if the Commission determines:
(a) The manufacturing license complies with the Atomic Energy Act
and the Commission's regulations and orders applicable and in effect at
the time the manufacturing license was originally issued; and
(b) Any new requirements the Commission may wish to impose are:
(1) Necessary for adequate protection to public health and safety
or common defense and security;
(2) Necessary for compliance with the Commission's regulations and
orders applicable and in effect at the time the manufacturing license
was originally issued; or
(3) A substantial increase in overall protection of the public
health and safety or the common defense and security to be derived from
the new requirements, and the direct and indirect costs of
implementation of those requirements are justified in view of this
increased protection.
Sec. 52.181 Duration of renewal.
A renewed manufacturing license may be issued for a term of not
less than 5, nor more than 15 years, plus any remaining years on the
manufacturing license then in effect before renewal. The renewed
license shall be subject to the requirements of Sec. Sec. 52.171 and
52.175.
Subpart G--Reserved
Subpart H--Enforcement
Sec. 52.301 Violations.
(a) The Commission may obtain an injunction or other court order to
prevent a violation of the provisions of--
(1) The Atomic Energy Act of 1954, as amended;
(2) Title II of the Energy Reorganization Act of 1974, as amended;
or
(3) A regulation or order issued under those Acts.
(b) The Commission may obtain a court order for the payment of a
civil penalty imposed under Section 234 of the Atomic Energy Act:
(1) For violations of--
(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of
the Atomic Energy Act of 1954, as amended;
(ii) Section 206 of the Energy Reorganization Act;
(iii) Any regulation, or order issued under the sections specified
in paragraph (b)(1)(i) of this section;
(iv) Any term, condition, or limitation of any license issued under
the sections specified in paragraph (b)(1)(i) of this section.
(2) For any violation for which a license may be revoked under
Section 186 of the Atomic Energy Act of 1954, as amended.
Sec. 52.303 Criminal penalties.
(a) Section 223 of the Atomic Energy Act of 1954, as amended,
provides for criminal sanctions for willful violation of, attempted
violation of, or conspiracy to violate, any regulation issued under
Sections 161b, 161i, or 161o of the Act. For purposes of Section 223,
all the regulations in part 52 are issued under one or more of Sections
161b, 161i, or 160o, except for the sections listed in paragraph (b) of
this section.
(b) The regulations in part 52 that are not issued under Sections
161b, 161i, or 161o for the purposes of Section 223 are as follows:
Sec. Sec. 52.0, 52.1, 52.2, 52.3, 52.7, 52.8, 52.9, 52.10, 52.11,
52.12, 52.13, 52.15, 52.16, 52.17, 52.18, 52.21, 52.23, 52.24, 52.27,
52.28, 52.29, 52.31, 52.33, 52.39, 52.41, 52.43, 52.45, 52.46, 52.47,
52.48, 52.51, 52.53, 52.54, 52.55, 52.57, 52.59, 52.61, 52.63, 52.71,
52.73, 52.75, 52.77, 52.79, 52.80, 52.81, 52.83, 52.85, 52.87, 52.93,
52.97, 52.98, 52.103, 52.104, 52.105, 52.107, 52.109, 52.131,
[[Page 49545]]
52.133, 52.135, 52.136, 52.137, 52.139, 52.141, 52.143, 52.145, 52.147,
52.151, 52.153, 52.155, 52.156, 52.157, 52.158, 52.159, 52.161, 52.163,
52.165, 52.167, 52.171, 52.173, 52.175, 52.177, 52.179, 52.181, 52.301,
and 52.303.
Appendix A to Part 52--Design Certification Rule for the U.S. Advanced
Boiling Water Reactor
I. Introduction
Appendix A constitutes the standard design certification for the
U.S. Advanced Boiling Water Reactor (ABWR) design, in accordance
with 10 CFR part 52, subpart B. The applicant for certification of
the U.S. ABWR design was GE Nuclear Energy.
II. Definitions
A. Generic design control document (generic DCD) means the
document containing the Tier 1 and Tier 2 information and generic
technical specifications that is incorporated by reference into this
appendix.
B. Generic technical specifications means the information,
required by 10 CFR 50.36 and 50.36a, for the portion of the plant
that is within the scope of this appendix.
C. Plant-specific DCD means the document, maintained by an
applicant or licensee who references this appendix, consisting of
the information in the generic DCD, as modified and supplemented by
the plant-specific departures and exemptions made under Section VIII
of this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (hereinafter Tier 1 information). The design descriptions,
interface requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by
this appendix (Tier 2 information). Compliance with Tier 2 is
required, but generic changes to and plant-specific departures from
Tier 2 are governed by Section VIII of this appendix. Compliance
with Tier 2 provides a sufficient, but not the only acceptable,
method for complying with Tier 1. Compliance methods differing from
Tier 2 must satisfy the change process in Section VIII of this
appendix. Regardless of these differences, an applicant or licensee
must meet the requirement in Section III.B of this appendix to
reference Tier 2 when referencing Tier 1. Tier 2 information
includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c),
with the exception of generic technical specifications and
conceptual design information;
2. Supporting information on the inspections, tests, and
analyses that will be performed to demonstrate that the acceptance
criteria in the ITAAC have been met; and
3. Combined license (COL) action items (COL license
information), which identify certain matters that must be addressed
in the site-specific portion of the final safety analysis report
(FSAR) by an applicant who references this appendix. These items
constitute information requirements but are not the only acceptable
set of information in the FSAR. An applicant may depart from or omit
these items, provided that the departure or omission is identified
and justified in the FSAR. After issuance of a construction permit
or COL, these items are not requirements for the licensee unless
such items are restated in the FSAR.
F. Tier 2* means the portion of the Tier 2 information,
designated as such in the generic DCD, which is subject to the
change process in Section VIII.B.6 of this appendix. This
designation expires for some Tier 2* information under Section
VIII.B.6.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
(1) Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are
conservative or essentially the same; or
(2) Changing from a method described in the plant-specific DCD
to another method unless that method has been approved by NRC for
the intended application.
H. All other terms in this appendix have the meaning set out in
10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954,
as amended, as applicable.
III. Scope and Contents
A. Tier 1, Tier 2, and the generic technical specifications in
the U.S. ABWR Design Control Document, GE Nuclear Energy, Revision 4
dated March 1997, are approved for incorporation by reference by the
Director of the Office of the Federal Register in accordance with 5
U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD may be
obtained from the National Technical Information Service, 5285 Port
Royal Road, Springfield, Virginia 22161. A copy is available for
examination and copying at the NRC Public Document Room located at
One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland 20852. Copies are also available for examination
at the NRC Library located at Two White Flint North, 11545 Rockville
Pike, Rockville, Maryland 20582 and the Office of the Federal
Register, 800 North Capitol Street, NW., Suite 700, Washington, DC.
B. An applicant or licensee referencing this appendix, in
accordance with Section IV of this appendix, shall incorporate by
reference and comply with the requirements of this appendix,
including Tier 1, Tier 2, and the generic technical specifications
except as otherwise provided in this appendix. Conceptual design
information, as set forth in the generic DCD, and the ``Technical
Support Document for the ABWR'' are not part of this appendix. Tier
2 references to the probabilistic risk assessment (PRA) in the ABWR
standard safety analysis report do not incorporate the PRA into Tier
2.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD,
then Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the U.S. ABWR design or
NUREG-1503, ``Final Safety Evaluation Report related to the
Certification of the Advanced Boiling Water Reactor Design'' (FSER),
and Supplement No. 1, then the generic DCD controls.
E. Design activities for structures, systems, and components
that are wholly outside the scope of this appendix may be performed
using site characteristics, provided the design activities do not
affect the DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference
this appendix shall, in addition to complying with the requirements
of 10 CFR 52.77, 52.79, and 52.80, comply with the following
requirements:
1. Incorporate by reference, as part of its application, this
appendix;
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information
and using the same organization and numbering as the generic DCD for
the U.S. ABWR design, as modified and supplemented by the
applicant's exemptions and departures;
b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
c. Plant-specific technical specifications, consisting of the
generic and site-specific technical specifications, that are
required by 10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters
and interface requirements;
e. Information that addresses the COL action items; and
f. Information required by 10 CFR 52.47 that is not within the
scope of this appendix.
3. Include, in the plant-specific DCD, the proprietary
information and safeguards information referenced in the U.S. ABWR
DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the U.S. ABWR design are in 10 CFR parts
20, 50, 73, and 100, codified as of May 2, 1997, that are applicable
and technically relevant, as described in the FSER (NUREG-1503) and
Supplement No. 1.
B. The U.S. ABWR design is exempt from portions of the following
regulations:
1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Separate Plant Safety
Parameter Display Console;
2. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-Accident
Sampling for Boron, Chloride, and Dissolved Gases; and
3. Paragraph (f)(3)(iv) of 10 CFR 50.34--Dedicated Containment
Penetration.
[[Page 49546]]
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the U.S. ABWR design comply with
the provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of
the public. A conclusion that a matter is resolved includes the
finding that additional or alternative structures, systems,
components, design features, design criteria, testing, analyses,
acceptance criteria, or justifications are not necessary for the
U.S. ABWR design.
B. The Commission considers the following matters resolved
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings
for issuance of a combined license, amendment of a combined license,
or renewal of a combined license, proceedings held under 10 CFR
52.103, and enforcement proceedings involving plants referencing
this appendix:
1. All nuclear safety issues, except for the generic technical
specifications and other operational requirements, associated with
the information in the FSER and Supplement No. 1, Tier 1, Tier 2
(including referenced information which the context indicates is
intended as requirements), and the rulemaking record for
certification of the U.S. ABWR design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and
in context, are intended as requirements in the generic DCD for the
U.S. ABWR design;
3. All generic changes to the DCD under and in compliance with
the change processes in Sections VIII.A.1 and VIII.B.1 of this
appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix,
but only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix,
all departures from Tier 2 pursuant to and in compliance with the
change processes in paragraph VIII.B.5 of this appendix that do not
require prior NRC approval, but only for that plant;
7. All environmental issues concerning severe accident
mitigation design alternatives associated with the information in
the NRC's final environmental assessment for the U.S. ABWR design
and Revision 1 of the technical support document for the U.S. ABWR,
dated December 1994, for plants referencing this appendix whose site
parameters are within those specified in the technical support
document.
C. The Commission does not consider operational requirements for
an applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an
applicant or licensee who references this appendix by rule,
regulation, order, or license condition.
D. Except in accordance with the change processes in Section
VIII of this appendix, the Commission may not require an applicant
or licensee who references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures,
systems, components, or design features discussed in the generic
DCD.
E.1. Persons who wish to review proprietary and safeguards
information or other secondary references in the DCD for the U.S.
ABWR design, in order to request or participate in the hearing
required by 10 CFR 52.85 or the hearing provided under 10 CFR
52.103, or to request or participate in any other hearing relating
to this appendix in which interested persons have adjudicatory
hearing rights, shall first request access to such information from
GE Nuclear Energy. The request must state with particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the
public at the NRC Web site, http://www.nrc.gov, and/or at the NRC
Public Document Room, is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to
prepare a request for hearing, the request must be filed no later
than 15 days after publication in the Federal Register of the notice
required either by 10 CFR 52.85 or 10 CFR 52.103. If GE Nuclear
Energy declines to provide the information sought, GE Nuclear Energy
shall send a written response within 10 days of receiving the
request to the requesting person setting forth with particularity
the reasons for its refusal. The person may then request the
Commission (or presiding officer, if a proceeding has been
established) to order disclosure. The person shall include copies of
the original request (and any subsequent clarifying information
provided by the requesting party to the applicant) and the
applicant's response. The Commission and presiding officer shall
base their decisions solely on the person's original request
(including any clarifying information provided by the requesting
person to GE Nuclear Energy), and GE Nuclear Energy's response. The
Commission and presiding officer may order GE Nuclear Energy to
provide access to some or all of the requested information, subject
to an appropriate non-disclosure agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from
June 11, 1997, except as provided for in 10 CFR 52.55(b) and
52.57(b). This appendix remains valid for an applicant or licensee
who references this appendix until the application is withdrawn or
the license expires, including any period of extended operation
under a renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design.
B. Tier 2 information.
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under Sec. Sec. 52.55 or 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to assure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in 10 CFR 52.7 are present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 information. The Commission may
grant such a request only if it determines that the exemption will
comply with the requirements of 10 CFR 52.7. The Commission will
deny a request for an exemption from Tier 2, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design. The grant of an exemption
to an applicant must be subject to litigation in the same manner as
other issues material to the license hearing. The grant of an
exemption to a licensee must be subject to an opportunity for a
hearing in the same manner as license amendments.
5.a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information, Tier 2* information, or the technical specifications,
or requires a license amendment under paragraphs B.5.b or B.5.c
[[Page 49547]]
of this section. When evaluating the proposed departure, an
applicant or licensee shall consider all matters described in the
plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would:
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety previously evaluated in the plant-specific
DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of a SSC important to safety previously evaluated
in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important
to safety with a different result than any evaluated previously in
the plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe
accident previously reviewed and determined to be not credible could
become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously
reviewed.
d. If a departure requires a license amendment pursuant to
paragraphs B.5.b or B.5.c of this section, it is governed by 10 CFR
50.90.
e. A departure from Tier 2 information that is made under
paragraph B.5 of this section does not require an exemption from
this appendix.
f. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.103(a), who believes that an applicant or licensee who
references this appendix has not complied with paragraph VIII.B.5 of
this appendix when departing from Tier 2 information, may petition
the NRC to admit into the proceeding such a contention. In addition
to compliance with the general requirements of 10 CFR 2.309, the
petition must demonstrate that the departure does not comply with
paragraph VIII.B.5 of this appendix. Further, the petition must
demonstrate that the change bears on an asserted noncompliance with
an ITAAC acceptance criterion in the case of a 10 CFR 52.103
preoperational hearing, or that the change bears directly on the
amendment request in the case of a hearing on a license amendment.
Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall
certify the matter directly to the Commission for determination of
the admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6.a. An applicant who references this appendix may not depart
from Tier 2* information, which is designated with italicized text
or brackets and an asterisk in the generic DCD, without NRC
approval. The departure will not be considered a resolved issue,
within the meaning of Section VI of this appendix and 10 CFR
52.63(a)(5).
b. A licensee who references this appendix may not depart from
the following Tier 2* matters without prior NRC approval. A request
for a departure will be treated as a request for a license amendment
under 10 CFR 50.90.
(1) Fuel burnup limit (4.2).
(2) Fuel design evaluation (4.2.3).
(3) Fuel licensing acceptance criteria (appendix 4B).
c. A licensee who references this appendix may not, before the
plant first achieves full power following the finding required by 10
CFR 52.103(g), depart from the following Tier 2* matters except in
accordance with paragraph B.6.b of this section. After the plant
first achieves full power, the following Tier 2* matters revert to
Tier 2 status and are thereafter subject to the departure provisions
in paragraph B.5 of this section.
(1) ASME Boiler & Pressure Vessel Code, Section III.
(2) ACI 349 and ANSI/AISC-690.
(3) Motor-operated valves.
(4) Equipment seismic qualification methods.
(5) Piping design acceptance criteria.
(6) Fuel system and assembly design (4.2), except burnup limit.
(7) Nuclear design (4.3).
(8) Equilibrium cycle and control rod patterns (App. 4A).
(9) Control rod licensing acceptance criteria (App. 4C).
(10) Instrument setpoint methodology.
(11) EMS performance specifications and architecture.
(12) SSLC hardware and software qualification.
(13) Self-test system design testing features and commitments.
(14) Human factors engineering design and implementation
process.
d. Departures from Tier 2* information that are made under
paragraph B.6 of this section do not require an exemption from this
appendix.
C. Operational requirements.
1. Generic changes to generic technical specifications and other
operational requirements that were completely reviewed and approved
in the design certification rulemaking and do not require a change
to a design feature in the generic DCD are governed by the
requirements in 10 CFR 50.109. Generic changes that do require a
change to a design feature in the generic DCD are governed by the
requirements in paragraphs A or B of this section.
2. Generic changes to generic TS and other operational
requirements are applicable to all applicants who reference this
appendix, except those for which the change has been rendered
technically irrelevant by action taken under paragraphs C.3 or C.4
of this section.
3. The Commission may require plant-specific departures on
generic technical specifications and other operational requirements
that were completely reviewed and approved, provided a change to a
design feature in the generic DCD is not required and special
circumstances as defined in 10 CFR 2.335 are present. The Commission
may modify or supplement generic technical specifications and other
operational requirements that were not completely reviewed and
approved or require additional technical specifications and other
operational requirements on a plant-specific basis, provided a
change to a design feature in the generic DCD is not required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other
operational requirements. The Commission may grant such a request
only if it determines that the exemption will comply with the
requirements of 10 CFR 52.7. The grant of an exemption must be
subject to litigation in the same manner as other issues material to
the license hearing.
5. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.103(a), who believes that an operational requirement
approved in the DCD or a technical specification derived from the
generic technical specifications must be changed may petition to
admit into the proceeding such a contention. Such petition must
comply with the general requirements of 10 CFR 2.309 and must
demonstrate why special circumstances as defined in 10 CFR 2.335 are
present, or for compliance with the Commission's regulations in
effect at the time this appendix was approved, as set forth in
Section V of this appendix. Any other party may file a response
thereto. If, on the basis of the petition and any response, the
presiding officer determines that a sufficient showing has been
made, the presiding officer shall certify the matter directly to the
Commission for determination of the admissibility of the contention.
All other issues with respect to the plant-specific technical
specifications or other operational requirements are subject to a
hearing as part of the license proceeding.
6. After issuance of a license, the generic technical
specifications have no further effect on the plant-specific
technical specifications and changes to the plant-specific technical
specifications will be treated as license amendments under 10 CFR
50.90.
[[Page 49548]]
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1. An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities, and a licensee may proceed at its own risk with design,
procurement, construction, and preoperational activities, even
though the NRC may not have found that any particular ITAAC has been
met.
2. The licensee who references this appendix shall notify the
NRC that the required inspections, tests, and analyses in the ITAAC
have been successfully completed and that the corresponding
acceptance criteria have been met.
3. In the event that an activity is subject to an ITAAC, and the
applicant or licensee who references this appendix has not
demonstrated that the ITAAC has been met, the applicant or licensee
may either take corrective actions to successfully complete that
ITAAC, request an exemption from the ITAAC in accordance with
Section VIII of this appendix and 10 CFR 52.97(b), or petition for
rulemaking to amend this appendix by changing the requirements of
the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes
to the ITAAC must meet the requirements of paragraph VIII.A.1 of
this appendix.
B.1. The NRC shall ensure that the required inspections, tests,
and analyses in the ITAAC are performed. The NRC shall verify that
the inspections, tests, and analyses referenced by the licensee have
been successfully completed and, based solely thereon, find the
prescribed acceptance criteria have been met. At appropriate
intervals during construction, the NRC shall publish notices of the
successful completion of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall
find that the acceptance criteria in the ITAAC for the license are
met before fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the
subject of a Sec. 52.103(a) hearing, their expiration will occur
upon final Commission action in such proceeding. However, subsequent
modifications must comply with the Tier 1 and Tier 2 design
descriptions in the plant-specific DCD unless the licensee has
complied with the applicable requirements of 10 CFR 52.98 and
Section VIII of this appendix.
X. Records and Reporting
A. Records.
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and
the generic TS and other operational requirements. The applicant
shall maintain the proprietary and safeguards information referenced
in the generic DCD for the period that this appendix may be
referenced, as specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application
and for the term of the license (including any period of renewal).
3. An applicant or licensee who references this appendix shall
prepare and maintain written evaluations which provide the bases for
the determinations required by Section VIII of this appendix. These
evaluations must be retained throughout the period of application
and for the term of the license (including any period of renewal).
B. Reporting.
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any
plant-specific departures from the DCD, including a summary of the
evaluation of each. This report must be filed in accordance with the
filing requirements applicable to reports in 10 CFR 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its DCD, which reflect the generic changes and the
plant-specific departures from the generic DCD made under Section
VIII of this appendix. These updates must be filed under the filing
requirements applicable to final safety analysis report updates in
10 CFR 52.3 and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and
X.B.2 must be submitted as follows:
a. On the date that an application for a license referencing
this appendix is submitted, the application must include the report
and any updates to the generic DCD.
b. During the interval from the date of application for a
license to the date the Commission makes the finding required by 10
CFR 52.103(g), the report must be submitted semiannually. Updates to
the plant-specific DCD must be submitted annually and may be
submitted along with amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the
final safety analysis report for the facility, at the intervals
required by 10 CFR 50.59(d)(2) and 10 CFR 50.71(e)(4), respectively,
or at shorter intervals as specified in the license.
Appendix B to Part 52--Design Certification Rule for the System 80+
Design
I. Introduction
Appendix B constitutes design certification for the System 80+
\1\ standard plant design, in accordance with 10 CFR part 52,
subpart B. The applicant for certification of the System 80+ design
was Combustion Engineering, Inc. (ABB-CE), which is now Westinghouse
Electric Company LLC.
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\1\ ``System 80+'' is a trademark of Westinghouse Electric
Company LLC.
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II. Definitions
A. Generic design control document (generic DCD) means the
document containing the Tier 1 and Tier 2 information and generic
technical specifications that is incorporated by reference into this
appendix.
B. Generic technical specifications means the information,
required by 10 CFR 50.36 and 50.36a, for the portion of the plant
that is within the scope of this appendix.
C. Plant-specific DCD means the document, maintained by an
applicant or licensee who references this appendix, consisting of
the information in the generic DCD, as modified and supplemented by
the plant-specific departures and exemptions made under Section VIII
of this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (hereinafter Tier 1 information). The design descriptions,
interface requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by
this appendix (Tier 2 information). Compliance with Tier 2 is
required, but generic changes to and plant-specific departures from
Tier 2 are governed by Section VIII of this appendix. Compliance
with Tier 2 provides a sufficient, but not the only acceptable,
method for complying with Tier 1. Compliance methods differing from
Tier 2 must satisfy the change process in Section VIII of this
appendix. Regardless of these differences, an applicant or licensee
must meet the requirement in Section III.B of this appendix to
reference Tier 2 when referencing Tier 1. Tier 2 information
includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c),
with the exception of generic technical specifications and
conceptual design information;
2. Supporting information on the inspections, tests, and
analyses that will be performed to demonstrate that the acceptance
criteria in the ITAAC have been met; and
3. Combined license (COL) action items (COL license
information), which identify certain matters that must be addressed
in the site-specific portion of the final safety analysis report
(FSAR) by an applicant who references this appendix. These items
constitute information requirements but are not the only acceptable
set of information in the FSAR. An applicant may depart from or omit
these items, provided that the departure or omission is identified
and justified in the FSAR. After issuance of a construction permit
or COL, these items are not requirements for the licensee unless
such items are restated in the FSAR.
F. Tier 2* means the portion of the Tier 2 information,
designated as such in the generic DCD, which is subject to the
change process in Section VIII.B.6 of this appendix.
[[Page 49549]]
This designation expires for some Tier 2* information under Section
VIII.B.6 of this appendix.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
(1) Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are
conservative or essentially the same; or
(2) Changing from a method described in the plant-specific DCD
to another method unless that method has been approved by NRC for
the intended application.
H. All other terms in this appendix have the meaning set out in
10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954,
as amended, as applicable.
III. Scope and Contents
A. Tier 1, Tier 2, and the generic technical specifications in
the System 80+ Design Control Document, ABB-CE, with revisions dated
January 1997, are approved for incorporation by reference by the
Director of the Office of the Federal Register in accordance with 5
U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD may be
obtained from the National Technical Information Service, 5285 Port
Royal Road, Springfield, Virginia 22161. A copy is available for
examination and copying at the NRC Public Document Room located at
One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland 20852. Copies are also available for examination
at the NRC Library located at Two White Flint North, 11545 Rockville
Pike, Rockville, Maryland 20582 and the Office of the Federal
Register, 800 North Capitol Street, NW., Suite 700, Washington, DC.
B. An applicant or licensee referencing this appendix, in
accordance with Section IV of this appendix, shall incorporate by
reference and comply with the requirements of this appendix,
including Tier 1, Tier 2, and the generic technical specifications
except as otherwise provided in this appendix. Conceptual design
information, as set forth in the generic DCD, and the Technical
Support Document for the System 80+ design are not part of this
appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD,
then Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the System 80+ design or
NUREG-1462, ``Final Safety Evaluation Report Related to the
Certification of the System 80+ Design,'' (FSER) and Supplement No.
1, then the generic DCD controls.
E. Design activities for structures, systems, and components
that are wholly outside the scope of this appendix may be performed
using site characteristics, provided the design activities do not
affect the DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference
this appendix shall, in addition to complying with the requirements
of 10 CFR 52.77, 52.79, and 52.80, comply with the following
requirements:
1. Incorporate by reference, as part of its application, this
appendix;
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information
and using the same organization and numbering as the generic DCD for
the System 80+ design, as modified and supplemented by the
applicant's exemptions and departures;
b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
c. Plant-specific technical specifications, consisting of the
generic and site-specific technical specifications, that are
required by 10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters
and interface requirements;
e. Information that addresses the COL action items; and
f. Information required by 10 CFR 52.47 that is not within the
scope of this appendix.
3. Include, in the plant-specific DCD, the proprietary
information referenced in the System 80+ DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the System 80+ design are in 10 CFR parts
20, 50, 73, and 100, codified as of May 9, 1997, that are applicable
and technically relevant, as described in the FSER (NUREG-1462) and
Supplement No. 1.
B. The System 80+ design is exempt from portions of the
following regulations:
1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Separate Plant Safety
Parameter Display Console;
2. Paragraphs (f)(2) (vii), (viii), (xxvi), and (xxviii) of 10
CFR 50.34--Accident Source Terms;
3. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-Accident
Sampling for Hydrogen, Boron, Chloride, and Dissolved Gases;
4. Paragraph (f)(3)(iv) of 10 CFR 50.34--Dedicated Containment
Penetration; and
5. Paragraphs III.A.1(a) and III.C.3(b) of Appendix J to 10 CFR
50--Containment Leakage Testing.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the System 80+ design comply with
the provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of
the public. A conclusion that a matter is resolved includes the
finding that additional or alternative structures, systems,
components, design features, design criteria, testing, analyses,
acceptance criteria, or justifications are not necessary for the
System 80+ design.
B. The Commission considers the following matters resolved
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings
for issuance of a combined license, amendment of a combined license,
or renewal of a combined license, proceedings held under 10 CFR
52.103, and enforcement proceedings involving plants referencing
this appendix:
1. All nuclear safety issues, except for the generic technical
specifications and other operational requirements, associated with
the information in the FSER and Supplement No. 1, Tier 1, Tier 2
(including referenced information which the context indicates is
intended as requirements), and the rulemaking record for
certification of the System 80+ design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and
in context, are intended as requirements in the generic DCD for the
System 80+ design;
3. All generic changes to the DCD under and in compliance with
the change processes in Sections VIII.A.1 and VIII.B.1 of this
appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix,
but only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix,
all departures from Tier 2 under and in compliance with the change
processes in paragraph VIII.B.5 of this appendix that do not require
prior NRC approval, but only for that plant;
7. All environmental issues concerning severe accident
mitigation design alternatives associated with the information in
the NRC's final environmental assessment for the System 80+ design
and the technical support document for the System 80+ design, dated
January 1995, for plants referencing this appendix whose site
parameters are within those specified in the technical support
document.
C. The Commission does not consider operational requirements for
an applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an
applicant or licensee who references this appendix by rule,
regulation, order, or license condition.
D. Except in accordance with the change processes in Section
VIII of this appendix, the Commission may not require an applicant
or licensee who references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures,
systems, components, or design features discussed in the generic
DCD.
E.1. Persons who wish to review proprietary information or other
secondary references in the DCD for the System 80+ design, in order
to request or participate in
[[Page 49550]]
the hearing required by 10 CFR 52.85 or the hearing provided under
10 CFR 52.103, or to request or participate in any other hearing
relating to this appendix in which interested persons have
adjudicatory hearing rights, shall first request access to such
information from Westinghouse. The request must state with
particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the
public at the NRC Web site, http://www.nrc.gov, and/or at the NRC
Public Document Room, is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to
prepare a request for hearing, the request must be filed no later
than 15 days after publication in the Federal Register of the notice
required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse
declines to provide the information sought, Westinghouse shall send
a written response within ten (10) days of receiving the request to
the requesting person setting forth with particularity the reasons
for its refusal. The person may then request the Commission (or
presiding officer, if a proceeding has been established) to order
disclosure. The person shall include copies of the original request
(and any subsequent clarifying information provided by the
requesting party to the applicant) and the applicant's response. The
Commission and presiding officer shall base their decisions solely
on the person's original request (including any clarifying
information provided by the requesting person to Westinghouse), and
Westinghouse's response. The Commission and presiding officer may
order Westinghouse to provide access to some or all of the requested
information, subject to an appropriate non-disclosure agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from
June 20, 1997, except as provided for in 10 CFR 52.55(b) and
52.57(b). This appendix remains valid for an applicant or licensee
who references this appendix until the application is withdrawn or
the license expires, including any period of extended operation
under a renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design.
B. Tier 2 Information
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under Sec. Sec. 52.55 or 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to assure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in 10 CFR 52.7 are present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 information. The Commission may
grant such a request only if it determines that the exemption will
comply with the requirements of 10 CFR 52.7. The Commission will
deny a request for an exemption from Tier 2, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design. The grant of an exemption
to an applicant must be subject to litigation in the same manner as
other issues material to the license hearing. The grant of an
exemption to a licensee must be subject to an opportunity for a
hearing in the same manner as license amendments.
5.a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information, Tier 2* information, or the technical specifications,
or requires a license amendment under paragraphs B.5.b or B.5.c of
this section. When evaluating the proposed departure, an applicant
or licensee shall consider all matters described in the plant-
specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would--
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety previously evaluated in the plant-specific
DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of an SSC important to safety previously evaluated
in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important
to safety with a different result than any evaluated previously in
the plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe
accident previously reviewed and determined to be not credible could
become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously
reviewed.
d. If a departure requires a license amendment under paragraph
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
e. A departure from Tier 2 information that is made under
paragraph B.5 of this section does not require an exemption from
this appendix.
f. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.103(a), who believes that an applicant or licensee who
references this appendix has not complied with paragraph VIII.B.5 of
this appendix when departing from Tier 2 information, may petition
the NRC to admit into the proceeding such a contention. In addition
to compliance with the general requirements of 10 CFR 2.309, the
petition must demonstrate that the departure does not comply with
paragraph VIII.B.5 of this appendix. Further, the petition must
demonstrate that the change bears on an asserted noncompliance with
an ITAAC acceptance criterion in the case of a 10 CFR 52.103
preoperational hearing, or that the change bears directly on the
amendment request in the case of a hearing on a license amendment.
Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall
certify the matter directly to the Commission for determination of
the admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6.a. An applicant who references this appendix may not depart
from Tier 2*
[[Page 49551]]
information, which is designated with italicized text or brackets
and an asterisk in the generic DCD, without NRC approval. The
departure will not be considered a resolved issue, within the
meaning of Section VI of this appendix and 10 CFR 52.63(a)(5).
b. A licensee who references this appendix may not depart from
the following Tier 2* matters without prior NRC approval. A request
for a departure will be treated as a request for a license amendment
under 10 CFR 50.90.
(1) Maximum fuel rod average burnup.
(2) Control room human factors engineering.
c. A licensee who references this appendix may not, before the
plant first achieves full power following the finding required by 10
CFR 52.103(g), depart from the following Tier 2* matters except in
accordance with paragraph B.6.b of this section. After the plant
first achieves full power, the following Tier 2* matters revert to
Tier 2 status and are thereafter subject to the departure provisions
in paragraph B.5 of this section.
(1) ASME Boiler & Pressure Vessel Code, Section III.
(2) ACI 349 and ANSI/AISC-690.
(3) Motor-operated valves.
(4) Equipment seismic qualification methods.
(5) Piping design acceptance criteria.
(6) Fuel and control rod design, except burnup limit.
(7) Instrumentation and controls setpoint methodology.
(8) Instrumentation and controls hardware and software changes.
(9) Instrumentation and controls environmental qualification.
(10) Seismic design criteria for non-seismic Category I
structures.
d. Departures from Tier 2* information that are made under
paragraph B.6 of this section do not require an exemption from this
appendix.
C. Operational requirements.
1. Generic changes to generic technical specifications and other
operational requirements that were completely reviewed and approved
in the design certification rulemaking and do not require a change
to a design feature in the generic DCD are governed by the
requirements in 10 CFR 50.109. Generic changes that do require a
change to a design feature in the generic DCD are governed by the
requirements in paragraphs A or B of this section.
2. Generic changes to generic TS and other operational
requirements are applicable to all applicants who reference this
appendix, except those for which the change has been rendered
technically irrelevant by action taken under paragraphs C.3 or C.4
of this section.
3. The Commission may require plant-specific departures on
generic technical specifications and other operational requirements
that were completely reviewed and approved, provided a change to a
design feature in the generic DCD is not required and special
circumstances as defined in 10 CFR 2.335 are present. The Commission
may modify or supplement generic technical specifications and other
operational requirements that were not completely reviewed and
approved or require additional technical specifications and other
operational requirements on a plant-specific basis, provided a
change to a design feature in the generic DCD is not required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other
operational requirements. The Commission may grant such a request
only if it determines that the exemption will comply with the
requirements of 10 CFR 52.7. The grant of an exemption must be
subject to litigation in the same manner as other issues material to
the license hearing.
5. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.103(a), who believes that an operational requirement
approved in the DCD or a technical specification derived from the
generic technical specifications must be changed may petition to
admit into the proceeding such a contention. Such a petition must
comply with the general requirements of 10 CFR 2.309 and must
demonstrate why special circumstances as defined in 10 CFR 2.335 are
present, or for compliance with the Commission's regulations in
effect at the time this appendix was approved, as set forth in
Section V of this appendix. Any other party may file a response
thereto. If, on the basis of the petition and any response, the
presiding officer determines that a sufficient showing has been
made, the presiding officer shall certify the matter directly to the
Commission for determination of the admissibility of the contention.
All other issues with respect to the plant-specific technical
specifications or other operational requirements are subject to a
hearing as part of the license proceeding.
6. After issuance of a license, the generic technical
specifications have no further effect on the plant-specific
technical specifications and changes to the plant-specific technical
specifications will be treated as license amendments under 10 CFR
50.90.
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1 An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities, and a licensee may proceed at its own risk with design,
procurement, construction, and preoperational activities, even
though the NRC may not have found that any particular ITAAC has been
met.
2. The licensee who references this appendix shall notify the
NRC that the required inspections, tests, and analyses in the ITAAC
have been successfully completed and that the corresponding
acceptance criteria have been met.
3. In the event that an activity is subject to an ITAAC, and the
applicant or licensee who references this appendix has not
demonstrated that the ITAAC has been met, the applicant or licensee
may either take corrective actions to successfully complete that
ITAAC, request an exemption from the ITAAC in accordance with
Section VIII of this appendix and 10 CFR 52.97(b), or petition for
rulemaking to amend this appendix by changing the requirements of
the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes
to the ITAAC must meet the requirements of Section VIII.A.1 of this
appendix.
B.1 The NRC shall ensure that the required inspections, tests,
and analyses in the ITAAC are performed. The NRC shall verify that
the inspections, tests, and analyses referenced by the licensee have
been successfully completed and, based solely thereon, find the
prescribed acceptance criteria have been met. At appropriate
intervals during construction, the NRC shall publish notices of the
successful completion of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall
find that the acceptance criteria in the ITAAC for the license are
met before fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the
subject of a Sec. 52.103(a) hearing, their expiration will occur
upon final Commission action in such proceeding. However, subsequent
modifications must comply with the Tier 1 and Tier 2 design
descriptions in the plant-specific DCD unless the licensee has
complied with the applicable requirements of 10 CFR 52.98 and
Section VIII of this appendix.
X. Records and Reporting
A. Records.
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and
the generic TS and other operational requirements. The applicant
shall maintain the proprietary and safeguards information referenced
in the generic DCD for the period that this appendix may be
referenced, as specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application
and for the term of the license (including any period of renewal).
3. An applicant or licensee who references this appendix shall
prepare and maintain written evaluations which provide the bases for
the determinations required by Section VIII of this appendix. These
evaluations must be retained throughout the period of application
and for the term of the license (including any period of renewal).
B. Reporting.
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any
plant-specific departures from the DCD, including a summary of the
evaluation of each. This report must be filed in accordance with the
filing requirements applicable to reports in 10 CFR 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its
[[Page 49552]]
DCD, which reflect the generic changes to and plant-specific
departures from the generic DCD made under Section VIII of this
appendix. These updates must be filed under the filing requirements
applicable to final safety analysis report updates in 10 CFR 52.3
and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and
X.B.2 must be submitted as follows:
a. On the date that an application for a license referencing
this appendix is submitted, the application must include the report
and any updates to the generic DCD.
b. During the interval from the date of application for a
license to the date the Commission makes the finding required by 10
CFR 52.103(g), the report must be submitted semi-annually. Updates
to the plant-specific DCD must be submitted annually and may be
submitted along with amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), the reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the
final safety analysis report for the facility, at the intervals
required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at
shorter intervals as specified in the license.
Appendix C to Part 52--Design Certification Rule for the AP600 Design
I. Introduction
Appendix C constitutes the standard design certification for the
AP600 \1\ design, in accordance with 10 CFR part 52, subpart B. The
applicant for certification of the AP600 design is Westinghouse
Electric Company LLC.
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\1\ AP600 is a trademark of Westinghouse Electric Company LLC.
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II. Definitions
A. Generic design control document (generic DCD) means the
document containing the Tier 1 and Tier 2 information and generic
technical specifications that is incorporated by reference into this
appendix.
B. Generic technical specifications means the information,
required by 10 CFR 50.36 and 50.36a, for the portion of the plant
that is within the scope of this appendix.
C. Plant-specific DCD means the document, maintained by an
applicant or licensee who references this appendix, consisting of
the information in the generic DCD, as modified and supplemented by
the plant-specific departures and exemptions made under Section VIII
of this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (hereinafter Tier 1 information). The design descriptions,
interface requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by
this appendix (Tier 2 information). Compliance with Tier 2 is
required, but generic changes to and plant-specific departures from
Tier 2 are governed by Section VIII of this appendix. Compliance
with Tier 2 provides a sufficient, but not the only acceptable,
method for complying with Tier 1. Compliance methods differing from
Tier 2 must satisfy the change process in Section VIII of this
appendix. Regardless of these differences, an applicant or licensee
must meet the requirement in Section III.B of this appendix to
reference Tier 2 when referencing Tier 1. Tier 2 information
includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c),
with the exception of generic technical specifications and
conceptual design information;
2. Supporting information on the inspections, tests, and
analyses that will be performed to demonstrate that the acceptance
criteria in the ITAAC have been met; and
3. Combined license (COL) action items (COL license
information), which identify certain matters that must be addressed
in the site-specific portion of the final safety analysis report
(FSAR) by an applicant who references this appendix. These items
constitute information requirements but are not the only acceptable
set of information in the FSAR. An applicant may depart from or omit
these items, provided that the departure or omission is identified
and justified in the FSAR. After issuance of a construction permit
or COL, these items are not requirements for the licensee unless
such items are restated in the FSAR.
4. The investment protection short-term availability controls in
Section 16.3 of the DCD.
F. Tier 2* means the portion of the Tier 2 information,
designated as such in the generic DCD, which is subject to the
change process in Section VIII.B.6 of this appendix. This
designation expires for some Tier 2* information under Section
VIII.B.6.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
(1) Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are
conservative or essentially the same; or
(2) Changing from a method described in the plant-specific DCD
to another method unless that method has been approved by NRC for
the intended application.
H. All other terms in this appendix have the meaning set out in
10 CFR 50.2 or 52.1, or Section 11 of the Atomic Energy Act of 1954,
as amended, as applicable.
III. Scope and Contents
A. Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3), and the generic
technical specifications in the AP600 DCD (12/99 revision) are
approved for incorporation by reference by the Director of the
Office of the Federal Register on January 24, 2000, in accordance
with 5 U.S.C. 552(a) and 1 CFR part 51. Copies of the generic DCD
may be obtained from Ronald P. Vijuk, Manager, Passive Plant
Engineering, Westinghouse Electric Company, P.O. Box 355,
Pittsburgh, Pennsylvania 15230-0355. A copy of the generic DCD is
available for examination and copying at the NRC Public Document
Room located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. Copies are also available for
examination at the NRC Library located at Two White Flint North,
11545 Rockville Pike, Rockville, Maryland 20582; and the Office of
the Federal Register, 800 North Capitol Street, NW., Suite 700,
Washington, DC.
B. An applicant or licensee referencing this appendix, in
accordance with Section IV of this appendix, shall incorporate by
reference and comply with the requirements of this appendix,
including Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3), and the generic
technical specifications except as otherwise provided in this
appendix. Conceptual design information in the generic DCD and the
evaluation of severe accident mitigation design alternatives in
Appendix 1B of the generic DCD are not part of this appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD,
then Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the AP600 design or NUREG-
1512, ``Final Safety Evaluation Report Related to Certification of
the AP600 Standard Design,'' (FSER), then the generic DCD controls.
E. Design activities for structures, systems, and components
that are wholly outside the scope of this appendix may be performed
using site characteristics, provided the design activities do not
affect the DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference
this appendix shall, in addition to complying with the requirements
of 10 CFR 52.77, 52.79, and 52.80, comply with the following
requirements:
1. Incorporate by reference, as part of its application, this
appendix;
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information
and utilizing the same organization and numbering as the generic DCD
for the AP600 design, as modified and supplemented by the
applicant's exemptions and departures;
b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
c. Plant-specific technical specifications, consisting of the
generic and site-specific technical specifications, that are
required by 10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters
and interface requirements;
e. Information that addresses the COL action items; and
[[Page 49553]]
f. Information required by 10 CFR 52.47 that is not within the
scope of this appendix.
3. Include, in the plant-specific DCD, the proprietary
information and safeguards information referenced in the AP600 DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the AP600 design are in 10 CFR parts 20,
50, 73, and 100, codified as of December 16, 1999, that are
applicable and technically relevant, as described in the FSER
(NUREG-1512) and the supplementary information for this section.
B. The AP600 design is exempt from portions of the following
regulations:
1. Paragraph (a)(1) of 10 CFR 50.34--whole body dose criterion;
2. Paragraph (f)(2)(iv) of 10 CFR 50.34--Plant Safety Parameter
Display Console;
3. Paragraphs (f)(2)(vii), (viii), (xxvi), and (xxviii) of 10
CFR 50.34--Accident Source Term in TID 14844;
4. Paragraph (a)(2) of 10 CFR 50.55a--ASME Boiler and Pressure
Vessel Code;
5. Paragraph (c)(1) of 10 CFR 50.62--Auxiliary (or emergency)
feedwater system;
6. Appendix A to 10 CFR part 50, GDC 17--Offsite Power Sources;
and
7. Appendix A to 10 CFR part 50, GDC 19--whole body dose
criterion.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the AP600 design comply with the
provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of
the public. A conclusion that a matter is resolved includes the
finding that additional or alternative structures, systems,
components, design features, design criteria, testing, analyses,
acceptance criteria, or justifications are not necessary for the
AP600 design.
B. The Commission considers the following matters resolved
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings
for issuance of a combined license, amendment of a combined license,
or renewal of a combined license, proceedings held under 10 CFR
52.103, and enforcement proceedings involving plants referencing
this appendix:
1. All nuclear safety issues, except for the generic technical
specifications and other operational requirements, associated with
the information in the FSER and Supplement No. 1, Tier 1, Tier 2
(including referenced information which the context indicates is
intended as requirements and the investment protection short-term
availability controls in Section 16.3), and the rulemaking record
for certification of the AP600 design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and
in context, are intended as requirements in the generic DCD for the
AP600 design;
3. All generic changes to the DCD under and in compliance with
the change processes in Sections VIII.A.1 and VIII.B.1 of this
appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix,
but only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix,
all departures from Tier 2 under and in compliance with the change
processes in paragraph VIII.B.5 of this appendix that do not require
prior NRC approval, but only for that plant;
7. All environmental issues concerning severe accident
mitigation design alternatives associated with the information in
the NRC's environmental assessment for the AP600 design and appendix
1B of the generic DCD, for plants referencing this appendix whose
site parameters are within those specified in the severe accident
mitigation design alternatives evaluation.
C. The Commission does not consider operational requirements for
an applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an
applicant or licensee who references this appendix by rule,
regulation, order, or license condition.
D. Except in accordance with the change processes in Section
VIII of this appendix, the Commission may not require an applicant
or licensee who references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures,
systems, components, or design features discussed in the generic
DCD.
E.1. Persons who wish to review proprietary and safeguards
information or other secondary references in the AP600 DCD, in order
to request or participate in the hearing required by 10 CFR 52.85 or
the hearing provided under 10 CFR 52.103, or to request or
participate in any other hearing relating to this appendix in which
interested persons have adjudicatory hearing rights, shall first
request access to such information from Westinghouse. The request
must state with particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the
public at the NRC Web site, http://www.nrc.gov, and/or at the NRC
Public Document Room, is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to
prepare a request for hearing, the request must be filed no later
than 15 days after publication in the Federal Register of the notice
required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse
declines to provide the information sought, Westinghouse shall send
a written response within 10 days of receiving the request to the
requesting person setting forth with particularity the reasons for
its refusal. The person may then request the Commission (or
presiding officer, if a proceeding has been established) to order
disclosure. The person shall include copies of the original request
(and any subsequent clarifying information provided by the
requesting party to the applicant) and the applicant's response. The
Commission and presiding officer shall base their decisions solely
on the person's original request (including any clarifying
information provided by the requesting person to Westinghouse), and
Westinghouse's response. The Commission and presiding officer may
order Westinghouse to provide access to some or all of the requested
information, subject to an appropriate non-disclosure agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from
January 24, 2000, except as provided for in 10 CFR 52.55(b) and
52.57(b). This appendix remains valid for an applicant or licensee
who references this appendix until the application is withdrawn or
the license expires, including any period of extended operation
under a renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design.
B. Tier 2 information.
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under Sec. Sec. 52.55 or 52.61, unless:
[[Page 49554]]
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to assure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in 10 CFR 52.7 are present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 information. The Commission may
grant such a request only if it determines that the exemption will
comply with the requirements of 10 CFR 52.7. The Commission will
deny a request for an exemption from Tier 2, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design. The grant of an exemption
to an applicant must be subject to litigation in the same manner as
other issues material to the license hearing. The grant of an
exemption to a licensee must be subject to an opportunity for a
hearing in the same manner as license amendments.
5.a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information, Tier 2* information, or the technical specifications,
or requires a license amendment under paragraphs B.5.b or B.5.c of
this section. When evaluating the proposed departure, an applicant
or licensee shall consider all matters described in the plant-
specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would:
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety previously evaluated in the plant-specific
DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of an SSC important to safety previously evaluated
in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important
to safety with a different result than any evaluated previously in
the plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe
accident previously reviewed and determined to be not credible could
become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously
reviewed.
d. If a departure requires a license amendment under paragraphs
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
e. A departure from Tier 2 information that is made under
paragraph B.5 of this section does not require an exemption from
this appendix.
f. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.103(a), who believes that an applicant or licensee who
references this appendix has not complied with paragraph VIII.B.5 of
this appendix when departing from Tier 2 information, may petition
the NRC to admit into the proceeding such a contention. In addition
to compliance with the general requirements of 10 CFR 2.309, the
petition must demonstrate that the departure does not comply with
paragraph VIII.B.5 of this appendix. Further, the petition must
demonstrate that the change bears on an asserted noncompliance with
an ITAAC acceptance criterion in the case of a 10 CFR 52.103
preoperational hearing, or that the change bears directly on the
amendment request in the case of a hearing on a license amendment.
Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall
certify the matter directly to the Commission for determination of
the admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6a. An applicant who references this appendix may not depart
from Tier 2* information, which is designated with italicized text
or brackets and an asterisk in the generic DCD, without NRC
approval. The departure will not be considered a resolved issue,
within the meaning of Section VI of this appendix and 10 CFR
52.63(a)(5).
b. A licensee who references this appendix may not depart from
the following Tier 2* matters without prior NRC approval. A request
for a departure will be treated as a request for a license amendment
under 10 CFR 50.90.
(1) Maximum fuel rod average burn-up.
(2) Fuel principal design requirements.
(3) Fuel criteria evaluation process.
(4) Fire areas.
(5) Human factors engineering.
c. A licensee who references this appendix may not, before the
plant first achieves full power following the finding required by 10
CFR 52.103(g), depart from the following Tier 2* matters except in
accordance with paragraph B.6.b of this section. After the plant
first achieves full power, the following Tier 2* matters revert to
Tier 2 status and are thereafter subject to the departure provisions
in paragraph B.5 of this section.
(1) Nuclear Island structural dimensions.
(2) ASME Boiler and Pressure Vessel Code, Section III, and Code
Case--284.
(3) Design Summary of Critical Sections.
(4) ACI 318, ACI 349, and ANSI/AISC--690.
(5) Definition of critical locations and thicknesses.
(6) Seismic qualification methods and standards.
(7) Nuclear design of fuel and reactivity control system, except
burn-up limit.
(8) Motor-operated and power-operated valves.
(9) Instrumentation and control system design processes,
methods, and standards.
(10) PRHR natural circulation test (first plant only).
(11) ADS and CMT verification tests (first three plants only).
d. Departures from Tier 2* information that are made under
paragraph B.6 of this section do not require an exemption from this
appendix.
C. Operational requirements.
1. Generic changes to generic technical specifications and other
operational requirements that were completely reviewed and approved
in the design certification rulemaking and do not require a change
to a design feature in the generic DCD are governed by the
requirements in 10 CFR 50.109. Generic changes that do require a
change to a design feature in the generic DCD are governed by the
requirements in paragraphs A or B of this section.
2. Generic changes to generic TS and other operational
requirements are applicable to all applicants who reference this
appendix, except those for which the change has been rendered
technically irrelevant by action taken under paragraphs C.3 or C.4
of this section.
3. The Commission may require plant-specific departures on
generic technical specifications and other operational requirements
that were completely reviewed and approved, provided a change to a
design feature in the generic DCD is not required and special
circumstances as defined in 10 CFR 2.335 are present. The Commission
may modify or supplement generic technical specifications and other
operational requirements that were not completely reviewed and
approved or require additional technical specifications and other
operational requirements on a plant-specific basis, provided a
change to a design feature in the generic DCD is not required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other
operational requirements. The Commission may grant such a request
only if it determines that the exemption will comply with the
requirements of 10 CFR 52.7. The grant of an exemption must be
subject to litigation in the same manner as other issues material to
the license hearing.
[[Page 49555]]
5. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR 52.103(a), who believes that an operational requirement
approved in the DCD or a technical specification derived from the
generic technical specifications must be changed may petition to
admit into the proceeding such a contention. Such petition must
comply with the general requirements of 10 CFR 2.309 and must
demonstrate why special circumstances as defined in 10 CFR 2.335 are
present, or for compliance with the Commission's regulations in
effect at the time this appendix was approved, as set forth in
Section V of this appendix. Any other party may file a response
thereto. If, on the basis of the petition and any response, the
presiding officer determines that a sufficient showing has been
made, the presiding officer shall certify the matter directly to the
Commission for determination of the admissibility of the contention.
All other issues with respect to the plant-specific technical
specifications or other operational requirements are subject to a
hearing as part of the license proceeding.
6. After issuance of a license, the generic technical
specifications have no further effect on the plant-specific
technical specifications and changes to the plant-specific technical
specifications will be treated as license amendments under 10 CFR
50.90.
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1 An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities, and a licensee may proceed at its own risk with design,
procurement, construction, and preoperational activities, even
though the NRC may not have found that any particular ITAAC has been
met.
2. The licensee who references this appendix shall notify the
NRC that the required inspections, tests, and analyses in the ITAAC
have been successfully completed and that the corresponding
acceptance criteria have been met.
3. In the event that an activity is subject to an ITAAC, and the
applicant or licensee who references this appendix has not
demonstrated that the ITAAC has been met, the applicant or licensee
may either take corrective actions to successfully complete that
ITAAC, request an exemption from the ITAAC in accordance with
Section VIII of this appendix and 10 CFR 52.97(b), or petition for
rulemaking to amend this appendix by changing the requirements of
the ITAAC, under 10 CFR 2.802 and 52.97(b). Such rulemaking changes
to the ITAAC must meet the requirements of paragraph VIII.A.1 of
this appendix.
B.1. The NRC shall ensure that the required inspections, tests,
and analyses in the ITAAC are performed. The NRC shall verify that
the inspections, tests, and analyses referenced by the licensee have
been successfully completed and, based solely thereon, find the
prescribed acceptance criteria have been met. At appropriate
intervals during construction, the NRC shall publish notices of the
successful completion of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall
find that the acceptance criteria in the ITAAC for the license are
met before fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the
subject of a Sec. 52.103(a) hearing, their expiration will occur
upon final Commission action in such proceeding. However, subsequent
modifications must comply with the Tier 1 and Tier 2 design
descriptions in the plant-specific DCD unless the licensee has
complied with the applicable requirements of 10 CFR 52.98 and
Section VIII of this appendix.
X. Records and Reporting
A. Records.
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and
the generic TS and other operational requirements. The applicant
shall maintain the proprietary and safeguards information referenced
in the generic DCD for the period that this appendix may be
referenced, as specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application
and for the term of the license (including any period of renewal).
3. An applicant or licensee who references this appendix shall
prepare and maintain written evaluations which provide the bases for
the determinations required by Section VIII of this appendix. These
evaluations must be retained throughout the period of application
and for the term of the license (including any period of renewal).
B. Reporting.
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any
plant-specific departures from the DCD, including a summary of the
evaluation of each. This report must be filed in accordance with the
filing requirements applicable to reports in 10 CFR 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its DCD, which reflect the generic changes to and
plant-specific departures from the generic DCD made under Section
VIII of this appendix. These updates must be filed under the filing
requirements applicable to final safety analysis report updates in
10 CFR 52.3 and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and
X.B.2 must be submitted as follows:
a. On the date that an application for a license referencing
this appendix is submitted, the application must include the report
and any updates to the generic DCD.
b. During the interval from the date of application for a
license to the date the Commission makes the finding required by 10
CFR 52.103(g), the report must be submitted semi-annually. Updates
to the plant-specific DCD must be submitted annually and may be
submitted along with amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), the reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the
final safety analysis report for the facility, at the intervals
required by 10 CFR 50.59(d)(2) and 50.71(e), respectively, or at
shorter intervals as specified in the license.
Appendix D to Part 52--Design Certification Rule for the AP1000 Design
I. Introduction
Appendix D constitutes the standard design certification for the
AP1000 \1\ design, in accordance with 10 CFR part 52, subpart B. The
applicant for certification of the AP1000 design is Westinghouse
Electric Company LLC.
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\1\ AP1000 is a trademark of Westinghouse Electric Company LLC.
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II. Definitions
A. Generic design control document (generic DCD) means the
document containing the Tier 1 and Tier 2 information and generic
technical specifications that is incorporated by reference into this
appendix.
B. Generic technical specifications means the information
required by 10 CFR 50.36 and 50.36a for the portion of the plant
that is within the scope of this appendix.
C. Plant-specific DCD means the document maintained by an
applicant or licensee who references this appendix consisting of the
information in the generic DCD as modified and supplemented by the
plant-specific departures and exemptions made under Section VIII of
this appendix.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by
this appendix (Tier 2 information). Compliance with Tier 2 is
required, but generic changes to and plant-specific departures from
Tier 2 are governed by Section VIII of this appendix. Compliance
with Tier 2 provides a sufficient, but not the only acceptable,
method for complying with Tier 1. Compliance methods differing from
Tier 2 must satisfy the change process in Section VIII of this
appendix. Regardless of these differences, an applicant or licensee
must
[[Page 49556]]
meet the requirement in Section III.B of this appendix to reference
Tier 2 when referencing Tier 1. Tier 2 information includes:
1. Information required by Sec. Sec. 52.47(a) and 52.47(c),
with the exception of generic technical specifications and
conceptual design information;
2. Supporting information on the inspections, tests, and
analyses that will be performed to demonstrate that the acceptance
criteria in the ITAAC have been met; and
3. Combined license (COL) action items (COL license
information), which identify certain matters that must be addressed
in the site-specific portion of the final safety analysis report
(FSAR) by an applicant who references this appendix. These items
constitute information requirements but are not the only acceptable
set of information in the FSAR. An applicant may depart from or omit
these items, provided that the departure or omission is identified
and justified in the FSAR. After issuance of a construction permit
or COL, these items are not requirements for the licensee unless
such items are restated in the FSAR.
4. The investment protection short-term availability controls in
Section 16.3 of the DCD.
F. Tier 2* means the portion of the Tier 2 information,
designated as such in the generic DCD, which is subject to the
change process in Section VIII.B.6 of this appendix. This
designation expires for some Tier 2* information under paragraph
VIII.B.6.
G. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
1. Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are
conservative or essentially the same; or
2. Changing from a method described in the plant-specific DCD to
another method unless that method has been approved by the NRC for
the intended application.
H. All other terms in this appendix have the meaning set out in
10 CFR 50.2, or 52.1, or Section 11 of the Atomic Energy Act of
1954, as amended, as applicable.
III. Scope and Contents
A. Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3), and the generic TS in
the AP1000 DCD (Revision 15, dated December 8, 2005) are approved
for incorporation by reference by the Director of the Office of the
Federal Register on February 27, 2006, under 5 U.S.C. 552(a) and 1
CFR part 51. Copies of the generic DCD may be obtained from Ronald
P. Vijuk, Manager, Passive Plant Engineering, Westinghouse Electric
Company, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355. A copy
of the generic DCD is also available for examination and copying at
the NRC Public Document Room, One White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852. Copies are available for
examination at the NRC Library, Two White Flint North, 11545
Rockville Pike, Rockville, Maryland, telephone (301) 415-5610, e-
mail [email protected] or at the National Archives and Records
Administration (NARA). For information on the availability of this
material at NARA, call (202) 741-6030 or go to http://www.archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html.
B. An applicant or licensee referencing this appendix, in
accordance with Section IV of this appendix, shall incorporate by
reference and comply with the requirements of this appendix,
including Tier 1, Tier 2 (including the investment protection short-
term availability controls in Section 16.3 of the DCD), and the
generic TS except as otherwise provided in this appendix. Conceptual
design information in the generic DCD and the evaluation of severe
accident mitigation design alternatives in appendix 1B of the
generic DCD are not part of this appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD,
then Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for design certification of the AP1000 design or NUREG-
1793, ``Final Safety Evaluation Report Related to Certification of
the AP1000 Standard Design,'' (FSER) and Supplement No. 1, then the
generic DCD controls.
E. Design activities for structures, systems, and components
that are wholly outside the scope of this appendix may be performed
using site characteristics, provided the design activities do not
affect the DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a combined license that wishes to reference
this appendix shall, in addition to complying with the requirements
of 10 CFR 52.77, 52.79, and 52.80, comply with the following
requirements:
1. Incorporate by reference, as part of its application, this
appendix.
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information
and using the same organization and numbering as the generic DCD for
the AP1000 design, as modified and supplemented by the applicant's
exemptions and departures;
b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
c. Plant-specific TS, consisting of the generic and site-
specific TS that are required by 10 CFR 50.36 and 50.36a;
d. Information demonstrating compliance with the site parameters
and interface requirements;
e. Information that addresses the COL action items; and
f. Information required by 10 CFR 52.47(a) that is not within
the scope of this appendix.
3. Include, in the plant-specific DCD, the proprietary
information and safeguards information referenced in the AP1000 DCD.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to the AP1000 design are in 10 CFR parts 20,
50, 73, and 100, codified as of January 23, 2006, that are
applicable and technically relevant, as described in the FSER
(NUREG-1793) and Supplement No. 1.
B. The AP1000 design is exempt from portions of the following
regulations:
1. Paragraph (f)(2)(iv) of 10 CFR 50.34--Plant Safety Parameter
Display Console;
2. Paragraph (c)(1) of 10 CFR 50.62--Auxiliary (or emergency)
feedwater system; and
3. Appendix A to 10 CFR part 50, GDC 17--Second offsite power
supply circuit.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
components, and design features of the AP1000 design comply with the
provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of
the public. A conclusion that a matter is resolved includes the
finding that additional or alternative structures, systems,
components, design features, design criteria, testing, analyses,
acceptance criteria, or justifications are not necessary for the
AP1000 design.
B. The Commission considers the following matters resolved
within the meaning of 10 CFR 52.63(a)(5) in subsequent proceedings
for issuance of a COL, amendment of a COL, or renewal of a COL,
proceedings held under 10 CFR 52.103, and enforcement proceedings
involving plants referencing this appendix:
1. All nuclear safety issues, except for the generic TS and
other operational requirements, associated with the information in
the FSER and Supplement No. 1, Tier 1, Tier 2 (including referenced
information, which the context indicates is intended as
requirements, and the investment protection short-term availability
controls in Section 16.3 of the DCD), and the rulemaking record for
certification of the AP1000 design;
2. All nuclear safety and safeguards issues associated with the
information in proprietary and safeguards documents, referenced and
in context, are intended as requirements in the generic DCD for the
AP1000 design;
3. All generic changes to the DCD under and in compliance with
the change processes in Sections VIII.A.1 and VIII.B.1 of this
appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in Sections VIII.A.4 and VIII.B.4 of this appendix,
but only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.f of this appendix,
all departures from Tier 2 under and in compliance with the change
processes in paragraph VIII.B.5 of this appendix that do not require
prior NRC approval, but only for that plant;
7. All environmental issues concerning severe accident
mitigation design alternatives associated with the information in
the NRC's EA for the AP1000 design and Appendix 1B
[[Page 49557]]
of the generic DCD, for plants referencing this appendix whose site
parameters are within those specified in the severe accident
mitigation design alternatives evaluation.
C. The Commission does not consider operational requirements for
an applicant or licensee who references this appendix to be matters
resolved within the meaning of 10 CFR 52.63(a)(5). The Commission
reserves the right to require operational requirements for an
applicant or licensee who references this appendix by rule,
regulation, order, or license condition.
D. Except under the change processes in Section VIII of this
appendix, the Commission may not require an applicant or licensee
who references this appendix to:
1. Modify structures, systems, components, or design features as
described in the generic DCD;
2. Provide additional or alternative structures, systems,
components, or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures,
systems, components, or design features discussed in the generic
DCD.
E.1. Persons who wish to review proprietary and safeguards
information or other secondary references in the AP1000 DCD, in
order to request or participate in the hearing required by 10 CFR
52.85 or the hearing provided under 10 CFR 52.103, or to request or
participate in any other hearing relating to this appendix in which
interested persons have adjudicatory hearing rights, shall first
request access to such information from Westinghouse. The request
must state with particularity:
a. The nature of the proprietary or other information sought;
b. The reason why the information currently available to the
public in the NRC's public document room is insufficient;
c. The relevance of the requested information to the hearing
issue(s) which the person proposes to raise; and
d. A showing that the requesting person has the capability to
understand and utilize the requested information.
2. If a person claims that the information is necessary to
prepare a request for hearing, the request must be filed no later
than 15 days after publication in the Federal Register of the notice
required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse
declines to provide the information sought, Westinghouse shall send
a written response within 10 days of receiving the request to the
requesting person setting forth with particularity the reasons for
its refusal. The person may then request the Commission (or
presiding officer, if a proceeding has been established) to order
disclosure. The person shall include copies of the original request
(and any subsequent clarifying information provided by the
requesting party to the applicant) and the applicant's response. The
Commission and presiding officer shall base their decisions solely
on the person's original request (including any clarifying
information provided by the requesting person to Westinghouse), and
Westinghouse's response. The Commission and presiding officer may
order Westinghouse to provide access to some or all of the requested
information, subject to an appropriate non-disclosure agreement.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from
February 27, 2006, except as provided for in 10 CFR 52.55(b) and
52.57(b). This appendix remains valid for an applicant or licensee
who references this appendix until the application is withdrawn or
the license expires, including any period of extended operation
under a renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 information.
1. Generic changes to Tier 1 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in 10 CFR 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in 10 CFR 52.63(b)(1) and 52.98(f). The Commission will
deny a request for an exemption from Tier 1, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design.
B. Tier 2 information.
1. Generic changes to Tier 2 information are governed by the
requirements in 10 CFR 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs B.3, B.4, B.5, or B.6 of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order while this appendix is in effect
under 10 CFR 52.55 or 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to ensure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in 10 CFR 50.12(a) are
present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 information. The Commission may
grant such a request only if it determines that the exemption will
comply with the requirements of 10 CFR 50.12(a). The Commission will
deny a request for an exemption from Tier 2, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design. The grant of an exemption
to an applicant must be subject to litigation in the same manner as
other issues material to the license hearing. The grant of an
exemption to a licensee must be subject to an opportunity for a
hearing in the same manner as license amendments.
5.a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information, Tier 2* information, or the TS, or requires a license
amendment under paragraphs B.5.b or B.5.c of this section. When
evaluating the proposed departure, an applicant or licensee shall
consider all matters described in the plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD, requires a license amendment if it would:
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
(SSC) important to safety and previously evaluated in the plant-
specific DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of an SSC important to safety previously evaluated
in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of an SSC important
to safety with a different result than any evaluated previously in
the plant-specific DCD;
(7) Result in a design basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2 affecting resolution of an
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe
accident previously reviewed and determined to be not credible could
become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously
reviewed.
d. If a departure requires a license amendment under paragraph
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
e. A departure from Tier 2 information that is made under
paragraph B.5 of this section does not require an exemption from
this appendix.
f. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
10 CFR
[[Page 49558]]
52.103(a), who believes that an applicant or licensee who references
this appendix has not complied with paragraph VIII.B.5 of this
appendix when departing from Tier 2 information, may petition to
admit into the proceeding such a contention. In addition to
compliance with the general requirements of 10 CFR 2.309, the
petition must demonstrate that the departure does not comply with
paragraph VIII.B.5 of this appendix. Further, the petition must
demonstrate that the change bears on an asserted noncompliance with
an ITAAC acceptance criterion in the case of a 10 CFR 52.103
preoperational hearing, or that the change bears directly on the
amendment request in the case of a hearing on a license amendment.
Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall
certify the matter directly to the Commission for determination of
the admissibility of the contention. The Commission may admit such a
contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
6.a. An applicant who references this appendix may not depart
from Tier 2* information, which is designated with italicized text
or brackets and an asterisk in the generic DCD, without NRC
approval. The departure will not be considered a resolved issue,
within the meaning of Section VI of this appendix and 10 CFR
52.63(a)(5).
b. A licensee who references this appendix may not depart from
the following Tier 2* matters without prior NRC approval. A request
for a departure will be treated as a request for a license amendment
under 10 CFR 50.90.
(1) Maximum fuel rod average burn-up.
(2) Fuel principal design requirements.
(3) Fuel criteria evaluation process.
(4) Fire areas.
(5) Human factors engineering.
(6) Small-break loss-of-coolant accident (LOCA) analysis
methodology.
c. A licensee who references this appendix may not, before the
plant first achieves full power following the finding required by 10
CFR 52.103(g), depart from the following Tier 2* matters except
under paragraph B.6.b of this section. After the plant first
achieves full power, the following Tier 2* matters revert to Tier 2
status and are subject to the departure provisions in paragraph B.5
of this section.
(1) Nuclear Island structural dimensions.
(2) American Society of Mechanical Engineers Boiler & Pressure
Vessel Code (ASME Code), Section III, and Code Case-284.
(3) Design Summary of Critical Sections.
(4) American Concrete Institute (ACI) 318, ACI 349, American
National Standards Institute/American Institute of Steel
Construction (ANSI/AISC)-690, and American Iron and Steel Institute
(AISI), ``Specification for the Design of Cold Formed Steel
Structural Members, Part 1 and 2,'' 1996 Edition and 2000
Supplement.
(5) Definition of critical locations and thicknesses.
(6) Seismic qualification methods and standards.
(7) Nuclear design of fuel and reactivity control system, except
burn-up limit.
(8) Motor-operated and power-operated valves.
(9) Instrumentation and control system design processes,
methods, and standards.
(10) Passive residual heat removal (PRHR) natural circulation
test (first plant only).
(11) Automatic depressurization system (ADS) and core make-up
tank (CMT) verification tests (first three plants only).
(12) Polar crane parked orientation.
(13) Piping design acceptance criteria.
(14) Containment vessel design parameters.
d. Departures from Tier 2* information that are made under
paragraph B.6 of this section do not require an exemption from this
appendix.
C. Operational requirements.
1. Generic changes to generic TS and other operational
requirements that were completely reviewed and approved in the
design certification rulemaking and do not require a change to a
design feature in the generic DCD are governed by the requirements
in 10 CFR 50.109. Generic changes that require a change to a design
feature in the generic DCD are governed by the requirements in
paragraphs A or B of this section.
2. Generic changes to generic TS and other operational
requirements are applicable to all applicants who reference this
appendix, except those for which the change has been rendered
technically irrelevant by action taken under paragraphs C.3 or C.4
of this section.
3. The Commission may require plant-specific departures on
generic TS and other operational requirements that were completely
reviewed and approved, provided a change to a design feature in the
generic DCD is not required and special circumstances as defined in
10 CFR 2.335 are present. The Commission may modify or supplement
generic TS and other operational requirements that were not
completely reviewed and approved or require additional TS and other
operational requirements on a plant-specific basis, provided a
change to a design feature in the generic DCD is not required.
4. An applicant who references this appendix may request an
exemption from the generic technical specifications or other
operational requirements. The Commission may grant such a request
only if it determines that the exemption will comply with the
requirements of 10 CFR 52.7. The grant of an exemption must be
subject to litigation in the same manner as other issues material to
the license hearing.
5. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license, or for operation under
10 CFR 52.103(a), who believes that an operational requirement
approved in the DCD or a TS derived from the generic TS must be
changed may petition to admit such a contention into the proceeding.
The petition must comply with the general requirements of 10 CFR
2.309 and must demonstrate why special circumstances as defined in
10 CFR 2.335 are present, or demonstrate compliance with the
Commission's regulations in effect at the time this appendix was
approved, as set forth in Section V of this appendix. Any other
party may file a response to the petition. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall
certify the matter directly to the Commission for determination of
the admissibility of the contention. All other issues with respect
to the plant-specific TS or other operational requirements are
subject to a hearing as part of the license proceeding.
6. After issuance of a license, the generic TS have no further
effect on the plant-specific TS. Changes to the plant-specific TS
will be treated as license amendments under 10 CFR 50.90.
IX. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
A.1. An applicant or licensee who references this appendix shall
perform and demonstrate conformance with the ITAAC before fuel load.
With respect to activities subject to an ITAAC, an applicant for a
license may proceed at its own risk with design and procurement
activities. A licensee may also proceed at its own risk with design,
procurement, construction, and preoperational activities, even
though the NRC may not have found that any particular ITAAC has been
met.
2. The licensee who references this appendix shall notify the
NRC that the required inspections, tests, and analyses in the ITAAC
have been successfully completed and that the corresponding
acceptance criteria have been met.
3. If an activity is subject to an ITAAC and the applicant or
licensee who references this appendix has not demonstrated that the
ITAAC has been met, the applicant or licensee may either take
corrective actions to successfully complete that ITAAC, request an
exemption from the ITAAC under Section VIII of this appendix and 10
CFR 52.97(b), or petition for rulemaking to amend this appendix by
changing the requirements of the ITAAC, under 10 CFR 2.802 and
52.97(b). Such rulemaking changes to the ITAAC must meet the
requirements of paragraph VIII.A.1 of this appendix.
B.1. The NRC shall ensure that the required inspections, tests,
and analyses in the ITAAC are performed. The NRC shall verify that
the inspections, tests, and analyses referenced by the licensee have
been successfully completed and, based solely thereon, find that the
prescribed acceptance criteria have been met. At appropriate
intervals during construction, the NRC shall publish notices of the
successful completion of ITAAC in the Federal Register.
2. In accordance with 10 CFR 52.103(g), the Commission shall
find that the acceptance criteria in the ITAAC for the license are
met before fuel load.
3. After the Commission has made the finding required by 10 CFR
52.103(g), the ITAAC do not, by virtue of their inclusion within the
DCD, constitute regulatory requirements either for licensees or for
renewal of the license; except for specific ITAAC, which are the
subject of a Sec. 52.103(a) hearing, their expiration will occur
upon final Commission action in such a
[[Page 49559]]
proceeding. However, subsequent modifications must comply with the
Tier 1 and Tier 2 design descriptions in the plant-specific DCD
unless the licensee has complied with the applicable requirements of
10 CFR 52.98 and Section VIII of this appendix.
X. Records and Reporting
A. Records
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes to Tier 1, Tier 2, and
the generic TS and other operational requirements. The applicant
shall maintain the proprietary and safeguards information referenced
in the generic DCD for the period that this appendix may be
referenced, as specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application
and for the term of the license (including any period of renewal).
3. An applicant or licensee who references this appendix shall
prepare and maintain written evaluations which provide the bases for
the determinations required by Section VIII of this appendix. These
evaluations must be retained throughout the period of application
and for the term of the license (including any period of renewal).
B. Reporting
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any
plant-specific departures from the DCD, including a summary of the
evaluation of each. This report must be filed in accordance with the
filing requirements applicable to reports in 10 CFR 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its DCD, which reflect the generic changes to and
plant-specific departures from the generic DCD made under Section
VIII of this appendix. These updates must be filed under the filing
requirements applicable to final safety analysis report updates in
10 CFR 52.3 and 50.71(e).
3. The reports and updates required by paragraphs X.B.1 and
X.B.2 must be submitted as follows:
a. On the date that an application for a license referencing
this appendix is submitted, the application must include the report
and any updates to the generic DCD.
b. During the interval from the date of application for a
license to the date the Commission makes its findings required by 10
CFR 52.103(g), the report must be submitted semi-annually. Updates
to the plant-specific DCD must be submitted annually and may be
submitted along with amendments to the application.
c. After the Commission makes the finding required by 10 CFR
52.103(g), the reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the
final safety analysis report for the facility, at the intervals
required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at
shorter intervals as specified in the license.
Appendices E Through M to Part 52 [Reserved]
Appendix N to Part 52--Standardization of Nuclear Power Plant Designs:
Combined Licenses To Construct and Operate Nuclear Power Reactors of
Identical Design at Multiple Sites
The Commission's regulations in part 2 of this chapter
specifically provide for the holding of hearings on particular
issues separately from other issues involved in hearings in
licensing proceedings, and for the consolidation of adjudicatory
proceedings and of the presentations of parties in adjudicatory
proceedings such as licensing proceedings (Sec. Sec. 2.316 and
2.317 of this chapter).
This appendix sets out the particular requirements and
provisions applicable to situations in which applications for
combined licenses under subpart C of this part are filed by one or
more applicants for licenses to construct and operate nuclear power
reactors of identical design (``common design'') to be located at
multiple sites.\1\
---------------------------------------------------------------------------
\1\ If the design for the power reactor(s) proposed in a
particular application is not identical to the others, that
application may not be processed under this appendix and subpart D
of part 2 of this chapter.
---------------------------------------------------------------------------
1. Except as otherwise specified in this appendix or as the
context otherwise indicates, the provisions of subpart C of this
part and subpart D of part 2 of this chapter apply to combined
license applications subject to this appendix.
2. Each combined license application submitted pursuant to this
appendix must be submitted as specified in Sec. 52.75 and 10 CFR
2.101. Each application must state that the applicant wishes to have
the application considered under 10 CFR part 52, appendix N, and
must list each of the applications to be treated together under this
appendix.
3. Each application must include the information required by
Sec. Sec. 52.77, 52.79, and 52.80(a), provided however, that the
application must identify the common design, and, if applicable,
reference a standard design certification under subpart B of this
part, or the use of a reactor manufactured under subpart F of this
part. The final safety analysis report for each application must
either incorporate by reference or include the final safety analysis
of the common design, including, if applicable, the final safety
analysis report for the referenced design certification or the
manufactured reactor.\2\
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\2\ As used in this appendix, the design of a nuclear power
reactor included in a single referenced safety analysis report means
the design of those structures, systems, and components important to
radiological health and safety and the common defense and security.
---------------------------------------------------------------------------
4. Each combined license application submitted pursuant to this
appendix must contain an environmental report as required by Sec.
52.80(b), and which complies with the applicable provisions of 10
CFR part 51, provided, however, that the application may incorporate
by reference a single environmental report on the environmental
impacts of the common design.
5. Upon a determination that each application is acceptable for
docketing under 10 CFR 2.101, each application will be docketed and
a notice of docketing for each application will be published in the
Federal Register, in accordance with 10 CFR 2.104, provided,
however, that the notice must state that the application will be
processed under the provisions of 10 CFR part 52, appendix N, and
subpart D of part 2 of this chapter. As the discretion of the
Commission, a single notice of docketing for multiple applications
may be published in the Federal Register.
6. The NRC staff shall prepare draft and final environmental
impact statements for each of the applications under part 51 of this
chapter. Scoping under 10 CFR 51.28 and 51.29 for each of the
combined license applications may be conducted simultaneously and
joint scoping may be conducted with respect to the environmental
issues relevant to the common design.
If the applications reference a standard design certification,
then the environmental impact statement for each of the applications
must incorporate by reference the design certification environmental
assessment. If the applications do not reference a standard design
certification, then the NRC staff shall prepare draft and final
supplemental environmental impact statements which address severe
accident mitigation design alternatives for the common design, which
must be incorporated by reference into the environmental impact
statement prepared for each application. Scoping under 10 CFR 51.28
and 51.29 for the supplemental environmental impact statement may be
conducted simultaneously, and may be part of the scoping for each of
the combined license applications.
7. The ACRS shall report on each of the applications as required
by Sec. 52.87. Each report must be limited to those safety matters
for each application which are not relevant to the common design. In
addition, the ACRS shall separately report on the safety of the
common design, provided, however, that the report need not address
the safety of a referenced standard design certification or reactor
manufactured under subpart F of this part.
8. The Commission shall designate a presiding officer to conduct
the proceeding with respect to the health and safety, common defense
and security, and environmental matters relating to the common
design. The hearing will be governed by the applicable provisions of
subparts A, C, G, L, N, and O of part 2 of this chapter relating to
applications for combined licenses. The presiding officer shall
issue a partial initial decision on the common design.
PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR
POWER PLANTS
0
151. The authority citation for part 54 continues to read as follows:
Authority: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68
Stat. 936, 937, 938, 948, 953, 954, 955, as amended, sec. 234, 83
Stat. 1244, as amended (42 U.S.C. 2132, 2133,
[[Page 49560]]
2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs 201, 202, 206,
88 Stat. 1242, 1244 as amended (42 U.S.C. 5841, 5842).
Section 54.17 also issued under E.O. 12829, 3 CFR, 1993 Comp.,
p. 570; E.O. 12958, as amended, 3 CFR, 1995 Comp., p. 333; E.O.
12968, 3 CFR, 1995 Comp., p. 391.
0
152. Section 54.1 is revised to read as follows:
Sec. 54.1 Purpose.
This part governs the issuance of renewed operating licenses and
renewed combined licenses for nuclear power plants licensed pursuant to
Sections 103 or 104b of the Atomic Energy Act of 1954, as amended, and
Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242)
0
153. In Sec. 54.3, paragraph (a), the definition for Current licensing
basis is revised, and the definition for Renewed combined license is
added to read as follows:
Sec. 54.3 Definitions.
(a) * * *
Current licensing basis (CLB) is the set of NRC requirements
applicable to a specific plant and a licensee's written commitments for
ensuring compliance with and operation within applicable NRC
requirements and the plant-specific design basis (including all
modifications and additions to such commitments over the life of the
license) that are docketed and in effect. The CLB includes the NRC
regulations contained in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50,
51, 52, 54, 55, 70, 72, 73, 100 and appendices thereto; orders; license
conditions; exemptions; and technical specifications. It also includes
the plant-specific design-basis information defined in 10 CFR 50.2 as
documented in the most recent final safety analysis report (FSAR) as
required by 10 CFR 50.71 and the licensee's commitments remaining in
effect that were made in docketed licensing correspondence such as
licensee responses to NRC bulletins, generic letters, and enforcement
actions, as well as licensee commitments documented in NRC safety
evaluations or licensee event reports.
* * * * *
Renewed combined license means a combined license originally issued
under part 52 of this chapter for which an application for renewal is
filed in accordance with 10 CFR 52.107 and issued under this part.
* * * * *
0
154. In Sec. 54.17, paragraph (c) is revised to read as follows:
Sec. 54.17 Filing of application.
* * * * *
(c) An application for a renewed license may not be submitted to
the Commission earlier than 20 years before the expiration of the
operating license or combined license currently in effect.
* * * * *
0
155. Section 54.27 is revised to read as follows:
Sec. 54.27 Hearings.
A notice of an opportunity for a hearing will be published in the
Federal Register in accordance with 10 CFR 2.105. In the absence of a
request for a hearing filed within 30 days by a person whose interest
may be affected, the Commission may issue a renewed operating license
or renewed combined license without a hearing upon 30-day notice and
publication in the Federal Register of its intent to do so.
0
156. In Section 54.31, paragraphs (a), (b), and (c) are revised to read
as follows:
Sec. 54.31 Issuance of a renewed license.
(a) A renewed license will be of the class for which the operating
license or combined license currently in effect was issued.
(b) A renewed license will be issued for a fixed period of time,
which is the sum of the additional amount of time beyond the expiration
of the operating license or combined license (not to exceed 20 years)
that is requested in a renewal application plus the remaining number of
years on the operating license or combined license currently in effect.
The term of any renewed license may not exceed 40 years.
(c) A renewed license will become effective immediately upon its
issuance, thereby superseding the operating license or combined license
previously in effect. If a renewed license is subsequently set aside
upon further administrative or judicial appeal, the operating license
or combined license previously in effect will be reinstated unless its
term has expired and the renewal application was not filed in a timely
manner.
* * * * *
0
157. Section 54.35 is revised to read as follows:
Sec. 54.35 Requirements during term of renewed license.
During the term of a renewed license, licensees shall be subject to
and shall continue to comply with all Commission regulations contained
in 10 CFR parts 2, 19, 20, 21, 26, 30, 40, 50, 51, 52, 54, 55, 70, 72,
73, and 100, and the appendices to these parts that are applicable to
holders of operating licenses or combined licenses, respectively.
0
158. In Sec. 54.37, paragraph (a) is revised to read as follows:
Sec. 54.37 Additional records and recordkeeping requirements.
(a) The licensee shall retain in an auditable and retrievable form
for the term of the renewed operating license or renewed combined
license all information and documentation required by, or otherwise
necessary to document compliance with, the provisions of this part.
* * * * *
PART 55--OPERATORS' LICENSES
0
159. The authority citation for part 55 continues to read as follows:
Authority: Secs. 107, 161, 182, 68 Stat. 939, 948, 953, as
amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2137, 2201,
2232, 2282); secs. 201, as amended, 202, 88 Stat. 1242, as amended,
1244 (42 U.S.C. 5841, 5842); sec. 1704, 112 Stat. 2750 (44 U.S.C.
3504 note). Sections 55.41, 55.43, 55.45, and 55.59 also issued
under sec. 306, Pub. L. 97-425, 96 Stat. 2262 (42 U.S.C. 10226).
Section 55.61 also issued under secs. 186, 187, 68 Stat. 955 (42
U.S.C. 2236, 2237).
0
160. In Sec. 55.1, paragraph (a) is revised to read as follows:
Sec. 55.1 Purpose.
* * * * *
(a) Establish procedures and criteria for the issuance of licenses
to operators and senior operators of utilization facilities licensed
under the Atomic Energy Act of 1954, as amended, or Section 202 of the
Energy Reorganization Act of 1974, as amended, and part 50, part 52, or
part 54 of this chapter,
* * * * *
0
161. In Sec. 55.2, paragraph (a) is revised to read as follows:
Sec. 55.2 Scope.
* * * * *
(a) Any individual who manipulates the controls of any utilization
facility licensed under parts 50, 52, or 54 of this chapter,
* * * * *
0
162. In Sec. 55.5, paragraph (b)(1) and the introductory text of
paragraph (b)(2) are revised to read as follows:
Sec. 55.5 Communications.
* * * * *
(b)(1) Except for test and research reactor facilities, the
Director of New Reactors or the Director of Nuclear Reactor Regulation,
as appropriate, has delegated to the Regional Administrators of Regions
I, II, III, and IV authority and responsibility under the regulations
in this part for the
[[Page 49561]]
issuance and renewal of licenses for operators and senior operators of
nuclear power reactors licensed under 10 CFR part 50 or part 52 and
located in these regions.
(2) Any application for a license or license renewal filed under
the regulations in this part involving a nuclear power reactor licensed
under 10 CFR part 50 or part 52 and any related inquiry, communication,
information, or report must be submitted to the Regional Administrator
by an appropriate method listed in paragraph (a) of this section. The
Regional Administrator or the Administrator's designee will transmit to
the Director of New Reactors or the Director of Nuclear Reactor
Regulation, as appropriate, any matter that is not within the scope of
the Regional Administrator's delegated authority.
* * * * *
PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR
RELATED GREATER THAN CLASS C WASTE
0
163. The authority citation for part 72 continues to read as follows:
Authority: Secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183,
184, 186, 187, 189, 68 Stat. 929, 930, 932, 933, 934, 935, 948, 953,
954, 955, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.
2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2232, 2233,
2234, 2236, 2237, 2238, 2282); sec. 274, Pub. L. 86-373, 73 Stat.
688, as amended (42 U.S.C. 2021); sec. 201, as amended, 202, 206, 88
Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);
Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L. 102-
486, sec. 7902, 106 Stat. 3123 (42 U.S.C. 5851); sec. 102, Pub. L.
91-190, 83 Stat. 853 (42 U.S.C. 4332); secs. 131, 132, 133, 135,
137, 141, Pub. L. 97-425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148,
Pub. L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10151, 10152, 10153,
10155, 10157, 10161, 10168); sec. 1704, 112 Stat. 2750 (44 U.S.C.
3504 note).
Section 72.44(g) also issued under secs. 142(b) and 148(c), (d),
Pub. L. 100-203, 101 Stat. 1330-232, 1330-236 (42 U.S.C. 10162(b),
10168(c), (d)). Section 72.46 also issued under sec. 189, 68 Stat.
955 (42 U.S.C. 2239); sec. 134, Pub. L. 97-425, 96 Stat. 2230 (42
U.S.C. 10154). Section 72.96(d) also issued under sec. 145(g), Pub.
L. 100-203, 101 Stat. 1330-235 (42 U.S.C. 10165(g)). Subpart J also
issued under secs. 2(2), 2(15), 2(19), 117(a), 141(h), Pub. L. 97-
425, 96 Stat. 2202, 2203, 2204, 2222, 2224 (42 U.S.C. 10101,
10137(a), 10161(h)). Subparts K and L are also issued under sec.
133, 98 Stat. 2230 (42 U.S.C. 10153) and sec. 218(a), 96 Stat. 2252
(42 U.S.C. 10198).
0
164. Section 72.210 is revised to read as follows:
Sec. 72.210 General license issued.
A general license is hereby issued for the storage of spent fuel in
an independent spent fuel storage installation at power reactor sites
to persons authorized to possess or operate nuclear power reactors
under 10 CFR part 50 or 10 CFR part 52.
0
165. In Sec. 72.218, paragraph (b) is revised to read as follows:
Sec. 72.218 Termination of licenses.
* * * * *
(b) An application for termination of a reactor operating license
issued under 10 CFR part 50 and submitted under Sec. 50.82 of this
chapter, or a combined license issued under 10 CFR part 52 and
submitted under Sec. 52.110 of this chapter, must contain a
description of how the spent fuel stored under this general license
will be removed from the reactor site.
* * * * *
PART 73--PHYSICAL PROTECTION OF PLANTS AND MATERIALS
0
166. The authority citation for part 73 continues to read as follows:
Authority: Secs. 53, 161, 68 Stat. 930, 948, as amended, sec.
147, 94 Stat. 780 (42 U.S.C. 2073, 2167, 2201); sec. 201, as
amended, 204, 88 Stat. 1242, as amended, 1245, sec. 1701, 106 Stat.
2951, 2952, 2953 (42 U.S.C. 5841, 5844, 2297f); sec. 1704, 112 Stat.
2750 (44 U.S.C. 3504 note).
Section 73.1 also issued under secs. 135, 141, Pub. L. 97-425,
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161). Section 73.37(f) also
issued under sec. 301, Pub. L. 96-295, 94 Stat. 789 (42 U.S.C. 5841
note). Section 73.57 is issued under sec. 606, Pub. L. 99-399, 100
Stat. 876 (42 U.S.C. 2169).
0
167. In Sec. 73.1, paragraph (b)(1)(i) is revised to read as follows:
Sec. 73.1 Purpose and scope.
* * * * *
(b) * * *
(1) * * *
(i) The physical protection of production and utilization
facilities licensed under parts 50 or 52 of this chapter,
* * * * *
0
168. In Sec. 73.2, the introductory text of paragraph (a) is revised
to read as follows:
Sec. 73.2 Definitions.
* * * * *
(a) Terms defined in parts 50, 52, and 70 of this chapter have the
same meaning when used in this part.
* * * * *
0
169. In Sec. 73.50, the introductory text is revised to read as
follows:
Sec. 73.50 Requirements for physical protection of licensed
activities.
Each licensee who is not subject to Sec. 73.51, but who possesses,
uses, or stores formula quantities of strategic special nuclear
material that are not readily separable from other radioactive material
and which have total external radiation dose rates in excess of 100
rems per hour at a distance of 3 feet from any accessible surfaces
without intervening shielding other than at nuclear reactor facility
licensed under parts 50 or 52 of this chapter, shall comply with the
following:
* * * * *
0
170. In Sec. 73.56, paragraph (a)(3) is revised to read as follows:
Sec. 73.56 Personnel access authorization requirements for nuclear
power plants.
(a) * * *
(3) Each applicant for a license to operate a nuclear power reactor
under Sec. Sec. 50.21(b) or 50.22 of this chapter, including an
applicant for a combined license under part 52 of this chapter, whose
application is submitted after April 25, 1991, shall include the
required access authorization program as part of its Physical Security
Plan. The applicant, upon receipt of an operating license or upon
notice of the Commission's finding under Sec. 52.103(g) of this
chapter, shall implement the required access authorization program as
part of its site Physical Security Plan.
* * * * *
0
171. In Sec. 73.57, paragraphs (a)(1), (a)(2), and (a)(3) are revised
to read as follows:
Sec. 73.57 Requirements for criminal history checks of individuals
granted unescorted access to a nuclear power facility or access to
Safeguards Information by power reactor licensees.
(a) * * *
(1) Each licensee who is authorized to operate a nuclear power
reactor under part 50 of this chapter, or each holder of a combined
license under part 52 of this chapter upon receipt of notice of the
Commission's finding under Sec. 52.103(g), shall comply with the
requirements of this section.
(2) Each applicant for a license to operate a nuclear power reactor
under part 50 of this chapter and each applicant for a combined license
under part 52 of this chapter shall submit fingerprints for those
individuals who have or will have access to Safeguards Information.
(3) Before receiving its operating license under part 50 of this
chapter or before the Commission makes its finding under Sec.
52.103(g) of this chapter, each applicant for a license to
[[Page 49562]]
operate a nuclear power reactor (including an applicant for a combined
license) may submit fingerprints for those individuals who will require
unescorted access to the nuclear power facility.
* * * * *
0
172. In Appendix C to Part 73, the Introduction is revised to read as
follows:
Appendix C to Part 73--Licensee Safeguards Contingency Plans
Introduction
A licensee safeguards contingency plan is a documented plan to
give guidance to licensee personnel in order to accomplish specific
defined objectives in the event of threats, thefts, or radiological
sabotage relating to special nuclear material or nuclear facilities
licensed under the Atomic Energy Act of 1954, as amended. An
acceptable safeguards contingency plan must contain:
(1) A predetermined set of decisions and actions to satisfy
stated objectives;
(2) An identification of the data, criteria, procedures, and
mechanisms necessary to efficiently implement the decisions; and
(3) A stipulation of the individual, group, or organizational
entity responsible for each decision and action.
The goals of licensee safeguards contingency plans for
responding to threats, thefts, and radiological sabotage are:
(1) To organize the response effort at the licensee level;
(2) To provide predetermined, structured responses by licensees
to safeguards contingencies;
(3) To ensure the integration of the licensee response with the
responses by other entities; and
(4) To achieve a measurable performance in response capability.
Licensee safeguards contingency planning should result in
organizing the licensee's resources in such a way that the
participants will be identified, their several responsibilities
specified, and the responses coordinated. The responses should be
timely.
It is important to note that a licensee's safeguards contingency
plan is intended to be complementary to any emergency plans
developed under appendix E to part 50 of this chapter, Sec. 52.17
or Sec. 52.79, or to Sec. 70.22(i) of this chapter.
* * * * *
PART 75--SAFEGUARDS ON NUCLEAR MATERIAL--IMPLEMENTATION OF US/IAEA
AGREEMENT
0
173. The authority citation for part 75 continues to read as follows
Authority: Secs. 53, 63, 103, 104, 122, 161, 68 Stat. 930, 932,
936, 937, 939, 948, as amended (42 U.S.C. 2073, 2093, 2133, 2134,
2152, 2201); sec. 201, 88 Stat. 1242, as amended (42 U.S.C. 5841);
sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
Section 75.4 also issued under secs. 135, 141, Pub. L. 97-425,
96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161).
0
174. In Sec. 75.6, paragraph (b) is revised to read as follows:
Sec. 75.6 Maintenance of records and delivery of information,
reports, and other communications.
* * * * *
(b) If an installation is a nuclear power plant or a non-power
reactor for which a construction permit, operating license or a
combined license has been issued, whether or not a license to receive
and possess nuclear material at the installation has been issued, the
cognizant Director is either the Director, Office of New Reactors, or
the Director, Office of Nuclear Reactor Regulation. For all other
installations, the cognizant Director is the Director, Office of
Nuclear Material Safety and Safeguards.
* * * * *
PART 95--FACILITY SECURITY CLEARANCE AND SAFEGUARDING OF NATIONAL
SECURITY INFORMATION AND RESTRICTED DATA
0
175. The authority citation for Part 95 continues to read as follows:
Authority: Secs. 145, 161, 193, 68 Stat. 942, 948, as amended
(42 U.S.C. 2165, 2201); sec. 201, 88 Stat. 1242, as amended (42
U.S.C. 5841); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); E.O.
10865, as amended, 3 CFR 1959-1963 COMP., p. 398 (50 U.S.C. 401,
note); E.O. 12829, 3 CFR, 1993 Comp., p. 570; E.O. 12958, as
amended, 3 CFR, 1995 Comp., p. 333, as amended by E.O. 13292, 3 CFR,
2004 Comp., p. 196; E.O. 12968, 3 CFR, 1995 Comp., p. 391.
0
176. In Sec. 95.5, the definition of license is revised to read as
follows:
Sec. 95.5 Definitions.
* * * * *
License means a license issued under 10 CFR parts 50, 52, 54, 60,
63, 70, or 72.
* * * * *
0
177. In Sec. 95.13, paragraph (b) is revised to read as follows:
Sec. 95.13 Maintenance of records.
* * * * *
(b) Each record required by this part must be legible throughout
the retention period specified by each Commission regulation. The
record may be the original or a reproduced copy or a microform provided
that the copy or microform is authenticated by authorized personnel and
that the microform is capable of producing a clear copy throughout the
required retention period. The record may also be stored in electronic
media with the capability for producing legible, accurate, and complete
records during the required retention period. Records such as letters,
drawings, or specifications must include all pertinent information such
as stamps, initials, and signatures. The licensee, certificate holder,
or other person shall maintain adequate safeguards against tampering
with and loss of records.
0
178. In Sec. 95.19, the introductory text of paragraph (b) is revised
to read as follows:
Sec. 95.19 Changes to security practices and procedures.
* * * * *
(b) A licensee, certificate holder, or other person may effect a
minor, non-substantive change to an approved Standard Practice
Procedures Plan for the safeguarding of classified information without
receiving prior CSA approval. These minor changes that do not affect
the security of the facility may be submitted to the addressees noted
in paragraph (a) of this section within 30 days of the change. Page
changes rather than a complete rewrite of the plan may be submitted.
Some examples of minor, non-substantive changes to the Standard
Practice Procedures Plan include--
* * * * *
0
179. Section 95.20 is revised to read as follows:
Sec. 95.20 Grant, denial or termination of facility clearance.
The Division of Nuclear Security shall provide notification in
writing (or orally with written confirmation) to the licensee,
certificate holder, or other person of the Commission's grant,
acceptance of another agency's facility clearance, denial, or
termination of facility clearance. This information must also be
furnished to representatives of the NRC, NRC contractors, licensees,
certificate holders, or other person, or other Federal agencies having
a need to transmit classified information to the licensees or other
person.
0
180. In Sec. 95.23, paragraph (b) is revised to read as follows:
Sec. 95.23 Termination of facility clearance.
* * * * *
(b) When facility clearance is terminated, the licensee,
certificate holder, or other person will be notified in writing of the
determination and the procedures outlined in Sec. 95.53 apply.
0
181. Section 95.31 is revised to read as follows:
[[Page 49563]]
Sec. 95.31 Protective personnel.
Whenever protective personnel are used to protect classified
information they shall:
(a) Possess an ``L'' access authorization (or CSA equivalent) if
the licensee, certificate holder, or other person possesses information
classified Confidential National Security Information, Confidential
Restricted Data or Secret National Security Information.
(b) Possess a ``Q'' access authorization (or CSA equivalent) if the
licensee, certificate holder, or other person possesses Secret
Restricted Data related to nuclear weapons design, manufacturing and
vulnerability information; and certain particularly sensitive Naval
Nuclear Propulsion Program information (e.g., fuel manufacturing
technology) and the protective personnel require access as part of
their regular duties.
0
182. In Sec. 95.33, paragraph (c) is revised to read as follows:
Sec. 95.33 Security education.
* * * * *
(c) Temporary Help Suppliers. A temporary help supplier, or other
contractor who employs cleared individuals solely for dispatch
elsewhere, is responsible for ensuring that required briefings are
provided to their cleared personnel. The temporary help supplier or the
using licensee's, certificate holder's, or other person's facility may
conduct these briefings.
* * * * *
0
183. Section 95.34 is revised to read as follows:
Sec. 95.34 Control of visitors.
(a) Uncleared visitors. Licensees, certificate holders, or other
persons subject to this part shall take measures to preclude access to
classified information by uncleared visitors.
(b) Foreign visitors. Licensees, certificate holders, or other
persons subject to this part shall take measures as may be necessary to
preclude access to classified information by foreign visitors. The
licensee, certificate holder, or other person shall retain records of
visits for 5 years beyond the date of the visit.
0
184. In Sec. 95.35, the introductory text of paragraph (a), and
paragraph (a)(3) are revised to read as follows:
Sec. 95.35 Access to matter classified as National Security
Information and Restricted Data.
(a) Except as the Commission may authorize, no licensee,
certificate holder or other person subject to the regulations in this
part may receive or may permit any other licensee, certificate holder,
or other person to have access to matter revealing Secret or
Confidential National Security Information or Restricted Data unless
the individual has:
* * * * *
(3) NRC-approved storage facilities if classified documents or
material are to be transmitted to the licensee, certificate holder, or
other person.
* * * * *
0
185. In Sec. 95.36, paragraphs (c), (d), and (e) are revised to read
as follows:
Sec. 95.36 Access by representatives of the International Atomic
Energy Agency or by participants in other international agreements.
* * * * *
(c) In accordance with the specific disclosure authorization
provided by the Division of Nuclear Security, licensees, certificate
holders, or other persons subject to this part are authorized to
release (i.e., transfer possession of) copies of documents that contain
classified National Security Information directly to IAEA inspectors
and other representatives officially designated to request and receive
classified National Security Information documents. These documents
must be marked specifically for release to IAEA or other international
organizations in accordance with instructions contained in the NRC's
disclosure authorization letter. Licensees, certificate holders, and
other persons subject to this part may also forward these documents
through the NRC to the international organization's headquarters in
accordance with the NRC disclosure authorization. Licensees,
certificate holders, and other persons may not reproduce documents
containing classified National Security Information except as provided
in Sec. 95.43.
(d) Records regarding these visits and inspections must be
maintained for 5 years beyond the date of the visit or inspection.
These records must specifically identify each document released to an
authorized representative and indicate the date of the release. These
records must also identify (in such detail as the Division of Nuclear
Security, by letter, may require) the categories of documents that the
authorized representative has had access and the date of this access. A
licensee, certificate holder, or other person subject to this part
shall also retain Division of Nuclear Security disclosure
authorizations for 5 years beyond the date of any visit or inspection
when access to classified information was permitted.
(e) Licensees, certificate holders, or other persons subject to
this part shall take such measures as may be necessary to preclude
access to classified matter by participants of other international
agreements unless specifically provided for under the terms of a
specific agreement.
0
186. In Sec. 95.37, paragraphs (a), (b), and (h) are revised to read
as follows:
Sec. 95.37 Classification and preparation of documents.
(a) Classification. Classified information generated or possessed
by a licensee, certificate holder, or other person must be
appropriately marked. Classified material which is not conducive to
markings (e.g., equipment) may be exempt from this requirement. These
exemptions are subject to the approval of the CSA on a case-by-case
basis. If a person or facility generates or possesses information that
is believed to be classified based on guidance provided by the NRC or
by derivation from classified documents, but which no authorized
classifier has determined to be classified, the information must be
protected and marked with the appropriate classification markings
pending review and signature of an NRC authorized classifier. This
information shall be protected as classified information pending final
determination.
(b) Classification consistent with content. Each document
containing classified information shall be classified Secret or
Confidential according to its content. NRC licensees, certificate
holders, or other persons subject to the requirements of 10 CFR part 95
may not make original classification decisions.
* * * * *
(h) Classification challenges. Licensees, certificate holders, or
other persons in authorized possession of classified National Security
Information who in good faith believe that the information's
classification status (i.e., that the document), is classified at
either too high a level for its content (overclassification) or too low
for its content (underclassification) are expected to challenge its
classification status. Licensees, certificate holders, or other persons
who wish to challenge a classification status shall--
(1) Refer the document or information to the originator or to an
authorized NRC classifier for review. The authorized classifier shall
review the document and render a written classification decision to the
holder of the information.
(2) In the event of a question regarding classification review, the
[[Page 49564]]
holder of the information or the authorized classifier shall consult
the NRC Division of Facilities and Security, Information Security
Branch, for assistance.
(3) Licensees, certificate holders, or other persons who challenge
classification decisions have the right to appeal the classification
decision to the Interagency Security Classification Appeals Panel.
(4) Licensees, certificate holders, or other persons seeking to
challenge the classification of information will not be the subject of
retribution.
* * * * *
0
187. In Sec. 95.39, paragraph (a) is revised to read as follows:
Sec. 95.39 External transmission of documents and material.
(a) Restrictions. Documents and material containing classified
information received or originated in connection with an NRC license,
certificate, or standard design approval or standard design
certification under part 52 of this chapter must be transmitted only to
CSA approved security facilities.
* * * * *
0
188. In Sec. 95.43, paragraph (a) is revised to read as follows:
Sec. 95.43 Authority to reproduce.
(a) Each licensee, certificate holder, or other person possessing
classified information shall establish a reproduction control system to
ensure that reproduction of classified material is held to the minimum
consistent with operational requirements. Classified reproduction must
be accomplished by authorized employees knowledgeable of the procedures
for classified reproduction. The use of technology that prevents,
discourages, or detects the unauthorized reproduction of classified
documents is encouraged.
* * * * *
0
189. In Sec. 95.45, paragraph (d) is revised to read as follows:
Sec. 95.45 Changes in classification.
* * * * *
(d) Any licensee, certificate holder, or other person making a
change in classification or receiving notice of such a change shall
forward notice of the change in classification to holders of all copies
as shown on their records.
0
190. Section 95.49 is revised to read as follows:
Sec. 95.49 Security of automatic data processing (ADP) systems.
Classified data or information may not be processed or produced on
an ADP system unless the system and procedures to protect the
classified data or information have been approved by the CSA. Approval
of the ADP system and procedures is based on a satisfactory ADP
security proposal submitted as part of the licensee's, certificate
holder's, or other person's request for facility clearance outlined in
Sec. 95.15 or submitted as an amendment to its existing Standard
Practice Procedures Plan for the protection of classified information.
0
191. Section 95.51 is revised to read as follows:
Sec. 95.51 Retrieval of classified matter following suspension or
revocation of access authorization.
In any case where the access authorization of an individual is
suspended or revoked in accordance with the procedures set forth in
part 25 of this chapter, or other relevant CSA procedures, the
licensee, certificate holder, or other person shall, upon due notice
from the Commission of such suspension or revocation, retrieve all
classified information possessed by the individual and take the action
necessary to preclude that individual having further access to the
information.
0
192. Section 95.53 is revised to read as follows:
Sec. 95.53 Termination of facility clearance.
(a) If the need to use, process, store, reproduce, transmit,
transport, or handle classified matter no longer exists, the facility
clearance will be terminated. The licensee, certificate holder, or
other person for the facility may deliver all documents and matter
containing classified information to the Commission, or to a person
authorized to receive them, or must destroy all classified documents
and matter. In either case, the licensee, certificate holder, or other
person for the facility shall submit a certification of nonpossession
of classified information to the NRC Division of Nuclear Security
within 30 days of the termination of the facility clearance.
(b) In any instance where a facility clearance has been terminated
based on a determination of the CSA that further possession of
classified matter by the facility would not be in the interest of the
national security, the licensee, certificate holder, or other person
for the facility shall, upon notice from the CSA, dispose of classified
documents in a manner specified by the CSA.
0
193. In Sec. 95.57, the introductory paragraph is revised to read as
follows:
Sec. 95.57 Reports.
Each licensee, certificate holder, or other person having a
facility clearance shall report to the CSA and the Regional
Administrator of the appropriate NRC Regional Office listed in 10 CFR
part 73, appendix A:
* * * * *
0
194. Section 95.59 is revised to read as follows:
Sec. 95.59 Inspections.
The Commission shall make inspections and reviews of the premises,
activities, records and procedures of any licensee, certificate holder,
or other person subject to the regulations in this part as the
Commission and CSA deem necessary to effect the purposes of the Act,
E.O. 12958 and/or NRC rules.
PART 140--FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY
AGREEMENTS
0
195. The authority citation for part 140 continues to read as follows:
Authority: Secs. 161, 170, 68 Stat. 948, 71 Stat. 576, as
amended (42 U.S.C. 2201, 2210); secs. 201, as amended, 202, 88 Stat.
1242, as amended, 1244 (42 U.S.C. 841, 5842); Sec. 1704, 112 Stat.
2750 (44 U.S.C. 3504 note).
0
196. In Sec. 140.2, paragraphs (a)(1) and (a)(2) are revised to read
as follows:
Sec. 140.2 Scope.
(a) * * *
(1) To each person who is an applicant for or holder of a license
issued under 10 CFR parts 50, 52, or 54 to operate a nuclear reactor,
and
(2) With respect to an extraordinary nuclear occurrence, to each
person who is an applicant for or holder of a license to operate a
production facility or a utilization facility (including an operating
license issued under part 50 of this chapter and a combined license
under part 52 of this chapter), and to other persons indemnified with
respect to the involved facilities.
* * * * *
0
197. Section 140.10 is revised to read as follows:
Sec. 140.10 Scope.
This subpart applies to each person who is an applicant for or
holder of a license issued under 10 CFR parts 50 or 54 to operate a
nuclear reactor, or is the applicant for or holder of a combined
license issued under parts 52 or 54 of this chapter, except licenses
held by persons found by the Commission to be Federal agencies or
nonprofit educational institutions licensed to conduct educational
activities. This subpart also applies to persons licensed
[[Page 49565]]
to possess and use plutonium in a plutonium processing and fuel
fabrication plant.
0
198. In Sec. 140.11, paragraph (b) is revised to read as follows:
Sec. 140.11 Amounts of financial protection for certain reactors.
* * * * *
(b) In any case where a person is authorized under parts 50, 52, or
54 of this chapter to operate two or more nuclear reactors at the same
location, the total primary financial protection required of the
licensee for all such reactors is the highest amount which would
otherwise be required for any one of those reactors; provided, that
such primary financial protection covers all reactors at the location.
0
199. In Sec. 140.12, paragraph (c) is revised to read as follows:
Sec. 140.12 Amount of financial protection required for other
reactors.
* * * * *
(c) In any case where a person is authorized under parts 50, 52, or
54 of this chapter to operate two or more nuclear reactors at the same
location, the total financial protection required of the licensee for
all such reactors is the highest amount which would otherwise be
required for any one of those reactors; provided, that such financial
protection covers all reactors at the location.
* * * * *
0
200. Section 140.13 is revised to read as follows:
Sec. 140.13 Amount of financial protection required of certain
holders of construction permits and combined licenses under 10 CFR part
52.
Each holder of a part 50 construction permit, or a holder of a
combined license under part 52 of this chapter before the date that the
Commission had made the finding under 10 CFR 52.103(g), who also holds
a license under part 70 of this chapter authorizing ownership,
possession and storage only of special nuclear material at the site of
the nuclear reactor for use as fuel in operation of the nuclear reactor
after issuance of either an operating license under 10 CFR part 50 or
combined license under 10 CFR part 52, shall, during the period before
issuance of a license authorizing operation under 10 CFR part 50, or
the period before the Commission makes the finding under Sec.
52.103(g) of this chapter, as applicable, have and maintain financial
protection in the amount of $1,000,000. Proof of financial protection
shall be filed with the Commission in the manner specified in Sec.
140.15 of this chapter before issuance of the license under part 70 of
this chapter.
0
201. In Sec. 140.20, paragraph (a)(1)(ii) is revised, and paragraph
(a)(1)(iii) is added to read as follows:
Sec. 140.20 Indemnity agreements and liens.
(a) * * *
(1) * * *
(ii) The date that the Commission makes the finding under Sec.
52.103(g) of this chapter; or
(iii) The effective date of the license (issued under part 70 of
this chapter) authorizing the licensee to possess and store special
nuclear material at the site of the nuclear reactor for use as fuel in
operation of the nuclear reactor after issuance of an operating license
for the reactor, whichever is earlier. No such agreement, however,
shall be effective prior to September 26, 1957; or
* * * * *
0
202. In Sec. 140.81, paragraph (a) is revised to read as follows:
Sec. 140.81 Scope and purpose.
(a) Scope. This subpart applies to applicants for and holders of
licenses authorizing operation of production facilities and utilization
facilities, including combined licenses under part 52 of this chapter,
and to other persons indemnified with respect to such facilities.
* * * * *
0
203. In Sec. 140.93 Appendix C, Article VIII, paragraph 4 is revised
to read as follows:
Sec. 140.93 Appendix C--Form of indemnity agreement with licensees
furnishing proof of financial protection in the form of licensee's
resources.
* * * * *
Article VIII
* * * * *
4. If the Commission determines that the licensee is financially
able to reimburse the Commission for a deferred premium payment made
in its behalf, and the licensee, after notice of such determination
by the Commission fails to make such reimbursement within 120 days,
the Commission will take appropriate steps to suspend the license
for 30 days. The Commission may take any further action as necessary
if reimbursement is not made within the 30-day suspension period
including, but not limited to, termination of the operating license
or combined license.
* * * * *
0
204. Section 140.96 is revised to read as follows:
Sec. 140.96 Appendix F--Indemnity locations.
(a) Geographical boundaries of indemnity locations.
(1) In every indemnity agreement between the Commission and a
licensee which affords indemnity protection for the preoperational
storage of fuel at the site of a nuclear power reactor under
construction, the geographical boundaries of the indemnity location
will include the entire construction area of the nuclear power reactor,
as determined by the Commission. Such area will not necessarily be
coextensive with the indemnity location which will be established at
the time an operating license or combined license under 10 CFR part 52
is issued for such additional nuclear power reactors.
(2) In every indemnity agreement between the Commission and a
licensee which affords indemnity protection for an existing nuclear
power reactor, the geographical boundaries of the indemnity location
shall include the entire construction area of any additional nuclear
power reactor as determined by the Commission, built as part of the
same power station by the same licensee. Such area will not necessarily
be coextensive with the indemnity location which will be established at
the time an operating license or combined license is issued for such
additional nuclear power reactors.
(3) This section is effective May 1, 1973, as to construction
permits issued before March 2, 1973, and, as to construction permits
and combined licenses issued on or after March 2, 1973, the provisions
of this section will apply no later than such time as a construction
permit or combined license is issued authorizing construction of any
additional nuclear power reactor.
PART 170--FEES FOR FACILITIES, MATERIALS, IMPORT AND EXPORT
LICENSES, AND OTHER REGULATORY SERVICES UNDER THE ATOMIC ENERGY ACT
OF 1954, AS AMENDED
0
205. The authority citation for part 170 continues to read as follows:
Authority: Sec. 9701, Pub. L. 97-258, 96 Stat. 1051 (31 U.S.C.
9701); sec. 301, Pub. L. 92-314, 86 Stat. 227 (42 U.S.C. 2201w);
sec. 201, Pub. L. 93-438, 88 Stat. 1242, as amended (42 U.S.C.
5841); sec. 205a, Pub. L. 101-576, 104 Stat. 2842, as amended (31
U.S.C. 901, 902); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note).
0
206. In Sec. 170.2, paragraph (j) is removed and reserved, and
paragraphs (g) and (k) are revised to read as follows:
Sec. 170.2 Scope.
* * * * *
(g) An applicant for or holder of a production or utilization
facility construction permit or operating license issued under 10 CFR
part 50, or an early site permit, standard design
[[Page 49566]]
certification, standard design approval, manufacturing license, or
combined license issued under 10 CFR part 52;
* * * * *
(j) [Reserved]
(k) Applying for or already has applied for review, under appendix
Q to 10 CFR part 50 of a facility site before the submission of an
application for a construction permit;
* * * * *
PART 171--ANNUAL FEES FOR REACTOR LICENSES AND FUEL CYCLE LICENSES
AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF
COMPLIANCE, REGISTRATIONS, AND QUALITY ASSURANCE PROGRAM APPROVALS
AND GOVERNMENT AGENCIES LICENSED BY THE NRC
0
207. The authority citation for part 171 continues to read as follows:
Authority: Sec. 7601, Pub. L. 99-272, 100 Stat. 146, as amended
by sec. 5601, Pub. L. 100-203, 101 Stat. 1330 as amended by sec.
3201, Pub. L. 101-239, 103 Stat. 2132, as amended by sec. 6101, Pub.
L. 101-508, 104 Stat. 1388, as amended by sec. 2903a, Pub. L. 102-
486, 106 Stat. 3125 (42 U.S.C. 2213, 2214); sec. 301, Pub. L. 92-
314, 86 Stat. 227 (42 U.S.C. 2201w); sec. 201, Pub. L. 93-438, 88
Stat. 1242, as amended (42 U.S.C. 5841); sec. 1704, 112 Stat. 2750
(44 U.S.C. 3504 note).
0
208. In Sec. 171.15, paragraph (a) is revised to read as follows:
Sec. 171.15 Annual Fees: Reactor licenses and independent spent fuel
storage licenses.
(a) Each person holding an operating license for a power, test, or
research reactor; each person holding a combined license under part 52
of this chapter after the Commission has made the finding under Sec.
52.103(g); each person holding a part 50 or part 52 power reactor
license that is in decommissioning or possession only status, except
those that have no spent fuel onsite; and each person holding a part 72
license who does not hold a part 50 or part 52 license shall pay the
annual fee for each license held at any time during the Federal fiscal
year in which the fee is due. This paragraph does not apply to test and
research reactors exempted under Sec. 171.11(a).
* * * * *
Dated at Rockville, Maryland, this 1st day of August 2007.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 07-3861 Filed 8-20-07; 8:45 am]
BILLING CODE 7590-01-P