[Federal Register Volume 71, Number 48 (Monday, March 13, 2006)]
[Proposed Rules]
[Pages 12782-12932]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-1856]



[[Page 12781]]

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Part II





Nuclear Regulatory Commission





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10 CFR Parts 1, 2 et al.



Licenses, Certifications, and Approvals for Nuclear Power Plants; 
Proposed Rule

  Federal Register / Vol. 71, No. 48 / Monday, March 13, 2006 / 
Proposed Rules  

[[Page 12782]]


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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: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend 
its regulations by revising the provisions applicable to the licensing 
and approval processes for nuclear power plants and 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 proposed changes would clarify 
the applicability of various requirements to each of the licensing 
processes (i.e., early site permit, standard design approval, standard 
design certification, combined license, and manufacturing license). On 
July 3, 2003, the NRC published a proposed rulemaking to clarify and 
correct the NRC's regulations related to nuclear power plant licensing. 
Upon further consideration, the NRC is now proposing new requirements 
to enhance its licensing and approval processes and changes throughout 
the NRC's regulations to support these processes. This proposed rule 
supersedes the 2003 proposed rule. The Commission believes that this 
rulemaking action will improve the effectiveness and efficiency of the 
licensing and approval processes for future applicants.

DATES: Submit comments by May 30, 2006. Comments received after this 
date will be considered if it is practical to do so, but the Commission 
is able to ensure consideration only for comments received on or before 
this date.
    The NRC is holding a workshop on March 14, 2006 (see ADDRESSES 
section for the location).

ADDRESSES: You may submit comments by any one of the following methods. 
Please include the following number (RIN 3150-AG24) in the subject line 
of your comments. Comments on rulemakings submitted in writing or in 
electronic form will be made available to the public in their entirety 
on the NRC rulemaking Web site. Personal information will not be 
removed from your comments.
    Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
    E-mail comments to: [email protected]. If you do not receive a reply e-
mail confirming that we have received your comments, contact us 
directly at 301-415-1966. You may also submit comments via the NRC's 
rulemaking Web site at http://ruleforum.llnl.gov. Address questions 
about our rulemaking Web site to Carol Gallagher 301-415-5905; e-mail 
[email protected]. Comments may also be submitted via the Federal eRulemaking 
Portal http://www.regulations.gov.
    Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 
20852, between 7:30 a.m. and 4:15 p.m. Federal workdays. (Telephone 
301-415-1966.)
    Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 
301-415-1101.
    Publicly available documents related to this rulemaking may be 
examined and copied for a fee at the NRC's Public Document Room (PDR), 
Public File Area O1 F21, One White Flint North, 11555 Rockville Pike, 
Rockville, Maryland. Selected documents, including comments, can be 
viewed and downloaded electronically via the NRC rulemaking Web site at 
http://ruleforum.llnl.gov.
    Publicly available documents created or received at the NRC after 
November 1, 1999, are available electronically at the NRC's Electronic 
Reading Room at http://www.nrc.gov/NRC/ADAMS/index.html. From this 
site, the public can gain entry into the NRC's Agencywide Document 
Access and Management System (ADAMS), which provides text and image 
files of NRC's public documents. If you do not have access to ADAMS or 
if there are problems in accessing the documents located in ADAMS, 
contact the NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or 
by e-mail to [email protected].
    Workshop: The NRC workshop to be held on March 14, 2006, will take 
place in the Auditorium at the NRC offices at 11545 Rockville Pike, 
Rockville, Maryland, between 9 a.m. and 4 p.m. Please contact Nanette 
V. Gilles, Office of Nuclear Reactor Regulation, U.S. Nuclear 
Regulatory Commission, at telephone 301-415-1180 or e-mail [email protected] 
to pre-register for the workshop. Questions may be submitted in writing 
in advance of the workshop to Ms. Gilles at [email protected], or sent by 
mail to Ms. Gilles at the U.S. Nuclear Regulatory Commission, Mail Stop 
O-4D9A, Washington, DC 20555-0001.

FOR FURTHER INFORMATION CONTACT: Nanette V. Gilles, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, telephone 301-415-1180, e-mail [email protected]; or Jerry N. 
Wilson, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory 
Commission, Washington, D.C. 20555-0001, telephone 301-415-3145, e-mail 
[email protected].

SUPPLEMENTARY INFORMATION:

I. Workshop
II. Background
    A. Development of Proposed Rule
    B. Publication of Revised Proposed Rule
III. Reorganization of Part 52 and Conforming Changes in the NRC's 
Regulations
IV. Discussion of Substantive Changes
    A. Introduction.
    B. Testing Requirements for Advanced Reactors
    C. Proposed Changes to 10 CFR Part 52
    D. Proposed Changes to 10 CFR Part 50
    E. Proposed Change to 10 CFR Part 1
    F. Proposed Changes to 10 CFR Part 2
    G. Proposed Changes to 10 CFR Part 10
    H. Proposed Changes to 10 CFR Part 19
    I. Proposed Changes to 10 CFR Part 20
    J. Proposed Changes to 10 CFR Part 21
    K. Proposed Change to 10 CFR Part 25
    L. Proposed Changes to 10 CFR Part 26
    M. Proposed Changes to 10 CFR Part 51
    N. Proposed Changes to 10 CFR Part 54
    O. Proposed Changes to 10 CFR Part 55
    P. Proposed Changes to 10 CFR Part 72
    Q. Proposed Changes to 10 CFR Part 73
    R. Proposed Change to 10 CFR Part 75
    S. Proposed Changes to 10 CFR Part 95
    T. Proposed Changes to 10 CFR Part 140
    U. Proposed Changes to 10 CFR Part 170
V. Specific Request for Comments
VI. Availability of Documents
VII. Agreement State Compatibility
VIII. Plain Language
IX. Voluntary Consensus Standards
X. Environmental Impact--Categorical Exclusion
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Certification
XIV. Backfit Analysis

I. Workshop

    The NRC is holding a workshop on March 14, 2006, to provide 
additional information on the basis for the changes it is proposing in 
this document, to facilitate public discussion on the proposed 
rulemaking, and to answer stakeholder questions regarding the proposed 
rule. Questions may be submitted in writing in advance of the workshop 
as specified in the ADDRESSES section of this document. To facilitate 
complete and accurate responses to questions, the Commission requests 
that questions be submitted by March 10, 2006.
    Participants may provide informal oral comments during the 
workshop, but in order to receive a formal response in the final rule, 
participants must submit comments in writing as

[[Page 12783]]

indicated in the ADDRESSES section of this document. To aid the public 
in their development of comments on the proposed rule, the workshop 
will be transcribed and the transcript will be made available 
electronically at the NRC rulemaking Web site at http://ruleforum.llnl.gov. and at the NRC's Electronic Reading Room at http://www.nrc.gov/NRC/ADAMS/index.html.

II. 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. The NRC is issuing a revised proposed 
rule that rewrites part 52, makes changes throughout the Commission's 
regulations to ensure that all licensing processes in part 52 are 
addressed, and clarifies 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 supersedes 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) designs and codified these designs in appendices A, B, C, and 
D of 10 CFR part 52, respectively.
    The NRC had 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 
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 
initial 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 original 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 has decided to revise the July 2003 
proposed rule and publish the revised proposed rule for public comment. 
As discussed in more detail in Section III, Reorganization of Part 52 
and Conforming Changes in the NRC's regulations, this revised proposed 
rule contains 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 (i.e., early site 
permit, standard design approval, standard design certification, 
combined license, and manufacturing license). 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 has decided that 
developing responses to comments received on the July 2003 proposed 
rule is not an effective use of agency resources. The NRC requests that 
commenters on the July 2003 proposed rule who believe that their 
earlier comments are not adequately addressed in this proposed rule 
resubmit their comments. The NRC will provide resolutions for comments 
received on the revised proposed rule in the statement of 
considerations for the final rule. The NRC will not be providing a 
comment resolution for all of the comments received on the original 
July 2003 proposed rule.

III. Reorganization of Part 52 and Conforming Changes in the NRC's 
Regulations

    Since the NRC first adopted 10 CFR part 52 in 1989, the NRC and its 
external stakeholders have identified a number of interrelated issues 
and concerns. One significant concern is that the overall regulatory 
relationship between part 50 and part 52 is not always clear. It is 
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 1 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 part 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 does 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

[[Page 12784]]

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 current 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 the user experience 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 1 requirements 
following the receipt of public comments on the 2003 proposed rule, the 
NRC concludes that the 2003 proposed rule would not adequately address 
and resolve these issues.
    Accordingly, the NRC now proposes to take 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 1 to reflect the licensing and approval processes in part 52. 
The NRC has also attempted to coordinate and reconcile 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 proposed reorganization of part 52, the existing 
appendices O and M on standard design approvals and manufacturing 
licenses, respectively, would be redesignated as new subparts in part 
52. Redesignating these appendices as subparts in part 52 would result 
in a consistent format and organization of the requirements applicable 
to each of the licensing and approval processes. In addition, the 
redesignation would clarify 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 the proposed redesignation, 
either favor or disfavor the processes in the current appendices M and 
O. Rather, the Commission is simply attempting to standardize the 
format and organization of part 52, and to clarify 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 
proposes to retitle 10 CFR part 52 as ``Licenses, Certifications, and 
Approvals for Nuclear Power Plants.''
    The NRC also proposes to reorganize and expand 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 
Sec.  50.52), jurisdictional limits (analogous to Sec.  50.53), and 
attacks and destructive acts (analogous to Sec.  50.13). In general, 
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 10 CFR.
    Appendix N, which addresses duplicate design licenses, would be 
removed from part 52 and would be retained in part 50 because the 
duplicate design license is a part 50 operating license. Appendix Q, 
which addresses early staff review of site suitability issues, would 
also be 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 that there appears to be some redundancy 
between the early review of site suitability issues and the early site 
permit process. Accordingly, the NRC proposes to remove appendix Q from 
part 52 and retain it only in part 50.
    Inasmuch as the NRC may, in the future, adopt other regulatory 
processes for nuclear power plants, the NRC proposes to reserve 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 current appendices M and O. The 
standard format and content was modeled on the existing organization 
and content of subparts A and C.
    Perhaps most importantly, the NRC has reviewed the existing 
regulations in 10 CFR Chapter 1 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 NRC's proposed conforming 
changes 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 proposed conforming changes for each affected part 
follows.
    The NRC has prepared the following table that cross-references the 
proposed reorganized provisions of part 52 with the current 
requirements in part 52:

[[Page 12785]]



 Table 1.--Cross-References Between Proposed 10 CFR Part 52 and Existing
                              Requirements
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               Proposed rule                    Existing requirements
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                           General Provisions
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52.0......................................  52.1
52.1......................................  52.3
52.2......................................  52.5
52.3......................................  None
52.4......................................  52.9
52.5......................................  None
52.6......................................  None
52.7......................................  None
52.8......................................  None
52.9......................................  None
52.10.....................................  None
52.11.....................................  52.8
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                      Subpart A--Early Site Permits
------------------------------------------------------------------------
52.12.....................................  52.11
52.13.....................................  52.13
52.15.....................................  52.15
52.16.....................................  None
52.17.....................................  52.17
52.18.....................................  52.18
None......................................  52.19
52.21.....................................  52.21
52.23.....................................  52.23
52.24.....................................  52.24
52.25.....................................  52.25
52.27.....................................  52.27
52.28.....................................  None
52.29.....................................  52.29
52.31.....................................  52.31
52.33.....................................  52.33
52.35.....................................  52.35
None......................................  52.37
52.39.....................................  52.39
------------------------------------------------------------------------
                Subpart B--Standard Design Certifications
------------------------------------------------------------------------
52.41.....................................  52.41 and 52.45
52.43.....................................  52.43
52.45.....................................  52.45 and 52.49
52.46.....................................  None
52.47.....................................  52.47
52.48.....................................  52.48
52.51.....................................  52.51
52.53.....................................  52.53
52.54.....................................  52.54
52.55.....................................  52.55
52.57.....................................  52.57
52.59.....................................  52.59
52.61.....................................  52.61
52.63.....................................  52.63
------------------------------------------------------------------------
                      Subpart C--Combined Licenses
------------------------------------------------------------------------
52.71.....................................  52.71
52.73.....................................  52.73
52.75.....................................  52.75
52.77.....................................  52.77
None......................................  52.78
52.79/52.80...............................  52.79
52.81.....................................  52.81
None......................................  52.83
52.85.....................................  52.85
52.87.....................................  52.87
52.80.....................................  52.89
52.91.....................................  52.91
52.93.....................................  52.93
52.97.....................................  52.97
52.98.....................................  None
52.99.....................................  52.99
52.103....................................  52.103
52.104....................................  None
52.105....................................  None
52.107....................................  None
52.109....................................  None
52.110....................................  None
------------------------------------------------------------------------
                           Subpart D--Reserved
                  Subpart E--Standard Design Approvals
------------------------------------------------------------------------
52.131....................................  App. O, Introduction
52.133....................................  None
52.135(a).................................  App. O, Paragraph 1
52.135(b).................................  App. O, Paragraph 2
52.135(c).................................  None
52.136....................................  App. O, Paragraph 3
52.137....................................  App. O, Paragraph 3
52.139....................................  None
52.141....................................  App. O, Paragraph 4
52.143....................................  App. O, Paragraph 5
52.145(a).................................  App. O, Paragraph 5
52.145(b).................................  App. O, Paragraph 6
52.145(c).................................  App. O, Paragraph 7
52.147....................................  None
------------------------------------------------------------------------
                    Subpart F--Manufacturing Licenses
------------------------------------------------------------------------
52.151....................................  App. M, Introduction
52.153(a).................................  App. M, Paragraph 8
52.153(b).................................  N/A
52.155....................................  App. M, Paragraphs 2 and 4
52.156....................................  App. M, Paragraph 4
52.157....................................  App. M, Paragraphs 2, 4, 5,
                                             6
52.158....................................  App. M, Paragraph 3
52.159....................................  App. M, Paragraph 1
52.161 [Reserved].........................  N/A
52.163....................................  App. M, Paragraph 1
52.165....................................  App. M, Paragraph 1
52.167....................................  App. M, Paragraphs 5,6,8, 10
52.169 [Reserved].........................  N/A
52.171....................................  App. M, Paragraphs 11 and 12
52.173....................................  App. M, Paragraph 6
52.175....................................  None
52.177....................................  None
52.179....................................  None
52.181....................................  None
------------------------------------------------------------------------
                           Subpart G--Reserved
                         Subpart H--Enforcement
------------------------------------------------------------------------
52.301....................................  52.111
52.303....................................  52.113
------------------------------------------------------------------------

IV. Discussion of Substantive Changes

A. Introduction

    The proposed changes in 10 CFR Chapter I are further discussed by 
part. Proposed changes to parts 52 and 50 are discussed first followed 
by proposed changes to other parts in numerical order. Within each 
part, general topics are discussed first, followed by discussion of 
proposed changes to individual sections as necessary. In addition to 
the substantive changes, existing rule language was revised to make 
conforming administrative changes (e.g., identification of regulations 
containing information collection requirements in Sec.  52.10), 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 proposed rule would amend Sec. Sec.  50.43, 52.47(b) (proposed 
Sec.  52.47(c)), 52.79, and appendix M to part 52 (proposed Sec.  
52.157) to achieve consistency in the requirements for testing advanced 
reactor designs and plants. This amendment would require applicants for 
a combined license, operating license, or manufacturing license that do 
not reference a certified advanced reactor design to also perform the 
design qualification testing required of applicants for design 
certification under the current Sec.  52.47(b)(2). If a combined 
license application references a certified design, the qualification 
testing required by the current Sec.  52.47(b)(2) will have been 
performed. The codification of testing requirements in Sec.  
52.47(b)(2) was a principal issue during the original development of 10 
CFR part 52 (see Section II of 54 FR 15372; April 18, 1989). The 
requirements in Sec.  52.47(b)(2), which demonstrate the performance of 
new safety features for nuclear power plants that differ significantly 
from evolutionary light-water reactors or 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, that the effects of systems interactions are 
acceptable, and to provide sufficient data to validate analytical 
codes. 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 (see Policy Statement at 51 FR 24643; 
July 8, 1986) and the Commission's goal of resolving all safety issues 
before authorizing construction.

[[Page 12786]]

    During the development of 10 CFR part 52, the focus of the nuclear 
industry and the NRC was on applications for design certification. That 
is why the testing requirements to qualify new or innovative safety 
features was only included in subpart B of part 52. Furthermore, the 
tests to qualify a new safety feature are different than verification 
tests, which are required by the current Sec.  52.79(c) and performed 
in accordance with Section XI, ``Test Control,'' of appendix B to part 
50. Verification tests are used to provide assurance that construction 
and installation of equipment (as-built) in the facility has been 
accomplished in accordance with the approved design.
    This amendment also proposes, in Sec. Sec.  50.43(e)(2) and 
52.79(a), a requirement for licensing a prototype plant, as defined in 
proposed Sec. Sec.  50.2 and 52.1, if it is used to meet the 
qualification testing requirements in proposed Sec.  50.43(e). New 
Sec.  50.43(e) states that, if a prototype plant is used to comply with 
the 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. Although the NRC 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 
Policy Statement at 51 FR 24646; July 8, 1986, and Section II of the 
final rule (54 FR 15372; April 18, 1989) on 10 CFR part 52), this 
revised proposed rule would not require the use of a prototype plant 
for qualification testing. Rather, this proposed rule would provide 
that if a prototype plant is used to qualify an advanced reactor 
design, then additional requirements may be required for licensing the 
prototype plant to compensate for any uncertainties with the unproven 
safety features. Also, the prototype plant could be used for commercial 
operation. Finally, it would be inconsistent for the NRC to require 
qualification testing only for design certification applications (paper 
designs) and not require testing for applications to build and operate 
an actual nuclear power plant. Therefore, the NRC proposes to amend the 
current Sec. Sec.  50.43, 52.47(b), 52.79, and appendix M to part 52 to 
implement its intent in adopting part 52 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 reactors 
manufactured under a manufacturing license).

C. Proposed 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, 
52.24, 52.39, 52.47, 52.54, 52.79, 52.93, 52.157, 52.158, 52.167, 
52.171, and Appendices A, B, and C
    The NRC believes that 10 CFR part 52 should be modified to clarify 
the use of the terms, site characteristics, site parameters, design 
characteristics, and design parameters, to present the NRC's 
requirements governing applications for and issuance of early site 
permits, design approvals, design certifications, combined licenses, 
and manufacturing licenses in clear and unambiguous terms. The proposed 
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 proposing to add 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 or 
reactors. Design characteristics are specified in a standard design 
approval, a standard design certification, or a combined license 
application. 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 has revised 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. Section 52.79 already 
contains 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 in this case to mean 
the site characteristics and design parameters as defined in proposed 
Sec.  52.1. The NRC proposes 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. Commission 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 reasons, the NRC believes it is important to 
clarify the use of these terms and their applicability to the part 52 
licensing processes.
2. Issuance of Combined and Manufacturing Licenses (Sec. Sec.  52.97 
and 52.163)
    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

[[Page 12787]]

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 
proposes to add the language in Sec. Sec.  52.97(d) for combined 
licenses, and 52.163 for manufacturing licenses, which are analogous to 
Sec.  50.50.
3. General Provisions
    a. Section 52.0, Scope; applicability of 10 CFR Chapter 1 
provisions. The NRC proposes to redesignate current Sec.  52.1, Scope, 
as Sec.  52.0, Scope; applicability of 10 CFR Chapter 1 provisions. In 
proposed Sec.  52.0, paragraph (a) consists of current Sec.  52.1 on 
the scope of part 52, and paragraph (b) addresses the applicability of 
10 CFR Chapter 1 provisions. Currently Sec.  52.1 states that part 52 
governs the issuance of early site permits, standard design 
certifications, and combined licenses for nuclear power facilities 
licensed under Section 103 or 104b of the Atomic Energy Act of 1954 
(AEA), as amended (68 Stat. 919), and Title II of the Energy 
Reorganization Act of 1974 (88 Stat. 1242). In proposed Sec.  52.0(a), 
the NRC proposes to revise this provision to include standard design 
approvals and manufacturing licenses within the scope of part 52 and to 
restrict licenses issued under part 52 to those 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.\1\ Accordingly, the NRC proposes to revise the scope of 
part 52 to limit its applicability to licenses issued under Section 103 
of the AEA.
---------------------------------------------------------------------------

    \1\ This may be an academic distinction, in light of the Energy 
Policy Act of 2005, Pub. L. 109-58, which removed the need for 
antitrust reviews of new utilization facilities.
---------------------------------------------------------------------------

    The addition of proposed Sec.  52.0(b) stems from the July 3, 2003 
(68 FR 40026) proposed rule. In that proposed rule, the NRC 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 proposed change was in response to a letter dated November 13, 
2001, from the Nuclear Energy Institute (NEI) that 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.

    The purpose of the amendment proposed in 2003 was to clarify that 
these 10 CFR part 50 provisions are applicable to the licensing 
processes that were formerly in 10 CFR part 50 (appendices M, N, O, and 
Q) and are now in 10 CFR part 52, as well as to the new licensing 
processes for early site permits, standard design certifications, and 
combined licenses. Although these provisions in 10 CFR part 50 did not 
refer to the additional licensing processes in 10 CFR part 52, the new 
Sec.  52.5 was proposed to make it clear that a holder of or applicant 
for an approval, certification, permit, or license issued under 10 CFR 
part 52 must comply with all requirements in these provisions that are 
otherwise applicable to applicants or licensees under 10 CFR part 50. 
In preparing the revised proposed rule, the NRC has taken into account 
the comments it received on the 2003 proposed rule which indicated that 
the previous change to add Sec.  52.5 was overly broad and would impose 
burdensome and seemingly inappropriate new requirements on applicants 
for design certifications that were not warranted for entities that 
were neither constructing nor operating a reactor.
    The NRC agrees that the amendment proposed in 2003 was not 
sufficiently detailed to make it clear which of the part 50 provisions 
applied to each of the part 52 licensing processes. The NRC has 
concluded that the most effective solution to this problem is to make 
conforming changes to all of the regulations in 10 CFR Chapter 1 that 
are applicable to the part 52 licensing processes. Accordingly, the NRC 
has reviewed all of 10 CFR Chapter 1 to identify requirements that 
apply to one or more of the licensing processes in 10 CFR part 52 and 
is proposing conforming changes to those requirements. As a result of 
this effort, the NRC proposes to add new Sec.  52.0(b) which makes it 
clear that the regulations in 10 CFR Chapter 1 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.
    b. Section 52.1, Definitions. The NRC proposes to amend Sec.  52.1 
by adding the definitions for decommission, license, licensee, 
manufacturing license, modular design, prototype plant, and standard 
design approval. The definition of decommission from 10 CFR part 50 
would be added to 10 CFR part 52 because the NRC is proposing that part 
52 address decommissioning of nuclear power facilities with combined 
licenses. The definitions of license and licensee are consistent with 
the definitions of the same terms that the NRC is proposing in 10 CFR 
parts 2 and 50. Definitions of manufacturing license and standard 
design approval would be added so that each of the part 52 license 
types are defined in this section.
    The definition of modular design would be added to explain the type 
of modular reactor design to which the NRC intended to refer to in the 
second sentence of the current Sec.  52.103(g). This special provision 
for modular designs would be 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 3 or 4 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. The NRC 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 definitions for modular design that may be used within the 
nuclear industry.
    The NRC proposes to add a definition for prototype plant to explain 
the type of nuclear power plant that the NRC intended in the current 
Sec.  52.47(b), and in the proposed Sec. Sec.  50.43, 52.47, 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

[[Page 12788]]

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 Section 52.4, Deliberate 
misconduct. The current section on interpretations in Sec.  52.5 is 
retained and redesignated as Sec.  52.2 and the current section on 
deliberate misconduct in Sec.  52.9 is retained and redesignated as 
Sec.  52.4.
    d. Section 52.3, Written communications; Section 52.5, Employee 
protection; Section 52.6, Completeness and accuracy of information; 
Section 52.7, Specific exemptions; Section 52.8, Combining licenses; 
Section 52.9, Jurisdictional limits; and Section 52.10, Attacks and 
destructive acts. The NRC proposes to clarify the regulatory structure 
of part 52 by proposing to add new Sec. Sec.  52.3, Written 
communications; 52.5, Employee protection; 52.6, Completeness and 
accuracy of information; 52.7, Specific exemptions; 52.8, Combining 
licenses; 52.9, Jurisdictional limits; and 52.10, Attacks and 
destructive acts. The Commission proposes to add Sec.  52.3, Written 
communications, which is essentially identical with the current Sec.  
50.4, 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.
    The Commission proposes to add Sec.  52.5, to address 
discrimination against an employee for engaging in certain protected 
activities concerning the regulations in part 52. Accordingly, the 
Commission proposes to add Sec.  52.5, which is essentially identical 
with the current Sec.  50.7, with the exception of the addition of a 
provision on coordination with the requirements in 10 CFR part 19.
    The Commission proposes to add Sec.  52.6, which is identical with 
the current Sec.  50.9, 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 proposes to add Sec.  52.7, which is essentially 
identical with current Sec.  50.12, 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, for example the 
provisions 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 with regard 
to design certification information must meet the criteria in the 
specific design certification rule or Sec.  52.63(b)(1), as applicable. 
If the applicant or licensee seeks an exemption from other provisions 
of Subpart B or other provisions of a particular standard design 
certification rule, then it may request an exemption under the more 
encompassing authority of Sec.  52.7. The exemption request must then 
demonstrate compliance with the additional criteria in Sec.  52.7.
    The NRC proposes to add Sec.  52.8, which is essentially identical 
with the current Sec.  50.31, 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. Although Sec.  50.31 contains a 
provision allowing a part 50 license, such as an operating license, to 
be combined with a part 52 license, such as an early site permit, Sec.  
50.31 does not address the Commission's authority to combine a part 52 
license with a non-part 50 license.
    The Commission proposes to add Sec.  52.9, which is identical with 
Sec.  50.53, 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.
    The Commission proposes to add Sec.  52.10 because there is no 
specific provision in part 52 that applies to part 52 processes 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. That 
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 facilities in Sec.  
50.13. Although it would be possible to revise Sec.  50.13 so that its 
provisions apply to part 52 licenses, early site permits, standard 
design certifications, and standard design approvals, this is 
inconsistent with the overall regulatory pattern of 10 CFR, whereby 
each part is treated as a separate and independent regulatory unit. 
Moreover, any changes to Sec.  50.13 may erroneously be viewed as 
changes to the Commission's substantive determination on this matter.
    For these reasons, the Commission is proposing to add Sec.  52.10, 
which is essentially identical with Sec.  50.13. Inclusion of this 
provision in part 52 would make clear that combined licenses, 
manufacturing licenses, design certification rulemakings, standard 
design approvals, and amendments to these licenses, rulemakings, and 
approvals under part 52--as with licenses issued under part 50--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 United States 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 United States 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.
4. Subpart A, Early Site Permits
    a. Emergency Preparedness Requirements for Early Site Permit 
Applicants. The NRC proposes to amend Sec. Sec.  52.17(b), 52.18, and 
52.39 to address changes to emergency preparedness requirements for 
early site permit applicants. The NRC proposes to

[[Page 12789]]

amend 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 proposes to add 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 proposes to amend 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 proposes to amend 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 would clarify 
the standards applicable to emergency preparedness information supplied 
with an early site permit application. In addition, the Commission 
proposes to add 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 proposing these amendments for consistency with the requirements in 
subpart C of part 52 regarding the review of emergency plans at the 
early site permit stage. 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 not be able to make the required finding 
without the inclusion of proposed inspections, tests, analyses, and 
acceptance criteria in an early site permit application that includes 
complete and integrated emergency plans.
    b. Section 52.13, Relationship to other subparts. The NRC proposes 
to retitle Sec.  52.13 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. The reference to Appendix Q and part 2 are no longer 
needed, consistent with the Commission's decision (discussed earlier in 
section II) to remove Appendix Q from part 52.
    c. Section 52.16, Contents of applications; general information and 
Section 52.17, Contents of applications; technical information. The NRC 
proposes to add Sec.  52.16 to include the general content requirements 
from Sec.  52.17(a)(1).
    The title of Sec.  52.17 would be revised to read, ``Contents of 
applications; technical information,'' Section 52.17(a)(1) would be 
amended to state that the early site permit application should 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, which is consistent with the language in paragraph 2 of 
current appendix Q to part 52, 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 early site permit 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-viii), 
industry recommends a 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 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 proposes to amend Sec.  52.17(a)(1) to eliminate all 
references to Sec.  50.34. The references to Sec.  50.34(a)(12) and 
(b)(10) would be 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, would also be removed. All requirements related to 
emergency planning for early site permits are addressed in Sec.  
52.17(b). Finally, the reference to the radiological consequence 
evaluation factors identified in Sec.  50.34(a)(1) would be removed and 
restated in Sec.  52.17(a)(1). The NRC is proposing to modify 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 proposed 
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

[[Page 12790]]

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 proposes to add several requirements to Sec.  52.17(a)(1). 
A requirement would be added to Sec.  52.17(a)(1)(xi) 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)(xii) would 
be 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 proposed change was made for 
consistency with proposed 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.''
    Two additional requirements would be added Sec.  52.17(a)(1) that 
are taken from Sec.  50.34(b), and which the NRC believes are 
applicable to early site permit applicants. Section 52.17(a)(1)(xii) 
would require applicants proposing to site nuclear power plants on 
sites which already have on them one or more licensed units to include 
in its application an evaluation of the potential hazards of 
construction activities to the structures, systems, and components 
important to safety of operating units, as well as a description of the 
managerial and administrative controls to be used to provide assurance 
that the limiting conditions for operation of the existing units are 
not exceeded as a result of construction activities. This requirement 
currently exists for applicants for construction permits, operating 
licenses, and combined licenses. The NRC believes it should also be 
applicable to applicants for early site permits so that all applicable 
issues are included in the NRC's review of site suitability before a 
decision is made on issuance of an early site permit, including issues 
that affect units already operating on the site (if this matter is 
addressed and resolved in an early site permit, this matter would have 
finality and need not be addressed in a referencing combined license 
proceeding). Section 52.17(a)(1)(xiii) would require that early site 
permit applications include an evaluation of the site against the 
applicable sections of the Standard Review Plan revision in effect 6 
months before the docket date of the application. This requirement 
currently exists for applicants for construction permits, operating 
licenses, design certifications, design approvals, combined licenses, 
and manufacturing licenses. The NRC believes it should also 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.
    The NRC would amend 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 current 
regulations already afford the early site permit applicant such 
flexibility, inasmuch as Sec.  52.17(a)(2) states 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 Sec. 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 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 alternative energy 
sources).

Id. at 48 (citations omitted). Accordingly, the NRC believes that Sec.  
52.17(a)(2) already provides 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 proposed 
rule would 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 proposes a 
conforming amendment to Sec. Sec.  52.18 and 52.21 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, under 
Sec.  52.17(a)(2), 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 under 
Sec. Sec.  52.17(a)(2), 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 under 
Sec.  52.17(a)(2) 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

[[Page 12791]]

Sec.  52.17(a)(2).\2\ With this clarification, the NRC does not believe 
that further changes to the language of Sec. Sec.  52.17 and 52.18 are 
necessary.
---------------------------------------------------------------------------

    \2\ The NRC emphasizes that under Sec.  52.17(a)(2), only the 
discussion of benefits (including need for power) of constructing 
and operating a nuclear power reactor (or reactors), and the 
discussion of alternative energy sources, may be deferred. The 
environmental report must always address the ``environmental impacts 
of construction and operation of a reactor, or reactors, which have 
characteristics which fall within the postulated site parameters.''
---------------------------------------------------------------------------

    The NRC would amend 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 would 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.  
52.24(c), those additional activities would have to be requested 
separately under Sec.  50.10(e)(1).
    d. Section 52.24, Issuance of early site permit. The Commission 
proposes to amend Sec.  52.24 to clarify the information that the NRC 
must include in the early site permit when it is issued. Section 52.24 
would also be 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 would be 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 is currently provided. The 
change would provide consistency with Sec.  52.39(a)(2), and in 
particular paragraph (a)(2)(iii) of the current regulations, which also 
refers to ``site parameters'' (corrected to ``site characteristics'' in 
the proposed rule) and ``terms and conditions.'' Section 52.24(c) would 
be added to require that the early site permit state the activities 
that the permit holder is authorized to perform at the site. This 
change would be 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).
    e. Section 52.28, Transfer of early site permit. Section 52.28 
would be 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 
Nuclear Energy Institute 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.
    f. Section 52.37, Reporting of defects and noncompliance; 
revocation, suspension, modification of permits for cause. Section 
52.37 would be removed because this provision only contains a cross-
reference to 10 CFR part 21 and Sec.  50.100, and the NRC is proposing 
conforming changes to those requirements to account for requirements 
for early site permits.
    g. Section 52.39, Finality of early site determinations; and 
Section 52.93, Exemptions and variances. Section 52.39 would be revised 
to address the finality of an early site permit. While some of the 
proposed changes are conforming or clarifying, some proposed changes 
represent a change from the finality provisions in the current Sec.  
52.39. Paragraph (a)(2) of the current rule distinguishes among issues 
alleging that: (i) 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)); (ii) 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 (iii) 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 any question about whether the referencing 
application complies with the early site permit should be regarded as a 
question material to the proceeding and admissible 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).
    The NRC also regards new emergency preparedness information 
submitted in the referencing application which materially changes the 
Commission's determination on emergency preparedness matters as an 
issue material to the proceeding and admissible as a contention in the 
referencing application proceeding. Any significant environmental issue 
material to the combined license application which was not considered 
in the early site permit proceeding is also subject to litigation 
during the proceeding on the referencing application to the extent the 
issue differs from issues discussed or reflects significant new 
information. 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

[[Page 12792]]

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 committed at the time of issuance (i.e., 
Commission failure to consider relevant information known and available 
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 proposed revisions to Sec.  52.39 are in agreement with these 
Commission conclusions. Section 52.39 would be 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 proposed rule language is based upon the existing 
regulation, but adds an additional circumstance. Section 
52.39(a)(1)(iii) would provide 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) would provide that the NRC may modify the early site 
permit if a variance is issued under proposed Sec.  52.39(d) (paragraph 
(b) in the current regulations); the NRC considers this a conforming 
change inasmuch as the current regulation provides for issuance of 
variances.
    The NRC proposes to clarify 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 current term, ``requirements.'' Under the proposed 
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 proposed 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,'' 
would be used consistently throughout Sec.  52.39 and corresponding 
provisions in the proposed revision to Sec.  52.79.
    Section 52.39(a)(2) would describe how the NRC would treat matters 
resolved in the early site permit proceeding in subsequent proceedings 
on applications referencing the early site permit, and is drawn from 
the current language of Sec.  52.39(a)(2). In addition, under the last 
sentence of proposed Sec.  52.39(a)(2), the NRC would finalize changes 
to an early site permit's emergency plan (or major features of it, as 
contemplated under Sec.  52.17(b)(2)) that are made after the issuance 
of the early site permit, but only if (1) the approved early site 
permit's emergency plan (or major feature) is based upon an emergency 
plan in use by a licensee of a nuclear power plant; (2) the changes to 
the early site permit emergency plan are identical to the changes in 
the referenced licensee's plan; and (3) the changes in the referenced 
licensee emergency plan are in compliance with Sec.  50.54(q). The 
Commission's proposal 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 should not require NRC 
review and approval as part of the licensing process. Any issues with 
compliance with Sec.  50.54(q) should be treated as an enforcement 
matter.
    Section 52.39(b) is discussed separately under Section IV.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) would replace the current criteria in Sec. Sec.  
52.39(a)(2)(i) through (iii), governing how the NRC would treat various 
issues with respect to the early site permits and its referencing in a 
combined license application. Matters regarding compliance with the 
early site permit which would be potentially valid subjects of 
contention under the proposed rule 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. 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, 
would be treated under proposed Sec.  52.39(c)(2) as challenges to the 
permit and processed in accordance with Sec.  2.206. The proposed rule 
would retain the current 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 current 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 proposed rule would redesignate the current provision in Sec.  
52.39(b) allowing an applicant for a license referencing an early site 
permit to request a variance from one of more ``elements'' of the early 
site permit as Sec.  52.39(d). The proposed rule would clarify 
``elements'' for which a variance may be sought by substituting the 
phrase, ``site characteristic, design parameter, term, or condition.'' 
The Commission notes that the admission of a contention on a proposed 
variance, which is currently addressed in Sec.  52.39(b), would now be 
addressed in Sec.  52.39(c)(iii) of the proposed rule. Finally, the 
proposed rule would preclude the Commission 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 combined license.
    Finally, the Commission proposes to add a new paragraph (e) to the 
``finality'' section in each subpart of part 52,

[[Page 12793]]

including Sec.  52.39, entitled ``Information requests,'' which would 
delineate the restrictions on the NRC for information requests to the 
holder of the early site permit. This provision is analogous to the 
current 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 proposed provision would be contained in 
Sec.  52.39(d), and would require 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.
5. 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 would be moved 
from the current Sec.  52.45(a) to this section.
    b. Section 52.43, Relationship to other subparts. This section 
defines the relationship of subpart B to other subparts in 10 CFR part 
52. The proposed rule would remove the requirements currently located 
in Sec. Sec.  52.43(c), 52.45(c), and 52.47(b)(2)(ii) because the 
Commission has decided not to require a final design approval (FDA) as 
a prerequisite for certification of a standard plant design under 
subpart B. This requirement was included in 10 CFR part 52 because, at 
the time of the original rulemaking, the NRC had no experience with 
design certification applications. 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 current 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 
would be formatted for consistency with the other subparts in 10 CFR 
part 52 and would replace the references to specific paragraphs within 
Sec. Sec.  50.4 and 50.30 with references to subpart H of part 2. 
Specific references are no longer needed because the NRC proposes 
conforming changes to Sec. Sec.  50.4 and 50.30 that clarify which 
provisions are applicable to combined license applications. A new Sec.  
52.45(c) on design certification review fees, which are currently set 
forth in Sec.  52.49, is included.
    d. Section 52.46, Contents of applications; general information. A 
new section would be added containing the appropriate general content 
requirements from 10 CFR 50.33 as a conforming amendment.
    e. Section 52.47, Contents of applications; technical information. 
This section presents the requirements for contents of a design 
certification application. Section 52.47 would be reorganized into 
separate provisions. The requirements for the final safety analysis 
report (FSAR) are proposed in Sec. Sec.  52.47(a) and 52.47(c), and the 
technical requirements for the remainder of the design certification 
application are proposed in Sec.  52.47(b). The current Sec.  
52.47(a)(1)(i) requires the submittal of information required of 
applicants for construction permits and operating licenses by parts 20, 
50 (including the applicable requirements from 10 CFR 50.34), 73, and 
100, and which is technically relevant to the design and not site-
specific. That requirement would be removed and replaced with the 
relevant requirements from the regulations that describe what must be 
included in an FSAR. In addition, the Commission proposes to codify 
technical positions that were developed after part 52 was adopted by 
the Commission in 1989, such as the proposed requirement in Sec.  
52.47(a)(19) requiring an explanation how relevant operating experience 
was incorporated into the standard design (see SRM on SECY-90-377, 
dated February 15, 1991, ML003707892). Also, the technical requirements 
in the regulations that are relevant would be revised to clearly state 
their applicability to design certifications. In doing so, the NRC has 
attempted to capture all relevant requirements regarding contents of 
the FSAR for a design certification application.
    A new Sec.  52.47(b) would be added to cover the required technical 
contents of a design certification application that are not contained 
in the FSAR. The proposed rule would conform the requirement for 
acceptable inspections, tests, analyses, and acceptance criteria 
(ITAAC) (proposed Sec.  52.47(b)(2)) with the AEA and the requirements 
in the current Sec.  52.97(b). This clarification of the current 
language, which was a condensed version of the language in Sec. Sec.  
52.79(c) and 52.97(b), is intended to avoid any future 
misunderstandings.
    The current Sec.  52.47(b) (proposed Sec.  52.47(c)) would be 
reorganized by separating the requirements on scope of design and 
modular configuration from the testing requirements. This is part of 
the NRC's goal to set forth the procedural requirements for the 
licensing processes in part 52 and the reactor safety requirements in 
part 50. As a result, the testing requirements would be relocated to 
Sec.  50.43(e), and the requirements on scope of design and modular 
configuration would remain in the proposed Sec.  52.47(c). Also, see 
the discussion on testing requirements for advanced nuclear reactors in 
Section B.1 of this document.
    f. Section 52.54, Issuance of standard design certification. 
Section 52.54 would be 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 Commission determines that there is 
reasonable assurance that the design conforms with the provisions of 
the AEA, and the Commission'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) 
would be added to indicate 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 is being added to indicate the NRC's 
expectation that design certification applicants implement the QA 
program that is required to be included in their application under 
Sec.  52.47(a)(21). The NRC is also considering whether a parallel 
requirement should be added to Part 50 (e.g., in a new Sec.  50.54a), 
similar to the requirements for QA program implementation contained in 
proposed Sec. Sec.  50.54(a) and 50.55(f). A new Sec.  52.54(b) would 
be added, consistent with Sec.  50.50, which states that a design 
certification shall specify the site parameters and design 
characteristics and any additional requirements and restrictions of the 
rule, as the Commission deems necessary and appropriate.
    The Commission is proposing to modify Sec.  52.54 to require that 
applicants

[[Page 12794]]

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 are met. Section 52.54 would be amended to include a 
new paragraph (c) which requires that every standard design 
certification rule 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 proposed 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 
proposed rule would amend the special backfit requirement in Sec.  
52.63(a)(1) to provide the Commission with the ability to make changes 
to the design certification rules or the certification information in 
the generic design control documents that reduce unnecessary regulatory 
burdens. Section 52.63(a)(1) currently states 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. The regulation does 
not appear to permit changes to the certification 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 may not be able to 
be made under the current Sec.  52.63(a)(1) is a proposed change to the 
three design certification rules 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).
    Section 52.63(a)(1) would include a new provision that explicitly 
allows the Commission to change the design certification rules in part 
52 to make future changes to reduce unnecessary regulatory burden, 
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, e.g., 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 Regulatory Guide 1.174). Changes to the 
design certification rules must be accomplished through rulemaking, 
with opportunity for public comment. Once a design certification rule 
is changed through rulemaking, under proposed Sec.  52.63(a)(2), the 
provisions would apply to all applications referencing the design 
certification rule as well as all current plants referencing the design 
certification, unless the change has been rendered ``technically 
irrelevant'' through other action taken under Sec. Sec.  52.63(a)(3) or 
(b)(1). Thus, standardization is maintained by ensuring that any 
changes to a design certification rule intended to reduce regulatory 
burden are imposed upon all nuclear power plants referencing the design 
certification rule.
    Section 52.63(a)(1) would be modified to replace ``a modification'' 
with ``the change,'' to clarify that the three criteria for changes 
apply to modifications, rescissions, or imposition of new requirements. 
Also, proposed Sec.  52.63 is amended to be consistent with its 
original intent (refer to 54 FR 15372; April 18, 1989) that the special 
backfit requirements apply to the certification information in the 
generic design control documents, not to the provisions in the design 
certification rules, e.g., Section VI.E of appendix A to part 52. Any 
proposed changes to these provisions that set forth how the design 
certification regulations are to be used are controlled by the normal 
backfit requirements in 10 CFR 50.109.
    The proposed rule would amend the current Sec.  52.63(a)(2) to 
delete the reference to Sec.  52.63(a)(4). The reference to 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 Sec.  52.63(a)(2) deals with modifications 
that the NRC may impose on a design certification rule under Sec. Sec.  
52.63(a)(3) or 52.63(b)(1). No substantive change is intended by the 
amendment which merely clarifies the original intent of the rule.
6. Subpart C, Combined Licenses
    a. Emergency Preparedness Requirements for a Combined License 
Applicant Referencing an Early Site Permit. The Commission proposes to 
modify current 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 
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 Commission agrees in part with the Illinois Department of 
Nuclear

[[Page 12795]]

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 Commission'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 Commission 
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) would be 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 proposed 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) would also be 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: 
(1) 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 (2) 
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 Commission'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) would 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 would 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 would 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 52.79(c), 52.85, 52.97(a), 52.99, 
and 52.103(a) and (g) would be amended to provide an applicant for a 
combined license with a process for resolving certain acceptance 
criteria in one or more of the inspection, test, analysis, and 
acceptance criteria (ITAAC) required by the proposed Sec.  52.79(c) 
before issuance of the combined license. In a letter dated November 13, 
2001 (comment 20 on draft proposed rule text), NEI recommended that 
subpart C be revised to allow for completion of design acceptance 
criteria (DAC) at the combined license application stage. NEI made this 
recommendation 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 detailed 
design information, such as information about the digital 
instrumentation and control system. DAC were originally written to be 
verified as part of the normal, post-combined license, ITAAC 
verification process; as such, DAC are in essence specialized ITAAC.
    The Commission agrees with NEI's recommendation that combined 
license applicants be permitted to demonstrate DAC completion as part 
of the combined license application, for several reasons. First, 
completion of the detailed 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 10 CFR part 52 was adopted, 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 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. Finally, the Commission believes that completion of DAC 
before issuance of the combined license will enhance public confidence 
in the overall licensing process because the public will have an 
opportunity to challenge whether the detailed design has been properly 
completed before construction begins. Accordingly, the Commission 
proposes that 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 new 
process would also allow findings on successful completion of 
inspections or tests of components procured before the issuance of the 
combined license. These matters would not be revisited after issuance 
of the combined license.
    Section 52.79(c) would be amended to provide a new provision that 
states that, if the application references an early site permit or a 
certified design, the application may include a notification that a 
required inspection, test, or analysis in the ITAAC has been 
successfully completed and that the

[[Page 12796]]

corresponding acceptance criterion has been met. Sections 52.79(c) and 
52.85 would be amended to require that the Federal Register 
notification required by Sec.  52.85 indicate that the application 
includes this notification, thereby ensuring that the public has 
adequate notice of the scope and nature of the application which the 
Commission is being requested to review.
    Sections 52.99 and 52.103 would be amended 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 Commission 
codified these ITAAC requirements into Section IX of the regulations. 
The purpose of the requirement in proposed 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. Proposed 
Sec.  52.99(c) would require the licensee to notify the NRC that the 
required inspections, tests, and analyses in the ITAAC have been 
completed and that the acceptance criteria have been met. For those 
inspections, tests, or analyses that are completed within 180 days 
before the scheduled date for initial loading of fuel, Sec.  52.99(c) 
would require that the licensee notify the NRC within 10 days of the 
successful completion of ITAAC. This immediate notification is 
necessary to ensure the NRC has sufficient time to verify successful 
completion of the ITAAC prior to the licensee's scheduled date for fuel 
load. Section 52.99(d) would state 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. Section 52.99(e) requires the 
NRC to ensure that the required inspections, tests, and analyses in the 
ITAAC are performed and also requires the NRC to publish, at 
appropriate intervals, notice in the Federal Register of the NRC 
staff's determination of the successful completion of inspections, 
tests, and analyses. 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 current 
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 (item 23), NEI requested that the 
NRC ``consider incorporating DCR [Design Certification Rule] general 
provisions into Subpart C as appropriate.'' The NRC has decided to add 
these ITAAC requirements to proposed 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.
    c. Section 52.73, Relationship to other subparts. Section 52.73 
would clarify that a design approval issued under proposed subpart E or 
a site report issued under proposed subpart B of part 52 may also be 
referenced in an application for a combined license application filed 
under 10 CFR part 52. This amendment would also add the requirements in 
the current Sec.  52.63(c) to the new Sec.  52.73(b) 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 (items 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 current Sec.  52.79 rather than the current Sec.  52.73. 
The NRC agrees with NEI that 10 CFR part 52 should be modified to 
clarify that the requirement in current Sec.  52.63(c) applies to 
applicants for a combined license, and that the last sentence be 
deleted. However, the Commission is adding the remaining provision to 
what was 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 
proposes to reformat 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 proposes conforming changes to Sec. Sec.  50.4 and 50.30 that 
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 would be 
deleted, and the requirements applicable to an applicant for, and 
holder of, a combined license with respect to the training program 
would be relocated 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 Section 52.80, Contents of 
application; additional technical information. Section 52.79 would be 
reformatted to divide the requirements for the technical contents of a 
combined license application into two separate provisions. Section 
52.79 would cover requirements for the contents of the FSAR, and Sec.  
52.80 would cover requirements for the remainder of the technical 
content of a combined license application.
    Current 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 
would be 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) 
would be 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 proposing these 
modifications, the NRC has attempted to capture all relevant 
requirements regarding contents of the FSAR for a combined license 
application.
    In addition, the proposed 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 
proposed amendment is in support of the

[[Page 12797]]

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 Commission proposes to add requirements for 
descriptions of operational programs. In doing so, the Commission has 
taken into account NEI's proposal in its letter dated August 31, 2005, 
to address SRM-SECY-04-0032.
    Section 52.79(b) would describe the variant on the requirements in 
Sec.  52.79(a) for a combined license application that references an 
early site permit. Section 52.79(a) does 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) would require 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's intent, as reflected in the words, ``have 
been met,'' is that all terms and conditions will be met before 
issuance of the combined license.
    Section 52.79(c) would describe 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 proposed 
requirements in Sec.  52.79(c) are essentially the same as those for a 
combined license application that references a standard design 
certification in proposed Sec.  52.79(d).
    Section 52.79(d) would describe the requirements for combined 
license applications that reference a standard design certification. 
Section 52.79(d) would state that the FSAR for a combined license 
application referencing a standard design certification need not 
contain information or analyses submitted to the Commission in 
connection with the design certification, but 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. 
Section 52.79(d) would require 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) would describe 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 Commission in 
connection with the manufacturing license, but must contain, in 
addition to the information and analyses otherwise required, 
information sufficient to demonstrate that the site parameters for the 
manufactured reactor are bounded by the site where the manufactured 
reactor is to be installed and used. Section 52.79(e) also would 
require 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 52.79 would require that emergency plans submitted with a 
combined license application be included in the FSAR (proposed Sec.  
52.79(a)). This modification is proposed for consistency with current 
Sec.  50.34 which requires that emergency plans be included in the FSAR 
for operating license applications.
    Section 52.80 would be added to cover the required technical 
contents of a combined license application that are not contained in 
the FSAR. These application contents include the PRA, ITAAC, and the 
environmental report.
    The NRC proposes to add a requirement in Sec.  52.80(a) that an 
applicant submit a plant-specific PRA as part of an application for a 
combined license. The current Sec.  52.79(b) references Sec.  
52.47(a)(1)(v), which requires a design-specific PRA within a design 
certification application. This amendment would add new Sec.  52.80(a) 
to require that if an application for a combined license references a 
standard design certification or standard design approval, or if the 
application proposes to use a nuclear power reactor manufactured under 
a manufacturing license under subpart F of this part, the plant-
specific PRA must use the PRA for the design certification, design 
approval, or manufactured reactor, as applicable, and must be updated 
to account for site-specific design information and any design changes, 
departures, or variances. In a letter dated April 3, 2001 (item 11.1a), 
NEI stated ``we agree on the NRC vision for a plant-specific PRA at COL 
that supplements the DC PRA with any changes that affect the DC PRA 
plus site-specific (interface) design information.'' A requirement 
would be added to Sec.  52.80(a) that a combined license application 
that does not reference a certified design must contain a plant-
specific PRA.
    The purpose of the requirement for a plant-specific PRA is to 
identify and address potential design and operational vulnerabilities; 
gain insights about the risk of the design; assess the balance between 
preventive and mitigative features in the design; determine 
quantitatively whether the design represents a reduction in risk over 
current operating plants; and, determine how the risk associated with 
the new design relates to the Commission's safety goals.
    g. Section 52.81, Standards for review of applications. 10 CFR 
parts 54 and 140 would be added to the list of standards that the NRC 
will use to review combined license applications. Part 54 would address 
applications for renewal of combined licenses and part 140 would 
include the requirements applicable to nuclear reactor licensees

[[Page 12798]]

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. The current 
Sec.  52.83, Applicability of part 50 provisions, would be removed and 
would be replaced by a new section addressing the finality of NRC 
approvals which are referenced in a combined license application. 
Current 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 
Commission believes that the current 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 1) to identify which requirements are applicable to combined 
license applicants and holders. Current 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 proposed revision to Sec.  52.83 would state 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 would clarify 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 would be 
removed and reserved for future use. Current Sec.  52.89 requires 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. Current Sec.  52.89 states further 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 would be removed 
because the requirements are captured in proposed Sec.  52.79(a) and in 
the proposed revisions to part 51.
    j. Section 52.91, Authorization to conduct site activities. Section 
52.91(a)(2) currently provides 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 would be 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 would be 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.
    k. Section 52.93, Exemptions and variances. Section 52.93 would 
include a discussion of the requirements regarding requests for an 
exemption from any part of a referenced design certification rule. The 
proposed Sec.  52.93 states that, 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.
    l. Section 52.97, Issuance of combined licenses. The NRC would 
modify 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) would be added, 
consistent with Sec.  50.50, which states that a combined license shall 
contain conditions and limitations, including technical specifications, 
as the Commission deems necessary and appropriate. Existing Sec.  
52.97(b)(2) would be 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 would be added to subpart C, consistent with 
the other subparts in 10 CFR part 52. Section 52.98 would provide 
provisions for the finality of combined license provisions. 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 license, except in accordance with the provisions of 
Sec. Sec.  52.103 or 50.109, as applicable.
    Section 52.98 would include 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

[[Page 12799]]

situation, both change processes may apply.
    Section 52.98(d) would be 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 variances from information within the scope 
of the manufactured reactor's design are subject to the change 
processes in Sec.  52.171. 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 proposes all of these 
requirements to clarify, in one location, the finality provisions 
applicable to all portions of a combined license.
    Finally, the Commission proposes to add a new paragraph (g) to the 
``finality'' section in each subpart of part 52, including Sec.  52.98, 
entitled ``Information requests,'' which would delineate the 
restrictions on the NRC for information requests to the holder of the 
combined license. This provision is analogous to the current 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 would be contained in Sec.  52.98(g), 
and would require 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 exceptions would be 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 would first have to be 
approved by the Executive Director for Operations (EDO) or his or her 
designee.
    n. Section 52.103, Operation under a combined license. Section 
52.103(g) currently requires the NRC to find that the acceptance 
criteria in the combined license are met before operation of the 
facility, but does not refer to loading of fuel. However, current Sec.  
52.103(f) states 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. It was the Commission's intent in 
the 1989 rulemaking that it find that the acceptance criteria have been 
met before fuel is loaded, and the failure to include the reference to 
loading of fuel was an inadvertent oversight. Therefore, this section 
would be 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. In addition, Section IX in each of appendices A, B, 
and C of part 52 requires that the Commission find that the acceptance 
criteria in the ITAAC for the license are met before fuel load. The NRC 
believes that this is the common interpretation of Sec.  52.103(g).
    o. Section 52.104, Duration of combined license; Section 52.105, 
Transfer of combined license; Section 52.107, Application for renewal; 
Section 52.109, Continuation of combined license; and Section 52.110, 
Termination of license. Five new provisions would be added to Part C 
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, would address the duration of a combined license and contains 
requirements that currently exist in Sec.  52.83. In addition, the 
Commission proposes to amend 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 would provide requirements for the transfer of a 
combined license that refer the applicant to Sec.  50.80. Section 
52.107 would provide a reference to 10 CFR part 54 for the renewal of a 
combined license.
    Section 52.109 would provide provisions for the continuation of a 
combined license and Sec.  52.110 would provide requirements for the 
termination of a combined license. Currently, part 52 does 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 would 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.
7. Subpart D, Reserved
8. Subpart E, Standard Design Approvals (Sec. Sec.  52.131 Through 
52.147)
    Appendix O to part 52 currently sets forth the NRC's requirements 
for approval of standard designs for nuclear plants 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 as part of four design certification 
reviews where the FDAs were a prerequisite to certification of the 
standard design. As part of this rulemaking, the NRC proposes to remove 
the requirement that FDAs are a prerequisite to a design certification 
under subpart B of part 52 (see the discussion on 10 CFR 52.43).
    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 NRC did undertake a re-
examination of appendix O to part 52 and proposed certain changes in 
the 2003 proposed rule. In view of the substantial reorganization and 
rewriting of part 52 proposed in this rulemaking, the Commission has 
given further consideration to the licensing process in appendix O to 
part 52 and proposes additional changes to enhance the regulatory 
effectiveness and efficiency of that process.
    The NRC 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 NRC favors the use of the design certification 
process, which suggests

[[Page 12800]]

that the design approval process could be eliminated. However, given 
the frequent use of appendix O to part 52 in the past, the NRC proposes 
to retain this process and to reorganize and reformat the design 
approval process to be consistent with the other subparts.
    The language that is currently in appendix O of part 52 has been 
relocated to a new subpart and formatted to be consistent with the 
other subparts. A new section (Sec.  52.133) would be created to 
describe the relationship of the design approval process with the other 
subparts. The proposed filing requirements are consistent with the 
other subparts. The applications 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 power plant designers are technically capable of developing 
essentially complete and final design information for Commission review 
and approval. Furthermore, the economic incentives with respect to 
design certification also apply to final design approvals. In addition, 
approval of a final reactor design removes 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 before 
there is any substantial commitment of resources associated with the 
construction of the plant, which will greatly enhance regulatory 
stability and predictability.
    The NRC also proposes that applications for standard design 
approvals provide essentially the same technical information that is 
required for design certification applications (e.g., demonstration of 
compliance with any technically relevant Three Mile Island (TMI) 
requirement, proposed technical resolutions of unresolved safety issues 
and medium- and high-priority generic safety issues, and a design-
specific probabilistic risk assessment). This proposal is consistent 
with past practice regarding applications for future designs and would 
implement the Commission's Policy Statements on Severe Reactor 
Accidents (50 FR 32138; August 8, 1985) and Nuclear Power Plant 
Standardization (52 FR 34884; September 15, 1987). However, this 
proposal would not require applicants for standard design approvals to 
submit ITAACs because FDAs may be referenced in applications for 
construction permits or operating licenses under 10 CFR part 50, and 
the verification process used for part 50 applications does not use 
ITAAC. In addition, this proposal would not require applicants to 
consider severe accident mitigation design alternatives.
    A new Sec.  52.139, which specifies the standards that will be used 
to review applications for standard design approvals and new Sec. Sec.  
52.145 and 52.147, which specify the finality and duration of standard 
design approvals consistent with other subparts would be added. In a 
letter dated November 13, 2001, NEI commented that ``Industry 
recommends FDAs be valid for 15 years.'' The NRC agrees with the 
industry's recommendation. The Commission has decided that the duration 
of standard design approvals should correspond to the duration of 
design certifications, inasmuch as both standard design approvals and 
design certifications constitute approvals of nuclear power plants 
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.
9. Subpart F, Manufacturing Licenses
    The following discussion explains the requirements in subpart F 
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.
    Appendix M of part 52 currently sets forth the NRC's requirements 
governing manufacturing licenses. Appendix M of part 52, which was 
first adopted by the NRC in 1973, provides 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 current licensing regime in appendix M 
of part 52, the NRC does not approve a final reactor design to be 
manufactured before issuance of the manufacturing license. Rather, 
analogous to the two-step process, the NRC issues a manufacturing 
license based upon the review of a preliminary design equivalent to 
that provided in a construction permit application. Upon approval of 
the preliminary design and associated information, the NRC issues a 
manufacturing license authorizing the manufacture--but not the removal 
from the manufacturing site--of one or more nuclear power reactors. 
Thereafter, manufacturing can commence, although the NRC must approve 
the final design of the manufactured reactor by license amendment (see 
appendix M of part 52, paragraph 7, Note). Under paragraph 8 of 
Appendix M of part 52, the manufactured reactor may not be removed from 
the place of manufacture until approval of the final design under 
paragraph 7 of appendix M of part 52.
    When the NRC adopted part 52 in 1989, the NRC did not re-examine 
the regulatory scheme for manufacturing licenses to determine if the 
bases for adopting part 52 and the licensing concepts used in part 52 
also would be an impetus for proposing changes to 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, in view of the substantial reorganization and rewriting of 10 
CFR Chapter 1 generally, the NRC has reconsidered the efficacy of the 
current manufacturing license process in appendix M of part 52 and 
proposes substantial changes to enhance regulatory effectiveness and 
efficiency.
    The most important shift in the manufacturing license concept 
proposed by the NRC 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 current 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

[[Page 12801]]

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 proposed Sec. Sec.  52.157 
and 51.158, 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 early site permits. Section 52.157 
would require the standard information to be submitted in support of 
the design of a reactor (derived from the existing requirements in 
current part 52, subparts B and C) for a standard design certification 
and combined license. In addition, the application must 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 proposed 
revisions to Sec.  50.40 (as discussed separately with respect to Sec.  
50.40, these testing and prototype requirements proposed to be 
incorporated into Sec.  50.40 were derived from the current 
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 a 
PRA, 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.
    The environmental report must address severe accident mitigation 
design alternatives (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.
    In light of the Commission's review and approval of a final design, 
the NRC proposes to provide a greater degree of finality to a 
manufacturing license. Under Sec.  52.171(a)(1) of the proposed rule, 
the same degree of issue finality accorded to the ``certified design'' 
would apply throughout the term of the manufacturing license. Under 
this provision, the approved design for the manufacturing license could 
not be changed or modified unless the NRC determines it is necessary 
either for adequate protection or for compliance with requirements 
applicable and in effect at the time the manufacturing license was 
issued. A comparable requirement is also included in Sec.  52.171(a)(4) 
which would restrict changes to the design of the manufactured reactor 
if it is referenced for use in a construction permit, operating 
license, or combined license. The NRC proposes not to provide the 
ability of the manufacturing license holder to make changes to the 
design, site parameters, design characteristics, or terms and 
conditions under the provisions of 10 CFR 50.59, which is comparable to 
the design certification process. The NRC believes that 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.
    The NRC proposes that the term of a manufacturing license be for no 
less than 5 or more than 15 years from the date of issuance. The 
licensee 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 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. The NRC does not 
propose to specify, as a separate matter, 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 nor does it authorize ``construction,'' there does 
not appear to be any legal need for the manufacturing license to 
specify, apart from its term, the earliest and latest date of 
completion of manufacture.
10. Subpart G, Reserved
11. Appendices A, B, and C--Design Certifications for ABWR, System 80+, 
and AP600
    The NRC proposes to amend paragraphs VI.B.4, 5, and 6 of the three 
design certification rules 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 design 
certification rule are resolved when an exemption has been granted for 
a plant referencing the design certification rule. Paragraph VI.B.5 
describes how issues are resolved when a plant referencing the design 
certification rule 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.

[[Page 12802]]

Thus, once a matter (e.g., an exemption in the case of paragraph 
VI.B.4) is addressed for a specific plant referencing a design 
certification rule, 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. Unfortunately, the three 
design certification rules use the phrase ``but only for that 
proceeding,'' which may lead to the erroneous conclusion that issue 
resolution exists only in the proceeding in which the matter was 
approved and/or adjudicated, and not in all subsequent proceedings for 
that plant.
    In letters dated November 12, 2001, and November 13, 2001, 
respectively, General Electric Company and Westinghouse Electric 
Company reiterated earlier recommendations the two companies had made 
that Sections VI.B.4 and 5 of the design certification rules state that 
exemptions and license amendments have finality ``but only for that 
plant.'' For the reasons previously discussed, the NRC proposes to 
substitute the phrase ``but only for that plant,'' to clarify that 
issue resolution on a matter applies in subsequent proceedings for that 
plant.
    Each of the design certification rules in appendices A, B, and C to 
part 52 includes a Section VIII on change processes. These processes 
apply to changes depending upon the category of design information 
affected. For plant-specific Tier 2 information, the change process 
established in the rule mirrors, in large part, that in the former 10 
CFR 50.59. The proposed rule would amend paragraph VIII.B.5 of the 
design certification rules to conform the terminology in the Sec.  
50.59-like change process to that used in the current Sec.  50.59. This 
amendment deletes references to unreviewed safety question and safety 
evaluation, and conforms the evaluation criteria concerning when prior 
NRC approval is needed. Also, a definition has been added to the design 
certification rules (paragraph II.G) for ``departure from a method of 
evaluation'' to support the evaluation criterion in Paragraph 
VIII.B.5.b(8).
    In an earlier rulemaking (see 64 FR 53582; October 4, 1999), the 
NRC revised Sec.  50.59 to incorporate new thresholds for permitting 
changes to a plant as described in the FSAR without NRC approval. For 
consistency and clarity, similar changes are being proposed for 10 CFR 
part 52 applicants or licensees. Because of some differences in how the 
change control requirements are structured in the design certification 
rules, certain definitions contained in Sec.  50.59 are not necessary 
for or applicable to 10 CFR part 52 and are not being included in this 
proposed rule. One definition that the NRC is including, is 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 the design certification rules will be 
implemented as intended.
    Each of the design certification rules in appendices A, B, and C to 
part 52 includes a section on records and reporting. The NRC proposes 
to amend paragraph X.B.3.b 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 
design certification rule to the date that the Commission makes its 
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 its 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 certification information, and to adjust the inspection program 
to reflect the design changes.
    The proposed amendment to paragraph X.B.3.b reduces the frequency 
of reporting during the period of construction and increases the 
frequency of reporting during the application review period. The 
Commission 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 under Sec.  52.103(g), the NRC must monitor 
the design changes made under Section VIII of the design certification 
rules. 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, detailed design of the plant before and during construction, 
and during preoperational 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 the requirements for operating 
plants.

D. Proposed Changes to 10 CFR Part 50

1. General Provisions, Sec.  50.2, Definitions
    The Commission proposes to add new definitions as conforming 
changes to Sec.  50.2. The definition of an applicant would be added to 
clarify that a person or entity applying for Commission ``permission or 
approval'' is an applicant. This would ensure that part 50 requirements 
for applicants would apply to a person or entity seeking an NRC 
approval not constituting a license, such as a standard design approval 
under part 52.
    The definitions for license and licensee would be 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.
    The definition for prototype plant would be added to explain the 
type of nuclear reactor that the NRC intends in the proposed 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) 
currently prohibits 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) would be revised to refer to combined licenses.
    Currently, Sec.  52.17(c) authorizes an early site permit applicant 
to request authority to perform the activities allowed under Sec.  
50.10(e)(1). The NRC notes that the current regulation does 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 propose to change the current restriction. 
It will conserve the

[[Page 12803]]

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 propose to 
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 existing 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. This section currently 
provides 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. The revised section clarifies 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 Part 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 NRC notes that Sec.  50.23 does not preclude the 
NRC from combining a construction permit and operating license with 
respect to production facilities or utilization facilities other than 
nuclear power plants under Section 161.h of the AEA.
    b. 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 proposes to make 
conforming changes throughout Sec.  50.30 to include necessary 
references to part 52 processes other than design certification (Part 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) would be 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.
    c. 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 qualification to carry out, in accordance with the NRC's 
regulations, the activities for which a license or permit is sought. 
Section 50.33 would be amended to require applicants for combined 
licenses to submit financial qualifications information. The proposed 
rule would not require financial qualifications information to be 
submitted by applicants for early site permits, 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 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 
under the proposed rule. 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) currently addresses radiological emergency 
response plans for State and local government entities that must be 
submitted in applications for operating licenses. The proposed rule 
would make 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)(iii) to seek NRC 
review and approval of complete emergency plans.
    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 will be provided. The proposed rule would make 
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 NRC's proposed 
revision to Sec.  50.75.
    d. Section 50.34, Contents of construction permit and operating 
license applications; technical information. The NRC is proposing to 
retitle Sec.  50.34 from Contents of applications; technical 
information to Contents of construction permit and operating license 
applications; technical information. Section 50.34(a) currently

[[Page 12804]]

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 
current 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, the current Sec.  
52.47 states 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, 
current Sec.  52.79 states 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 Commission 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 proposes to revise Sec.  50.34 to make it applicable to 
construction permit and operating license applications only and to 
provide separate sections for the technical 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 proposed 
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) would 
be amended to indicate that each applicant for a design certification, 
design approval, or combined 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), rather than repeating 
the requirements in each of the relevant sections in part 52.
    e. Section 50.34a, Design objectives for equipment to control 
releases of radioactive material in effluents--nuclear power reactors; 
and Section 50.36a, Technical specifications on effluents from nuclear 
power reactors. Section 50.34a currently 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 would be amended to clarify its applicability to the 10 
CFR part 52 licensing and approval processes. Section 50.34a currently 
applies to combined licenses by virtue of the provision in current 
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. Current 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 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 which is technically relevant to the design and not site-
specific. Current appendix O to 10 CFR part 52, section O.3, explicitly 
requires applicants for design approvals to include the applicable 
technical information required by Sec.  50.34a. Finally, current 
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 proposed 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 proposed 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.
    f. 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 
current 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 current appendix O, section O.3, which 
states that the submittal for review of a standard design shall include 
the applicable

[[Page 12805]]

technical information under Sec. Sec.  50.34 (a) and (b), as 
appropriate.
    Section 50.36 would be revised to clarify that design approval and 
design certification applications must also include proposed technical 
specifications. The new proposed provisions in Sec.  50.36(c) would 
require each applicant for a design approval or a design certification 
to include proposed generic technical specifications in its application 
for the portion of the plant that is within the scope of the design 
approval or design certification application.
    g. Section 50.36a, Technical specifications on effluents from 
nuclear power reactors. Section 50.36a(a) currently 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 existing language in Sec.  50.36a(a) 
encompasses 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 which is technically relevant 
to the design and not site-specific. Section 50.36a(a) would be amended 
to state that each licensee of a nuclear power reactor and each 
applicant for a design certification 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 NRC is proposing to make conforming changes to appendix I to 10 
CFR part 50. These proposed changes parallel the proposed changes to 
Sec. Sec.  50.34a and 50.36a.
    h. Section 50.37, Agreement limiting access to Classified 
Information. Section 50.37 currently 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. Current Sec.  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 so stated therein or not. 
The existing language in Sec.  50.37 encompasses early site permit, 
combined license, and manufacturing license applicants under 10 CFR 
part 52 because these products are all licenses. However, the NRC 
proposes to modify Sec.  50.37 to encompass applicants for design 
certification and for standard design approvals under 10 CFR part 52 
for consistency with the proposed changes to 10 CFR part 25, Access 
Authorization for Licensee Personnel. Part 25 sets forth the 
Commission's requirements governing the grant of access authorization 
to classified information to certain individuals, and the Commission is 
proposing modifications to part 25 to reflect the licensing and 
regulatory approval processes in part 52. Accordingly, the Commission 
proposes to make consistent changes to Sec.  50.37. The proposed Sec.  
50.37 would 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. Proposed 
Sec.  50.37 would also require that this agreement be part of the 
application and be deemed part of the license, or construction permit, 
or NRC standard design approval whether so stated therein or not. The 
NRC proposes to modify Sec.  52.54, Issuance of standard design 
certification, 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 proposed changes, along with the 
proposed changes to parts 25 and 95, are necessary to ensure that 
access to classified information is adequately controlled by all 
entities applying for NRC licenses, design certifications, or design 
approvals.
4. 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) 
would be 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 would be 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) would be 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 would conform 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 IV.B of this document.
    c. Section 50.45, Standards for construction permits, operating 
licenses, and combined licenses. This section would be revised to 
clarify that the standards for authorizing construction or alteration 
of a facility also apply to applications for combined licenses issued 
under part 52.
    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 systems (ECCS) evaluation models. The proposed rule would 
add conforming references to design approvals, design certifications, 
and licenses issued under part 52 so that the NRC will be notified of 
changes to or errors in acceptable evaluation models that were used in 
licenses, certifications, and approvals issued under part 52.
    e. Section 50.47, Emergency plans, Section 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. These 
regulations do not clearly address early site permit or combined 
license applicants or holders. Accordingly, the NRC proposes

[[Page 12806]]

to make a number of changes in these regulations. Section 50.47(a)(1) 
currently 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. Section 50.47(a)(1) would be revised to 
include combined licenses in these applicability statements. A new 
Sec.  50.47(a)(1)(ii) would be added to include similar requirements 
for early site permit applicants that submit complete and integrated 
emergency plans.
    Section 50.47(c)(1) provides a process for operating license 
applicants that fail to meet the applicable standards of Sec.  
50.47(b). Section 50.47(c)(1) would be revised to clarify that this 
process is applicable to combined license applicants as well.
    Section 50.47(d) currently provides that no NRC or Federal 
Emergency Management Agency (FEMA) 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 states that a 
license authorizing fuel loading and/or low-power testing and training 
may be 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. A new Sec.  50.47(e) would be added 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 proposed 
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), when it determines whether to authorize 
fuel loading and operation. 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. Proposed Sec.  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, FEMA 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. Proposed Sec.  50.54(gg) would also provide the 
standards by which the NRC will base such a finding.
    Appendix E to part 50 would be revised 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 proposes to add a 
parallel statement for combined license applicants, and to add a 
statement that an early site permit applicant may submit emergency 
plans. Similar modifications are proposed in Section III of Appendix E 
to part 50 regarding the content of final safety analysis reports and 
early site permit applications. In Section IV of Appendix E to part 50, 
Content of Emergency Plans, the NRC proposes to modify 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 proposes to designate the requirements 
for operating licenses as paragraph F.2.a.i, and to add a new paragraph 
F.2.a.ii that contains the requirements for combined licenses. Proposed 
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 NRC further 
proposes that, if FEMA 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.
    A new paragraph IV.F.2.a.iii would be added 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 FEMA 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) would 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

[[Page 12807]]

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 also proposes to modify section V of appendix E to part 50, 
Implementing Procedures, which states 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 proposes to clarify that 
paragraph V is also applicable to combined license 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 date that 
the Commission authorizes fuel load and operation under Sec.  52.103.
    f. Section 50.48, Fire protection. Section 50.48(a)(1) would be 
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) would be 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) and (k) 
would be revised to clarify that these programmatic requirements apply 
to applicants for and holders of operating licenses issued under part 
50 and combined licenses under part 52.
    h. Section 50.54, Conditions of licenses; and Section 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 proposed to be 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 proposed to be covered in Sec.  50.55. Combined licenses 
represent a special case, inasmuch as they address both construction 
and operation. The NRC proposes to address combined licenses by placing 
the conditions applicable to construction in Sec.  50.55, which would 
indicate that these conditions are applicable until the date that the 
NRC authorizes fuel load and operation under Sec.  52.103. Conditions 
which are applicable during operation would be set forth in Sec.  
50.54, and indicate that these conditions are applicable on the date 
that the NRC authorizes fuel load and operation under Sec.  52.103.
    The introductory paragraph of Sec.  50.54 would be revised to refer 
to combined licenses, and to exclude manufacturing licenses from its 
provisions. Section 50.54(a)(1) would be revised 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 proposes to revise Sec.  50.54(i-1) to indicate its 
applicability to combined licenses. Specifically, Sec.  50.54(i-1) 
would require that within three 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 proposes to add Sec.  50.54(gg). These revisions are 
discussed with related requirements in section IV.D.4.f of this Federal 
Register document, ``Section 50.47, Emergency plans, Section 50.54(gg), 
and appendix E to part 50, Emergency planning and preparedness for 
production and utilization facilities.''
    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 would be revised to 
specify which paragraphs apply to a construction permit, early site 
permit, combined license, and manufacturing license.
    Sections 50.55(a) and (b) would be revised to require a combined 
license and manufacturing license to state the earliest and latest 
dates for completion of construction or modification, and to provide 
for forfeiture of the combined license or manufacturing license if 
construction, manufacture, or modification is not completed by the 
stated date. In the case of a manufacturing license, the license would 
be required to state the earliest and latest date of manufacture for 
each reactor. The NRC believes that Section 185.a of the AEA requires 
that a construction permit state the earliest and latest date for 
completion of construction, and applies to a combined license because a 
combined license includes the authority granted under a construction 
permit. The NRC believes that the 1992 amendment of Section 185.b of 
the AEA addressing combined licenses did not supercede and render 
nugatory the provisions of Sec.  50.54a. 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. Nonetheless, because a manufacturing license 
authorizes activities which are analogous to those in a construction 
permit, it makes sense from a regulatory standpoint to treat 
manufacturing licenses similar to construction permits.
    Section 50.55(c) makes the conditions in Sec.  50.54 also apply to 
construction permits, unless otherwise modified. The NRC proposes to 
retain this paragraph and add a reference to combined licenses. 
Manufacturing licenses would not be referenced, because there does not 
appear to be any regulatory need to apply any of the conditions in 
Sec.  50.54 to manufacturing 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.

[[Page 12808]]

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 proposed changes to part 21, the NRC proposes to 
retain 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 (albeit limited to the reactor as opposed to 
the entire facility in the case of a construction permit or combined 
license). 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 proposed subpart E (current 
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 proposes to revise 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 
below in section J on part 21, early site permits do not authorize 
``construction'' or its functional equivalent. Therefore, early site 
permits would be subject to the requirements of part 21 rather than 
Sec.  50.55(e) under the proposed 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.12.b, ``Appendix B to Part 
50--Quality Assurance Criteria for Nuclear Power Plants and Fuel 
Reprocessing Plants.''
    i. Section 50.55a, Codes and standards. Section 50.55a currently 
provides requirements relating to codes and standards for construction 
permits and operating licenses for boiling or pressurized water-cooled 
nuclear power facilities. The proposed rule would amend 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 currently applies to combined licenses by virtue of the 
provision in current 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. Also, Sec.  50.55a 
currently applies to design certifications by virtue of the provision 
in current Sec.  52.48, Standards for review of applications, 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 current 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 proposing to 
require 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, current 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 proposes to modify 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 proposed 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) would be 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) would be 
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) would be 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.
    k. Section 50.61, Fracture toughness requirements for protection 
against pressurized thermal shock events. This section would be 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 would be 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. 
The proposed Sec.  50.62 would 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 Commission is proposing 
to add 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 
this information in their final safety analysis reports.

[[Page 12809]]

    m. Section 50.63, Loss of all alternating current power. Conforming 
changes would be 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 a maintenance program at nuclear plants. Section 
50.65(a) would be revised to clarify that holders of operating licenses 
issued under part 50 and combined licenses issued under part 52 must 
have a maintenance program. Section 50.65(c) would be revised to 
specify that for new licenses issued after the effective date of this 
regulation, the maintenance program must be implemented before the 
initial fuel loading of the reactor.
5. Inspections, Records, Reports, Notifications
    a. Section 50.70, Inspections. Section 50.70(a) currently 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 existing 
language in Sec.  50.70(a) encompasses combined license holders and 
manufacturing license holders because they are licensees. In addition, 
the provision in current Sec.  52.83, Applicability of part 50 
provisions, 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, current 
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. The proposed rule would amend Sec.  50.70(a) 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 proposing to clarify 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 proposed 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) currently 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 current 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 proposes to revise 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) currently 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 current 
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 apply. 
Accordingly, the NRC proposes to modify 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 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 proposes to amend 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 current regulation, however is deficient in two respects. First, it 
does not address the updating requirements for combined license holders 
where the combined license references a standard design certification. 
Second, the current regulation, if applied to manufacturing licenses as 
proposed under subpart F, would impose unnecessary regulatory burden 
with respect to periodic updating. The NRC's concept of a manufacturing 
license under subpart F is for a relatively stable, unchanging design. 
Hence, there should be no need for periodic updating. Rather, the 
updating should occur only as the result of Commission-approved changes 
to the design.
    Accordingly, the NRC proposes to amend Sec.  50.71(e) to specify 
the FSAR updating requirements for combined license holders where the 
license references a standard design certification. In addition, 
current Sec.  50.71(f) would be redesignated as Sec.  50.71(g), and add 
a new Sec.  50.71(f), addressing the FSAR update requirements for a 
manufacturing license. Proposed Sec.  50.71(f) would 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 proposed Sec.  
52.171, or any new analyses requested to be performed by the NRC. 
Periodic updating of a 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

[[Page 12810]]

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.
    c. Section 50.73, Licensee event report system. Section 50.73 
currently 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 current language does not impose 
the LER requirement on combined license holders. Accordingly, in a 
conforming change, the NRC proposes to extend 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 proposed 
rule does not extend the LER requirement to other part 52 processes for 
similar reasons, viz., 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 50.75, 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. In brief, Sec.  
50.75 currently requires a nuclear power plant operating license 
application to: (i) address the predicted costs of decommissioning; 
(ii) describe the method(s) for adjusting the cost prediction 
throughout the life of the plant to address the effects of inflation; 
and (iii) provide financial assurance by one of the alternatives 
specified in the regulation, and to submit evidence that one or more of 
these means has been established. The regulation also establishes a 
requirement to update the cost estimates for decommissioning, and to 
describe any adjustments to the amount of funds collected annually to 
reflect any changes in projected decommissioning cost.
    The current requirements are directed at the two phase construction 
permit followed by operating license patterns in part 50, and are not 
well-suited to address the licensing process associated with a combined 
license under part 52. For example, requiring the combined license 
applicant to comply with the current requirement in Sec.  50.75(b)(1) 
that the operating license applicant submit a copy of the financial 
instrument obtained to satisfy the requirements of Sec.  50.75(e), 
would in essence place a more stringent requirement on the combined 
license applicant inasmuch as it would be required to fund 
decommissioning assurance at an earlier date as compared with the 
operating license applicant. To address these discrepancies, the NRC 
proposes to revise Sec. Sec.  50.75(b) and 50.75(e)(1) to address 
decommissioning funding assurance for combined licenses. Under the 
proposed rule, the combined license applicant must submit a 
decommissioning report as required by Sec.  50.33(k), but it need not 
provide a financial instrument to fund decommissioning or to submit a 
copy to the NRC. Instead, under proposed Sec.  50.75(b)(1) and (4), the 
combined license must contain a certification that the financial 
assurance would be provided no later than 30 days after the NRC 
publishes notice in the Federal Register under Sec.  52.103(a). 
Following the issuance of a combined license, the holder must 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). No later than 30 days after the 
Commission publishes notice in the Federal Register under Sec.  
52.103(a), the holder is required to submit 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 authorization to load fuel and 
operate is provided to the license holder under Sec.  52.103, the 
combined license holder is 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.
    The Sec.  50.75 decommissioning funding requirements could be 
interpreted as applying to an applicant for, and holder of a 
manufacturing license under part 52. The NRC did not have such intent 
when it adopted Sec.  50.75. 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. 
Accordingly, a conforming change is proposed for Sec. Sec.  50.33(k) 
and 50.75(a) to exclude the applicants for and holders of manufacturing 
licenses under part 52 from compliance with the requirements of that 
section.
6. US/IAEA Safeguards Agreement
    a. Section 50.78, Installation information and verification. Since 
1980, the United States 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 currently 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, the NRC proposes to revise Sec.  
50.78 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.
7. 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

[[Page 12811]]

and all licenses under part 52, Sec.  50.80(a) would be revised to 
explicitly refer to permits under parts 50 and 52, and licenses under 
part 52.
    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 proposes to revise the definition of license and facility. The 
definition of license in this section would be revised to explicitly 
refer to all licenses under 10 CFR, and early site permits under part 
52. The definition of facility would be revised to add a new paragraph 
which would explicitly refer to an early site permit under part 52, and 
a reactor manufactured under a manufacturing license under part 52.
8. 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 
proposes to make 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 proposed to be amended in 
Sec.  50.2 would include combined licenses, manufacturing licenses, and 
early site permits under part 52, Sec.  50.92 would be 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 
proposes to amend 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 and 
manufacturing 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) would be 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 obtain 
a new construction permit in addition to a license amendment, inasmuch 
as NRC review of the alteration is assured.
9. 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 authorizes 
the NRC to suspend, modify or revoke 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 Commission proposes to revise Sec.  50.100 
by adding a new paragraph (b) explicitly addressing the Commission's 
authority.
10. Backfitting
    a. Section 50.109, Backfitting. The backfit rule provides certain 
protection to licensees against changes in the NRC requirements and NRC 
staff positions on those requirements. The backfitting provisions in 
Sec.  50.109 currently apply to standard design approvals, construction 
permits, and operating licenses, see Sec.  50.109(a)(1)(i)-(iv), but do 
not address combined licenses, or manufacturing licenses. Part 52 
contains special backfitting requirements on early site permits, design 
certification rules, but neither Sec.  50.109 or part 52 currently 
address 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, the Commission is 
proposing to revise Sec.  50.109(a)(1) 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, the proposed 
rule would reference 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.
11. 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. The NRC proposes a 
conforming amendment to add applicants for and holders of combined 
licenses as being subject to this provision.
12. 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 would be 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 sets forth requirements for a main control 
room in a nuclear power

[[Page 12812]]

plant. The NRC proposes 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 structures, systems, and components 
(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 proposes to revise 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 would state that every applicant for a combined license is 
required by the provisions of Sec.  52.79 to include in its final 
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 to the managerial and administrative 
controls to be used to assure safe operation. The introduction would 
also state 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. Finally, the introduction would state that every applicant 
for a design approval, design certification, or manufacturing license 
is required by the provisions of Sec. Sec.  52.137, 52.47, and 52.157, 
respectively, to include in its final 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.
    The NRC proposes to maintain the current regulatory structure for 
requirements that implement Appendix B 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 proposes to revise 
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).
    Appendix B to part 50 is currently 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. Appendix B to part 50 currently applies 
to design certifications by virtue of the provision in current Sec.  
52.48, Standards for review of applications, which states 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. Appendix O to part 52, section O.3, requires 
applicants for design approvals to include the information required by 
Sec. Sec.  50.34(a) and (b), as appropriate, and states 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 
currently applies to manufacturing licenses by virtue of the provision 
in current appendix M to part 52, section M.1, which 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.
    Early site permits are considered partial construction permits; 
therefore, the Commission 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 Commission 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 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 Commission proposes to modify appendix B 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, combined licenses, and manufacturing licenses.
    The title of Appendix C to part 50 would be revised. 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.
    With the exception of manufacturing licenses, none of the other 
regulatory

[[Page 12813]]

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.\3\ Therefore, the proposed 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 no such review is conducted by the NRC for fabricators of nuclear 
power plant systems and components.
---------------------------------------------------------------------------

    \3\ 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 
IV.D.4.f of this Federal Register notice.
    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 proposing 
changes to Appendix I that conform to the changes being proposed in 
Sec. Sec.  50.34a and 50.36a. Specifically, a statement would be added 
in Section I that states 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 would be revised to state that the guides on design objectives set 
forth in Appendix I 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). Finally, Section IV would be revised to state that the 
guides on limiting conditions for operation for light-water-cooled 
nuclear power reactors in Appendix I 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.
    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 (1) leakage through the 
primary reactor containment systems and components penetrating primary 
containment shall not exceed allowable leakage rate values as specified 
in the technical specifications or associated bases; and (2) 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 proposes to amend appendix J to 
part 50 to clarify that these requirements also apply to combined 
licenses under 10 CFR part 52, as is currently indicated by 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.
    g. Appendices M and O to part 50 [Removed]. The proposed rule would 
remove appendices M and O from 10 CFR part 50. Appendix M addresses 
Appendix M provides 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 addresses the early review of site suitability 
issues. 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 proposed 
rule would amend appendix S to part 50 to clarify that the requirements 
in appendix S to part 50 also apply to applicants for design approvals 
and manufacturing licenses issued under 10 CFR part 52. Although 
current appendix O to part 52 does not explicitly require applicants 
for design approvals to comply with the requirements of appendix S to 
part 50, the NRC is proposing to require 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, current 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 Commission proposes to modify 
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 proposes conforming changes to the Introduction, paragraph 
(a) to appendix S to part 50, and proposes to add definitions for 
design approval and manufacturing license to Section III, Definitions, 
of appendix S to

[[Page 12814]]

part 50, consistent with the definitions in proposed part 52.

E. Proposed Change to 10 CFR Part 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) would be 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. Proposed Changes to 10 CFR Part 2

1. Section 2.1, Scope
    The procedures in 10 CFR part 2 apply to, inter alia, proceedings 
concerning standard design approvals and standard design certifications 
under part 52. Moreover, subpart H of part 2 applied to rulemakings. 
Accordingly, the statement of scope for part 2 would be revised by 
adding a reference to rulemaking and standard design approvals.
2. Section 2.4, Definitions
    The definitions of contested proceeding, license, and licensee, 
would be revised in part 2 by adding conforming references, as 
appropriate, to the licensing processes in part 52. The revised 
definition of contested proceeding would clarify that contested 
proceedings include those involving permits, such as early site permits 
and construction permits. The revised definition of license, would 
ensure 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 definition of 
licensee would be revised to ensure 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 Part
    This section would be 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 is 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, as discussed in the discussion 
on Sec.  2.4, the definitions of ``license'' and ``licensee'' have been 
revised to explicitly refer to early site permits.
5. Section 2.102, Administrative Review of Application
    This section would be revised by adding conforming references in 
Sec.  2.102(a) to applications for early site permits, standard design 
approvals, and combined licenses and manufacturing licenses under part 
52. Under the revised section, the NRC staff would 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(a) identifies in general the content for notices of 
hearing published in the Federal Register. Section 2.104(a) would be 
revised by adding conforming references to a combined license and early 
site permit, to indicate that the NRC will provide at least 30 days 
notice in the Federal Register of a hearing.
    Currently, Sec.  2.104(b) establishes the minimum content of the 
notice of (mandatory) hearing for a construction permit, and Sec.  
2.104(c) establishes the minimum content of the notice of opportunity 
for hearing for an operating license under part 50. The NRC believes 
that there is some benefit, in terms of public transparency and 
regulatory efficiency and consistency, in establishing the minimum 
content for notices of hearing for part 52 licensing processes. 
Accordingly, current Sec.  2.104(d) would be redesignated as Sec.  
2.104(l), and Sec.  2.104(e) would be redesignated as Sec.  2.104(m); 
new Sec. Sec.  2.104(d), (e), and (f), would be added to establish the 
content of notices of hearing for early site permits, combined 
licenses, and manufacturing licenses, respectively. Each of these 
paragraphs is modeled on the notice of hearing for construction permit, 
but modified to reflect the criteria for determining the application, 
as reflected in Sec. Sec.  52.24, 52.97, and 52.167, for early site 
permits, combined licenses, and manufacturing licenses, respectively. 
The NRC notes that manufacturing licenses do not, per se, authorize 
construction of a nuclear power plant. Therefore, a mandatory hearing 
for a manufacturing license is not required under Section 189.a.(1)(A) 
of the AEA. The NRC proposes to provide a mandatory hearing as a matter 
of discretion, in large part because the NRC has never issued a 
manufacturing license of the type contemplated in proposed subpart F of 
part 52. Once the NRC has gained experience in the issuance of 
manufacturing licenses and their oversight, the NRC may in the future 
remove the requirement for a mandatory hearing associated with a 
manufacturing license.
    Section 2.104(e) currently requires the NRC to transmit a notice of 
a 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. As previously noted, the NRC 
proposes redesignating the Sec.  2.104(e) notice provisions as Sec.  
2.104(m). In addition, Sec.  2.104(m)(1) would be revised to clarify 
that the notice would be provided for applications for early site 
permits, combined licenses, but not for 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 in Section II.C.9). Because part 52 also provides an 
opportunity for hearing with respect to its finding under Sec.  52.103, 
the NRC proposes to place the language in Sec.  2.104(e)(2) in Sec.  
2.104(m)(3), and to add Sec.  2.104(m)(2) which indicates that notice 
of opportunity for hearing will be provided to the appropriate State 
official, and the chief executive of the municipality or county as 
applicable.
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, Sec.  2.105(a)(4) would be revised to 
clarify that the procedures in Sec.  2.105 also apply to applications 
for amendments of combined licenses and manufacturing licenses.

[[Page 12815]]

    Under current Sec.  52.103(a), the NRC publishes 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 proposes to revise Sec.  2.105 by adding Sec.  
2.105(a)(12) which addresses the notice required by Sec.  52.103(a). 
Finally, 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 that identifies the proposed Agency action as making the 
finding under Sec.  52.103(g).
8. Section 2.106, Notice of Issuance
    Section 2.106(a) currently provides 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, this section does not require publication of the document in 
the Federal Register that the Commission has made the finding under 
Sec.  52.103(g). The NRC proposes to revise Sec.  2.106(a) to require 
publication of such document in the Federal Register.
    The NRC also proposes to establish in Sec.  2.106(b)(2), the 
minimum requirements for the contents of such notice, viz., 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 in determining whether to 
authorize fuel load and operation for a combined license under part 52. 
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 these findings in connection with all of its 
regulatory and licensing approvals.
9. Section 2.109, Effect of Timely Renewal Application
    Section 2.109 would be revised to add conforming references to a 
combined license under subpart C of part 52. The revised language would 
clarify 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, Sec. Sec.  2.110(a) and (b) would be 
revised to refer to subpart E of part 52 and appendix Q of part 50. 
Section 2.110(c) would be 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; currently, 
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 
would be 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 would be 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. These approvals, by 
themselves, do not authorize third party action. Therefore, any agency 
action to condition their use would not require an NRC order to the 
holder of a standard design approval.
13. Section 2.340, Initial Decisions; Immediate Effectiveness of 
Certain Decisions
    Section 2.340, in paragraph (a), currently sets forth the 
Commission's provisions governing initial decisions in contested 
proceedings for facility operating licenses. Paragraph (a) reflects the 
Commission's longstanding determination that a presiding officer shall 
not address uncontested issues in operating license proceedings unless 
the presiding officer finds, and the Commission (upon referral of the 
matter) agrees with the presiding officer, that the issue represents a 
serious safety, environmental, or common defense and security matter. 
Paragraphs (b), (f) and (g) set forth the Commission's provisions 
governing the immediate effectiveness of initial decisions on issuance 
or amendment of construction permits and operating licenses. There are 
several apparent inadequacies with this section with respect to part 
52. First, Sec.  2.340(a) does not reflect the limits to the presiding 
officer's authority to decide issues that are not contested, and are 
not within the limited scope of hearings with respect to ITAAC under 
Sec.  52.103(g), and the procedure for challenges to ITAAC under Sec.  
52.103(f). Second, paragraphs (b) and (f), read literally, do not apply 
to either an early site permit proceeding (which is a partial 
construction permit), and paragraphs (f) and (g) do not apply to 
issuance of a combined license (which constitutes both a construction 
permit and operating license). Finally, the language of this section 
does not address the immediate effectiveness of the Commission's 
finding under Sec.  52.103(g) that a combined license's ITAAC have been 
met.
    Accordingly, the Commission proposes to revise Sec.  2.340 to 
address early site permits and combined licenses. The Commission 
proposes to simplify the title of this section, which the Commission 
regards as an editorial change. A new paragraph (a-1) would be adopted 
to reflect the procedure in Sec.  52.103(f) with respect to 
consideration of issues not related to meeting acceptance criteria in 
ITAAC. Paragraph (b) would be revised by adding references to early 
site permits, issuance or amendment of combined licenses, and a 
decision under Sec.  52.103(g) that acceptance criteria in an ITAAC for 
a combined license have been met. An editorial change is made to the 
last sentence of Sec.  2.340(b) to make clear that Commission review 
provisions of Sec.  2.341 are not applicable where the Commission 
itself is acting as the presiding officer.
    Paragraph (c) would be revised to make clear that the Director of 
NRR is authorized to issue an early site permit and combined license 
within 10 days of the issuance of an initial decision. The Commission 
notes that under part 52, no licensing action by the Director of NRR is 
necessary following a Sec.  52.103(g) finding that the combined license 
acceptance criteria have been met, in order for the combined license 
holder to commence fuel load and operation. Hence, no change to Sec.  
2.340 in this regard appears to be necessary.

[[Page 12816]]

    New paragraphs (e), (h), and (i) would be adopted to address 
immediate effectiveness of initial decisions in early site permit 
proceedings, combined license issuance, and amendment proceedings, and 
the Sec.  52.103(g) finding for a combined license, respectively. Each 
paragraph would also describe the Commission's consideration of a 
presiding officer's initial decision in such proceedings. Paragraph (e) 
on early site permits is modeled after current paragraph (f) which 
covers initial decisions in construction permit proceedings. Paragraph 
(h) is modeled on current paragraph (g) for issuance and amendment of 
operating licenses, but with changes to reflect the fact that issuance 
of a combined license does not, by itself, allow operation. Paragraph 
(i) is also modeled on current paragraph (g), but modified to focus on 
the Sec.  52.103(g) finding.
    Finally, existing paragraph (h) would be re-designated as a new 
paragraph (o), and the intervening paragraphs (j) through (n) would be 
reserved for future use to accommodate licensing and regulatory 
procedures that may be adopted by the Commission in the future.
14. Section 2.390, Public Inspections, Exemptions, Requests for 
Withholding
    Section 2.390(a) contains the Commission's general rule that NRC 
records and documents regarding a license, permit or order shall 
ordinarily be made available to the public, unless one or more 
provisions in Sec.  2.390 apply. This section would be revised to make 
clear that Sec.  2.390 also applies to NRC records and documents 
regarding standard design approvals under part 52.
15. Section 2.500, Scope of Subpart
    This section would be revised by adding a conforming reference to 
subpart F of part 52 on manufacturing licenses.
16. Section 2.501, Notice of Hearing on Application Under Subpart F of 
Part 52 for a License To Manufacture Nuclear Power Reactors
    This section would be revised by adding a conforming reference to 
subpart F of part 52 on manufacturing licenses. In addition, paragraph 
(b) of this section would be 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 proposes to 
consolidate in Sec.  2.104, the requirements governing the content of a 
notice of hearing with respect to all part 52 processes.
17. Sections 2.502, 2.503 and 2.504 are Removed and Reserved
    The matters addressed in these sections are addressed with greater 
specificity in proposed subpart F of part 52, consistent with the 
Commission's proposed concept for manufacturing licenses and the 
Commission's proposed prohibition on part 50 license applications 
referencing the use of reactors manufactured under a manufacturing 
license issued under subpart F of part 52.
18. 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, does 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 proposes to revise Sec.  2.800 to 
address standard design certifications. Section 2.800 would be 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 would 
be revised to reflect the additional function of this section. Sections 
2.811 through 2.819 would be 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), would be treated as an 
amending petition for rulemaking under the provisions of Sec. Sec.  
2.801-2.810.
19. Section 2.801, Initiation of Rulemaking
    A conforming change is proposed for Sec.  2.801 to refer to 
applications for standard design certification rulemaking.
20. Section 2.811, Filing of Standard Design Certification Application; 
Required Copies
    Section 2.811 would be added to clarify the requirements that are 
related to the filing of applications for standard design 
certifications, and 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).

[[Page 12817]]

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.
21. 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.
22. 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.
23. 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.
24. 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 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.

G. Proposed Change to 10 CFR Part 10

1. Section 10.1, Purpose; and Section 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, does not reflect the licensing and 
approval processes in part 52. Accordingly, the NRC proposes to make 
several 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.
    The NRC proposes to add Sec.  10.1(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, and 
revise Sec.  10.2(b) to refer to standard design approvals under part 
52 and applicants for consultants (to address the provision of services 
associated with design approvals, who may not be ``employees'' per se).

H. Proposed 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 existing 
regulatory language in part 19. Currently, part 19's regulatory 
requirements and proscriptions apply 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 do not cover holders of 10 CFR part 
52 licenses such as combined licenses, early site permits, and 
manufacturing licenses. Moreover, part 19 applies 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 current regulations would 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 
applies 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 
does 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 proposes to amend 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 
proposes to extend part 19 to cover applicants for and holders of 
standard design approvals and standard design certifications. The NRC 
believes that its regulatory

[[Page 12818]]

authority under Section 211 and Section 401 of the 1974 ERA is much 
broader than the current 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.\4\ 
The provisions in Section 401of 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 current scope of part 19 to address the non-licensing 
regulatory approvals in part 52.
---------------------------------------------------------------------------

    \4\ 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 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).
---------------------------------------------------------------------------

    To implement the NRC's proposed broadening of the scope of part 19, 
Sec. Sec.  19.1 and 19.2 would be 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 contractors remains unchanged in the proposed 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 proposes to amend 
Sec.  19.3 to the terms, regulated activities, and regulated entities. 
Regulated entities would be 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 proposes to delineate 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 would be revised to apply to regulated 
entities, as well as licensees.
    Section 19.31, governing exemptions from part 19, would be revised 
to use language consistent with Sec.  50.12 and proposed Sec.  52.6. 
Unlike the current regulation, which limits a request for exemption to 
a ``licensee,'' the proposed rule would allow ``interested persons,'' 
as well as licensees to request an exemption from one or more 
provisions of part 19. This would 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 would 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 would be 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 would 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. Proposed 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 would be 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 would be revised to 
add a conforming reference to facilities licensed under 10 CFR part 52.
3. Section 20.2203, Reports of Exposures, Radiation Levels, and 
Concentrations of Radioactive Material Exceeding the Constraints or 
Limits
    Sections 20.2203(c) and (d) would be revised to add a reference to 
holders of combined licenses to the procedures on submitting reports.

J. Proposed 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 sets 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 proposed rule did 
not address Section 206 reporting requirements with respect to standard 
design approvals or manufacturing licenses. Moreover, the NRC's 
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, the 
NRC withdraws its earlier proposal and has developed a more complete 
and integrated proposal 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).
    Key principles of reporting under section 206 of the ERA. The NRC 
believes that the extension of NRC's reporting requirements 
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 a NRC 
license, a standard design approval, or standard design certification. 
Second, reporting of defects or failures to comply with associated 
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 \5\ 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

[[Page 12819]]

ensure that it fulfills its Section 206 of the ERA reporting obligation 
in an accurate and timely manner.
---------------------------------------------------------------------------

    \5\ Throughout this discussion, reference to entities, licensees 
and/or applicants includes the contractors and subcontractors of 
those entities, 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 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, upon issuance of the early site permit, 
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 upon issuance of 
the early site permit, 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 a manufactured reactor (as 
applicable) must also continue and cannot be deemed to have ended.
    Some industry stakeholders have 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 at least three 
different 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 stakeholders does not 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

[[Page 12820]]

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 
NRC's view is that the NRC's reporting 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). Once an 
application has been granted, the NRC believes 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 NRC 
believes that 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 NRC'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, early site 
permit, design approval or design certification. Upon initial 
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 currently 
reflected in Sec.  21.31; the NRC proposes to add the corresponding 
requirement to Sec.  50.55(e)(7).
    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 proposed changes to part 21 and 
Sec.  50.55(e) reflect the NRC'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 proposed 
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 would 
be 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 NRC's 
proposed applicability of part 21 and Sec.  50.55(e) to each of the 
various approvals under part 52. The NRC requests comments on whether 
the existing division between part 21 and Sec.  50.55(e) should be 
maintained, or whether the substantive requirements of Sec.  50.55(e) 
should be

[[Page 12821]]

incorporated into part 21, with Sec.  50.55(e) (and/or perhaps another 
regulation in part 50) setting forth a cross-reference to part 21. Note 
that one of the principal differences between part 21 and Sec.  
50.55(e) is that Sec.  50.55(e)(1)(iii)(C) requires reporting of QA 
breakdowns in addition to defects and failures to comply associated 
with substantial safety hazards. The other is that the requirement 
governing commercial grade dedication is only found in part 21.
    Reporting requirements for early site permits. If the early site 
permit holder becomes aware of a significant safety concern with 
respect to its site (e.g., that the specified site parameter 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 
early site permit. This reporting attains special importance given the 
NRC's proposal not to impose an updating requirement for early site 
permit information other than that related to emergency preparedness. 
In order for the applicant for an early site permit 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 early site permit 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 and approval of the early site permit application 
would be subject to approval of the part 21 program.

 Table A-1.--Applicability of NRC Requirements Implementing Section 206 of the Energy Reorganization Act to Part
                                       52 Licensing and Approval Processes
----------------------------------------------------------------------------------------------------------------
                                                                                             Sanctions
  Part 52 Licensing or approval processes         Applicable NRC requirement     -------------------------------
                                             implementing section 206 of the ERA       Civil         Criminal
----------------------------------------------------------------------------------------------------------------
Early Site Permit (SDA); Subpart A
Application *..............................  part 21............................           21.61           21.62
Issuance of ESP............................  part 21............................           21.61           21.62
Standard Design Approval (SDA); Subpart E
Application *..............................  part 21............................           21.61           21.62
Issuance of SDA............................  part 21............................           21.61           21.62
Standard Design Certification Rule (DCR);
 Subpart B
Application *..............................  part 21............................           21.61           21.62
Final DCR rulemaking.......................  part 21............................           21.61           21.62
Combined License (COL); Subpart C
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); Subpart F
Application *..............................  50.55(e)...........................          50.110          50.111
Issuance of ML.............................  50.55(e)...........................          50.110         50.111
----------------------------------------------------------------------------------------------------------------
* Currently, there is no explicit requirement imposing part 21 on an applicant for a construction permit (CP).
  However, as a practical matter the NRC has required these applicants to implement a part 21 program before
  approval of the CP. The Commission proposes to take the same approach with respect to applicants for a COL,
  DCR, ESP, FDA, or ML.

    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 
above, 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 and the standard review plan, 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 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 NRC proposes to impose 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 NRC proposes that the requirements implementing Section 
206 of the ERA, which are applicable to standard design approvals, be 
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 proposes to impose 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

[[Page 12822]]

approval, a design certification does not actually authorize 
construction. Accordingly, the NRC proposes to revise Sec. Sec.  21.3, 
21.21, 21.51, and 21.61 to explicitly refer to an applicant for a 
standard design certification, rather than to revise Sec.  50.55(e).
    Some industry stakeholders 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 reactor) 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 
reactor. Therefore, the NRC proposes that the requirements implementing 
Section 206 of the ERA which are applicable to standard design 
certifications be 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 proposed 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 proposes to revise 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 authorization of fuel load and 
operation under Sec.  52.103, 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.
    Reporting requirements for manufacturing licenses. Under proposed 
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 incorporate 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 manufacturing license, the NRC believes 
that the combined license holder using the manufactured reactor 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.\6\ 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, 
should be contained in Sec.  50.55(e). Accordingly, the NRC proposes to 
revise Sec.  50.55(e) to specifically apply its provisions to holders 
of manufacturing licenses.
---------------------------------------------------------------------------

    \6\ 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.
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K. Proposed Change to 10 CFR Part 25

1. Section 25.35, Classified Visits
    Part 25, which sets forth the NRC's requirements governing the 
granting of access authorization to classified information to certain 
individuals, does not currently reflect the licensing and approval 
processes in part 52. Accordingly, the NRC proposes to make changes to 
ensure that individuals who seek a license, standard design approval, 
or standard design certification under part 52 and require access 
authorization, are subject to the provisions of part 25. Because part 
52 involves entities other than licensees, the NRC proposes to revise 
the title of part 25 to simply read, ``Access Authorization.'' The NRC 
also proposes to revise Sec.  25.35 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.

[[Page 12823]]

L. Proposed Changes to 10 CFR Part 26

1. Section 26.2, Scope, Section 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 implement a subset of part 26 requirements--Sec. Sec.  
26.10, 26.20, 26.23, 26.70, and 26.73--which excludes the drug testing 
provisions in part 26.
    The NRC proposes to extend 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 fuel load and operation under 
Sec.  52.103, analogous to holder of an operating license under part 
50. By contrast, holders of combined licenses, before the date that the 
NRC authorizes fuel load and operation would be required to comply with 
the more limited set of part 26 provisions currently applicable to 
construction permit holders. Similarly, holders of manufacturing 
licenses under subpart F of part 52 would be treated the same as 
holders of construction permits. Finally, persons authorized to conduct 
the limited construction activities allowed under Sec.  50.10(e)(3) 
would also be treated the same as a construction permit holder. The 
proposed rule would accomplish this by: (1) Revising Sec.  26.2(a) to 
refer to combined license holders after the date that the NRC 
authorizes fuel load and operation under Sec.  52.103; (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 under part 
52. 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, the proposed rule would not be 
imposed on applicants for and holders of standard design approvals, and 
applicants for standard design certifications under part 52.

M. Proposed Changes to 10 CFR Part 51

    The proposed rule would make 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 three 
design certification rules in Appendices A, B, and C of part 52, as 
well as the proposed design certification rule for the AP1000 design, 
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 severe accident 
mitigation design alternatives (SAMDAs) into the design certification 
rule. Based upon this experience, the NRC proposes to make changes to 
part 51 to accomplish two objectives.
    First, the NRC proposes to eliminate 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. Therefore, 
the environmental effects of construction and operation of a nuclear 
power facility using the referenced design certification are to be 
addressed in the environmental impact statement (EIS) for the combined 
license. This is practical inasmuch as the full scope and details of 
the benefits and environmental impacts of constructing and operating a 
nuclear power reactor using the design approved in the design 
certification are most likely known at the time when the design 
certification is proposed to be used in a specific nuclear power 
facility at a particular site; this rationale will remain the same for 
all future design certifications. The NRC proposes to revise 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 proposes to require 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 SAMDA 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 has 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 assessments for design certifications should address 
SAMDAs. However, under the current 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

[[Page 12824]]

determining whether an EIS must be prepared, or that a FONSI is 
justified. Accordingly, the NRC proposes that SAMDAs be addressed in 
environmental reports and environmental assessments for design 
certifications.
    The NRC proposes to make a number of changes to accomplish these 
two objectives. Existing Sec.  51.55 would be redesignated as Sec.  
51.58, andSec.  51.55 would be added 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.
    Section 51.30 would be revised to provide for a new Sec.  51.30(d) 
establishing the scope of an environmental assessment for a design 
certification. Section 51.32 (b)(1) and (2) would be added 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 proposes to include 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. Section 51.31 would be modified by 
adding Sec.  51.30(b) specifying the information on the environmental 
assessment to be included in the proposed rulemaking on the design 
certification published in the Federal Register.
    Section 51.50(c)(2) would be revised 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.\7\
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    \7\ 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, Sec.  52.75(c)(2) would be added 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 proposed 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 proposes 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,\8\ 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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    \8\ 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.
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    The NRC proposes to make a number of changes to part 51, in some 
cases parallel to those described above with respect to design 
certifications, consistent with its views on manufacturing licenses. 
Existing Sec.  51.54 would be revised 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 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. Section 51.20(b)(6), which currently

[[Page 12825]]

requires preparation of an EIS for issuance of a manufacturing license, 
and Sec.  51.76, which currently addresses the subject matter of an EIS 
for a manufacturing license, would both be removed from part 51.
    Section 51.30(e) would be revised to establish the scope of an 
environmental assessment prepared for a manufacturing license. Section 
51.32(b)(3) and (4) would be added 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 proposes to include an explicit statement 
of no significant environmental impact for manufacturing licenses in 
Sec.  51.32(b)(3) and (4) to facilitate external stakeholder's 
understanding of the nature of the Commission's action. Section 
51.31(c) would be added to describe the NRC's process for determining 
the manufacturing license with respect to environmental issues covered 
by NEPA.
    Section 51.50(c)(3) would be added 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 would state that 
the environmental report need not address the environmental impacts 
associated with manufacturing the reactor under the manufacturing 
license.
    Finally, Sec.  51.75(c)(3) would be added to indicate that if the 
combined license application proposed 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,\9\ 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.  52.75(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) would explicitly 
provide that the combined license EIS will not address the 
environmental impacts of manufacturing the reactor under the 
manufacturing license.
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    \9\ 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.
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    NEPA obligations associated with Sec.  52.103(g) findings on ITAAC. 
Currently, neither part 51 nor subpart C of part 52 explicitly 
addresses 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 current 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 proposes 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, early site 
permit, 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, early site 
permit, 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.
    More specific changes to individual sections in part 51 are 
discussed below.
    Section 51.20, Criteria for and identification of licensing and 
regulatory actions requiring environmental impact statements. Section 
51.20(b) would be revised to identify the part 52 licensing processes 
that require an environmental impact statement or a supplement to an 
environmental impact statement. Specifically, Sec.  51.20(b)(1) would 
be revised to indicate that issuance of an early site permit requires 
an EIS. Section 51.20(b)(2) would be revised to indicate that issuance 
of a combined license requires an EIS. Also, paragraph (b)(6) would be 
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.
    Section 51.22, Criterion for categorical exclusion; identification 
of licensing and regulatory actions eligible for categorical exclusion 
or otherwise not requiring environmental review. Section 51.22(c) would 
be revised to identify part 52 licensing processes that are eligible 
for categorical exclusion or otherwise do not require environmental 
review.
    Section 51.23, Temporary storage of spent fuel after cessation of 
reactor operation--generic determination of no significant 
environmental impact.

[[Page 12826]]

Sections 51.23(b) and (c) would be revised to indicate that the 
provisions of these paragraphs also apply to combined licenses.
    Section 51.45, Environmental report. Section 51.45(c) would be 
revised to indicate that the analysis in an environmental report 
prepared for an early site permit need not include consideration of the 
economic, technical, and other benefits and costs of the proposed 
action and of energy alternatives. This change is proposed for 
consistency with the provisions of Sec.  52.17(a)(2), which states that 
an environmental report included in an early site permit application 
need not include an assessment of the benefits (for example, need for 
power) of the proposed action and the Commission's denial of a Petition 
for Rulemaking (See PRM-52-02 (October 28, 2003; 68 FR 55905)).
    Section 51.50, Environmental report--construction permit, early 
site permit, or combined license stage. The proposed rule would revise 
the title of Sec.  51.50 to ``Environmental report--construction 
permit, early site permit, or combined license stage,'' and include 
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.
    Where a combined license applicant is referencing an early site 
permit, the NRC staff is proposing to add a requirement in Sec.  51.50 
that the applicant's environmental report need not contain information 
or analyses submitted to the Commission in the early site permit stage, 
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 other significant environmental issue not considered in the 
early site permit proceeding, either for the site or design; and (3) 
any new and significant information on the site or design to the extent 
that it differs from, or is in addition to, that discussed in the early 
site permit EIS. The NRC staff is also proposing to add a requirement 
that the applicant must have a reasonable process for identifying any 
new and significant information regarding the NRC's conclusions in the 
early site permit EIS.
    The NRC's regulations and the applicable case law interpreting the 
National Environment Policy Act of 1969, as amended (NEPA), support the 
NRC staff's belief 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 early site permit stage and the 
NRC staff intends to use tiering and incorporation-by-reference 
whenever it is appropriate to do so. The combined license applicant 
must address any other significant environmental issue not considered 
in any previous proceeding, such as issues deferred from the early site 
permit stage to the combined license stage (e.g., the benefits 
assessment).
    For an early site permit, the NRC prepares an EIS that resolves 
numerous issues within certain bounding conditions. These issues are 
candidates for issue preclusion at the combined license, CP or OL 
stage. If the issue could be deferred and the combined license 
applicant elected to do so, e.g., the benefits assessment, then the 
combined license applicant would be required to address the issue in 
its combined license, CP, or OL application. A combined license, CP, or 
OL application must also demonstrate that the design of the facility 
falls within the parameters specified in the early site permit. In 
addition, the application should indicate whether the site is in 
compliance with the terms of the early site permit. The information 
supporting a conclusion that the site is in compliance with the early 
site permit should be maintained in an auditable form by the applicant. 
While the NRC is ultimately responsible for completing any required 
NEPA review, for example, to ensure that the conclusions for a resolved 
early site permit environmental issue remain valid for a combined 
license action, 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 for issues for which the combined license 
applicant does not identify any new and significant information.
    Under 10 CFR 51.70(b), the NRC is required to independently 
evaluate and be responsible for the reliability of all information used 
in the EIS, including an EIS prepared for a combined license. In 
carrying out its responsibilities under 10 CFR 51.70(b), the NRC staff 
may (1) inquire into the continued validity of information disclosed in 
an EIS for an early site permit that is referenced in a combined 
license application; and (2) look for any new information that may 
affect the assumptions, analysis, or conclusions reached in the early 
site permit EIS.
    The initial burden to assess newly identified information and those 
issues that were deferred to the combined license, CP, or OL 
application falls to the applicant. The applicant is required to 
provide information sufficient to resolve any other significant 
environmental issue not considered in the early site permit proceeding, 
either for the site or design, 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). Therefore, the 
environmental report must contain new and significant information on 
the site or design to the extent that it differs from, or is in 
addition to, that discussed in the early site permit EIS.
    The NRC staff, in the context of a combined license application 
that references an early site permit, defines ``new'' in the phrase 
``new and significant information'' as any information that was not 
contained or referenced in the early site permit application or the 
early site permit EIS. This new information may include (but is not 
limited to) specific design information that was not contained in the 
application, especially where the design interacts with the 
environment, or information that was in the early site permit 
application, but has changed by the time of the combined license 
application. This new information may or may not be significant.
    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 staff 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

[[Page 12827]]

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.
    In conclusion, the NRC staff 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. For matters resolved at the ESP stage, 
if there is no new and significant information that differs from that 
discussed in the ESP EIS, then the staff will rely upon (``tier off'') 
the early site permit EIS and disclose the NRC conclusion for matters 
covered in the early site permit review. Such matters will not be 
subject to litigation at the combined license stage.
    Section 51.51, Uranium fuel cycle environmental data--Table S-3. 
Section 51.51 would be revised 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.
    Section 51.52, Environmental effects of transportation of fuel and 
waste--Table S-4. Section 51.52 would be amended 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.
    Section 51.53, Postconstruction environmental reports. Section 
51.53(a) would be revised 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. Section 51.53(a) also would be revised 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, Sec.  51.53(c)(3) would be revised 
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.
    Section 51.54, Environmental report--manufacturing license. The 
proposed rule would amend 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.
    Section 51.55, Environmental report--standard design certification. 
The provisions in current Sec.  51.55 would be transferred to a new 
Sec.  51.58 (discussed in Sec.  51.58), and this section would be 
revised to address the contents of environmental reports for design 
certifications under subpart B of part 52. The structure of proposed 
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.
    Section 51.58, Environmental report--number of copies; 
distribution. The matters previously addressed in Sec.  51.55 would be 
addressed in a proposed new Sec.  51.58. Section 51.58(a) would add 
conforming references for early site permits and combined licenses. 
Section

[[Page 12828]]

51.58(b) would make a conforming reference to subpart F of part 52.
    Section 51.71, Draft environmental impact statement--contents. 
Section 51.71(d) and its associated Footnote 3 would be revised to 
include a separate discussion with specific requirements for the 
content of draft environmental impact statements at the early site 
permit and combined license stages.
    Section 51.75, Draft environmental impact statement--construction 
permit, early site permit, or combined license. Sections 51.75(b) and 
(c) and a new Footnote 5 would be added to include separate 
requirements for the preparation of draft EISs at the early site permit 
and combined license stages. Section 51.75(c) would be organized into 
separate subparagraphs, which would 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 both, or proposes to use a manufactured reactor. For 
example, Sec.  51.75(c)(3) would provide that the combined license EIS 
will not address the environmental impacts associated with 
manufacturing the reactor under the manufacturing license.
    Section 51.95, Postconstruction environmental impact statements. 
Section 51.95(a) would be revised 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, Sec.  51.95(c) would be revised to 
correct the address for the NRC Public Document Room. Section 51.95 
would be revised 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, Sec.  51.95(d) would be 
revised 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.
    Section 51.105, Public hearings in proceedings for issuance of 
construction permits or early site permits. The section heading and 
Sec.  51.105(a) would be revised 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, Sec.  
51.105(b) would be added 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. In accordance with Sec.  52.17, applicants are not 
required to address the benefits assessment (e.g., need for power) or 
alternative energy sources at the early site permit stage.
    Section 51.105a, Public hearings in proceedings for issuance of 
manufacturing licenses. Section 51.105a would be added to provide 
requirements for public hearings in proceedings for issuance of 
manufacturing licenses. Specifically, Sec.  51.105a would establish 
that the presiding officer in a proceeding for the issuance of a 
manufacturing license will (1) Determine, in an uncontested proceeding, 
whether the NEPA review conducted by the NRC staff has been adequate to 
identify all reasonable SAMDAs for the design of the reactor to be 
manufactured, and evaluate the environmental, technical, economic, and 
other benefits and costs of each SAMDA; and (2) determine, in a 
contested proceeding, whether the manufacturing license should be 
issued as proposed by the NRC staff director (Director of Nuclear 
Reactor Regulation).
    Section 51.107, Public hearings in proceedings for issuance of 
combined licenses. Section 51.107 would be added 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.

N. Proposed 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 would be made to this section to include 
renewed combined licenses.
2. Section 54.3, Definitions
    The definition for renewed combined license would be 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) would be revised to add a conforming reference to 
combined licenses issued under 10 CFR part 52.
4. Section 54.27, Hearings
    This section would be 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) would be revised to include 
conforming references to combined licenses in this procedure on 
issuance of renewed licenses.
6. Section 54.35, Requirements During Term of Renewed License
    This section would be 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) would be revised to include a conforming reference 
to a renewed combined license.

O. Proposed 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. Currently, the provisions in 
part 55 refer 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 would be revised by adding a 
reference to part 52. This would clarify 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 would clarify that these 
operators, as well as holders of combined licenses issued under part 52

[[Page 12829]]

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. Proposed 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) would be revised to include a conforming 
reference to combined licenses issued under part 52.

Q. Proposed 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. Currently, 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, do 
not refer to combined licenses issued under part 52. However, part 73 
is currently 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 would be revised to clarify that the 
regulations in part 73 apply to persons who receive combined licenses 
under part 52, and Sec.  73.2 would be revised to state that terms 
defined in part 52 have the same meaning when used in part 73. The NRC 
proposes to address 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 authorizes 
fuel load and operation under Sec.  52.103 for a combined license. 
Additional conforming changes to include part 52 licenses are proposed 
for Sec. Sec.  73.50 and 73.56, and Appendix C to part 73.

R. Proposed 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) would be 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. Proposed Changes to 10 CFR Part 95

    The following discussion explains the requirements in part 95 
generically and covers Sections 95.5, 95.13, 95.19, 95.20, 95.23, 
95.31, 95.33-95.37, 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 proposes to revise 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 
proposes to revise Sec.  95.39 to use the phrase, ``NRC license, 
certificate, or standard design approval or standard design 
certification under part 52.''

T. Proposed 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 currently does not address the indemnification and financial 
protection requirements of combined license holders. Accordingly, 
various sections in part 140 are being revised 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 authorizes the possession or use of nuclear fuel or 
other nuclear materials.

[[Page 12830]]

U. Proposed Changes to 10 CFR Part 170

    Part 170 sets out the fees charged for licensing services performed 
by the NRC. Sections 170.2(g) and (k) would be revised to add 
conforming references to manufacturing licenses and standard design 
approvals issued under part 52, remove the reference to Appendix Q that 
will be returned to part 50, and delete the reference to a 
manufacturing license issued under part 50 (which is proposed to be 
removed from part 50 because of its transfer to part 52 in the 1989 
rulemaking adopting part 52).

V. Specific Request for Comments

    In addition to the general invitation to submit comments on the 
proposed rule, the NRC also requests comments on the following 
questions:
    1. 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 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.
    2. Currently, 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 the concept of 
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: (i) Address one or more of the sixteen standards in 
Sec.  50.47(b), and (ii) 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, 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 ITAAC, at the 
early site permit stage.
    3. As indicated in Section IV, Discussion of Substantive Changes, 
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 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.
    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 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.

[[Page 12831]]

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.
    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 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.
    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 requirement were adopted.
    7. As discussed in Section IV.C.6.f of this 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 
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 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.
    8. 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

[[Page 12832]]

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.
    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 EIS which require supplementation.\10\ 
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 emergency preparedness 
(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 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.
---------------------------------------------------------------------------

    \10\ 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 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

[[Page 12833]]

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 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 
above, 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

[[Page 12834]]

an ESP updating process as described above, 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.
    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 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.
    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-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 at 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-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-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-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

[[Page 12835]]

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 the 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 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) 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

[[Page 12836]]

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 Section 52.99(e) and the Commission's ITAAC finding under 
proposed Section 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.
    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 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.
    13. As discussed in Section IV.F.6 of this statement of 
considerations, the Commission proposes in this rulemaking, 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

[[Page 12837]]

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.
    14. As discussed in Section IV.C.5.g of this SOC, 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).
    15. In Section IV.J of the Supplementary Information of this 
Federal Register Notice, the NRC 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 this 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 this 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.

VI. 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 www.nrc.gov/reading-rm.html.
    The NRC staff contact. Nanette V. Gilles, Mail Stop O-4D9A, 
Washington, DC 20555, 301-415-1180.

[[Page 12838]]



----------------------------------------------------------------------------------------------------------------
            Document                     PDR                Web                  EPDR              NRC staff
----------------------------------------------------------------------------------------------------------------
Comments received...............  X................  X................  X....................  .................
Regulatory Analysis.............  X................  X................  ML...................  X
Regulatory History Index for      .................  .................  ML032810026..........  .................
 July 2003 proposed rule.
----------------------------------------------------------------------------------------------------------------

VII. 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 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 are those 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 are 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 are those 
program elements that address areas of regulation that cannot be 
relinquished to Agreement States pursuant to the Atomic Energy Act, as 
amended, or provisions of Title 10 of the Code of Federal Regulations 
and should not be adopted by Agreement States. Category H&S are 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. The proposed revisions are categorized as follows:

                               List of Changes 10 CFR Part 52 Proposed Rulemaking
----------------------------------------------------------------------------------------------------------------
                                                                                              Comments regarding
         Proposed sections            Description--new,        Compatibility designation         compatibility
                                           changes                                                designation
----------------------------------------------------------------------------------------------------------------
                 10 CFR Part 2--Rules of Practice for Domestic Licensing and Issuance of Orders
----------------------------------------------------------------------------------------------------------------
2.1...............................  Scope...............  [D]...............................  Agreement States
                                                                                               may adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.4...............................  Definitions.........
                                    Contested             [D]...............................  Agreement States
                                     proceedings.                                              may adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
                                    License.............  [D]...............................  Agreement States
                                                                                               adopt similar
                                                                                               definition as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs. This
                                                                                               definition
                                                                                               appears in 10 CFR
                                                                                               Sec.   20.1003.
                                                                                               For purposes of
                                                                                               compatibility,
                                                                                               Agreement States
                                                                                               should use the
                                                                                               language of the
                                                                                               Part 20
                                                                                               definition, which
                                                                                               is assigned a
                                                                                               Compatibility
                                                                                               Category D.
                                    Licensee............  [D]...............................  Agreement States
                                                                                               adopt a similar
                                                                                               definition as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs. This
                                                                                               definition
                                                                                               appears in 10 CFR
                                                                                               Sec.   20.1003.
                                                                                               For purposes of
                                                                                               compatibility,
                                                                                               Agreement States
                                                                                               should use the
                                                                                               language of the
                                                                                               Part 20
                                                                                               definition, which
                                                                                               is assigned a
                                                                                               Compatibility
                                                                                               Category D.
Subpart A
 2.100............................  Scope of parts......  [D]...............................  Agreement States
                                                                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.101.............................  Filing of             [D]...............................  Agreement States
                                     application.                                              adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.

[[Page 12839]]

 
2.102.............................  Administrative        [D]...............................  Agreement States
                                     review of                                                 adopt similar
                                     application.                                              provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
                                                                                               These similar
                                                                                               provisions
                                                                                               appears in 10 CFR
                                                                                               Sec.   30. For
                                                                                               purposes of
                                                                                               compatibility,
                                                                                               Agreement States
                                                                                               should use the
                                                                                               language in Part
                                                                                               30, which is
                                                                                               assigned a
                                                                                               Compatibility
                                                                                               Category D.
2.104.............................  Notice of hearing...  [D]...............................  Agreement States
                                                                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.105.............................  Notice of proposed    [D]...............................  Agreement States
                                     action.                                                   adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.106.............................  Notice of issuances.  [D]...............................  Agreement States
                                     Added notice for                                          adopt similar
                                     COL in FR.                                                provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.109.............................  Effect of timely      [D]...............................  Agreement States
                                     renewal application.                                      adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
                                                                                               These similar
                                                                                               provisions
                                                                                               appears in 10 CFR
                                                                                               Sec.   30. For
                                                                                               purposes of
                                                                                               compatibility,
                                                                                               Agreement States
                                                                                               should use the
                                                                                               language in Part
                                                                                               30, which is
                                                                                               assigned a
                                                                                               Compatibility
                                                                                               Category D.
2.110.............................  Filing and            [D]...............................  Agreement States
                                     administrative                                            adopt similar
                                     action on submittal                                       provisions as a
                                     for design review                                         part of their
                                     of site suitability.                                      regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.111.............................  Prohibition of sex    [D]...............................  Agreement States
                                     discrimination.                                           may adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
Subpart B
2.200.............................  Scope of subpart....  [D]...............................  Agreement States
                                                                                               may adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.202.............................  Orders..............  [D]...............................  Agreement States
                                                                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
Subpart C
2.390.............................  Public inspections,   [D]...............................  Agreement States
                                     exemptions,                                               adopt similar
                                     requests for                                              provisions as a
                                     withholding.                                              part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
Subpart E
2.500.............................  Scope of subpart....  NRC...............................  This provision is
                                                                                               designated a
                                                                                               Compatibility
                                                                                               Category NRC
                                                                                               because it
                                                                                               addresses
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
2.501.............................  Notice of hearing on  NRC...............................  This provision is
                                     application for                                           designated a
                                     license to                                                Compatibility
                                     manufacture nuclear                                       Category NRC
                                     power plants.                                             because it
                                                                                               addresses
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.

[[Page 12840]]

 
2.502.............................  Notice of hearing on  NRC...............................  This provision is
                                     application for a                                         designated a
                                     construction permit                                       Compatibility
                                     for a nuclear power                                       Category NRC
                                     reactor                                                   because it
                                     manufactured at the                                       addresses
                                     site at which the                                         activities
                                     reactor is to be                                          reserved to the
                                     operated.                                                 Commission.
2.503.............................  Finality of           NRC...............................  This provision is
                                     decisions on                                              designated a
                                     separate issues.                                          Compatibility
                                                                                               Category NRC
                                                                                               because it
                                                                                               addresses
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
2.504.............................  Applicability of      NRC...............................  This provision is
                                     other sections.                                           designated a
                                                                                               Compatibility
                                                                                               Category NRC
                                                                                               because it
                                                                                               addresses
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
Subpart H
2.800.............................  Scope of rulemaking.  [D]...............................  Agreement States
                                                                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.801.............................  Initiation of         [D]...............................  Agreement States
                                     rulemaking.                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.811.............................  Filing of standard    [D]...............................  Agreement States
                                     design                                                    adopt similar
                                     certification                                             provisions as a
                                     application                                               part of their
                                     required copies.                                          regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.813.............................  Written               [D]...............................  Agreement States
                                     communications.                                           adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.815.............................  Docketing and         [D]...............................  Agreement States
                                     acceptance review.                                        adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.817.............................  Withdrawal of         [D]...............................  Agreement States
                                     application.                                              adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
2.819.............................  Denial of             [D]...............................  Agreement States
                                     application for                                           adopt similar
                                     failure to supply                                         provisions as a
                                     information.                                              part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
----------------------------------------------------------------------------------------------------------------
  10 CFR Part 10--Criteria and Procedures for Determining Eligibility for Access to Restricted Data or National
                                 Security Information or an Employment Clearance
----------------------------------------------------------------------------------------------------------------
10.1..............................  Purpose.............  NRC...............................  This provision is
                                                                                               designated a
                                                                                               Compatibility
                                                                                               Category NRC
                                                                                               because it
                                                                                               addresses
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10.2..............................  Scope...............  NRC...............................  This provision is
                                                                                               designated a
                                                                                               Compatibility
                                                                                               Category NRC
                                                                                               because it
                                                                                               addresses
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
----------------------------------------------------------------------------------------------------------------
           10 CFR Part 19--Notices, Instructions and Reports to Workers; Inspection and Investigations
----------------------------------------------------------------------------------------------------------------
19.1..............................  Purpose.............  D.................................  Agreement States
                                                                                               may adopt similar
                                                                                               provisions
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
19.2..............................  Scope...............  D.................................  Agreement States
                                                                                               may adopt similar
                                                                                               provisions
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
19.3..............................  Definitions.........

[[Page 12841]]

 
                                    Regulated activities  D.................................  Agreement States
                                                                                               may adopt a
                                                                                               similar
                                                                                               definition
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
                                    Regulated entities..  D.................................  Agreement States
                                                                                               may adopt a
                                                                                               similar
                                                                                               definition
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
                                    Worker..............  C.................................  This provision is
                                                                                               currently
                                                                                               designated a
                                                                                               Compatibility
                                                                                               Category C.
                                                                                               However, since
                                                                                               the proposed
                                                                                               revisions address
                                                                                               areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction,
                                                                                               Agreement States
                                                                                               should not adopt
                                                                                               these amendments.
19.11.............................  Posting of notices    C.................................  This provision is
                                     to workers.                                               currently
                                                                                               designated a
                                                                                               Compatibility
                                                                                               Category C.
                                                                                               However, since
                                                                                               the proposed
                                                                                               revisions address
                                                                                               areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction,
                                                                                               Agreement States
                                                                                               should not adopt
                                                                                               these amendments.
19.14.............................  Presence of           C.................................  This provision is
                                     representatives of                                        currently
                                     licensees and                                             designated a
                                     workers during                                            Compatibility
                                     inspections.                                              Category C.
                                                                                               However, since
                                                                                               the proposed
                                                                                               revisions address
                                                                                               areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction,
                                                                                               Agreement States
                                                                                               should not adopt
                                                                                               these amendments.
19.20.............................  Employee protection.  D.................................  Agreement States
                                                                                               may adopt similar
                                                                                               provisions
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
19.31.............................  Application for       D.................................  Agreement States
                                     exemptions.                                               may adopt similar
                                                                                               provisions
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
19.32.............................  Discrimination        D.................................  Agreement States
                                     prohibited.                                               may adopt similar
                                                                                               provisions
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
----------------------------------------------------------------------------------------------------------------
                                     10 CFR Part 20--Standards of Protection
----------------------------------------------------------------------------------------------------------------
20.1002...........................  Scope...............  D.................................  Agreement States
                                                                                               may adopt similar
                                                                                               provisions
                                                                                               consistent with
                                                                                               their regulatory
                                                                                               authority, but
                                                                                               should not
                                                                                               address areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction.
20.1401...........................  General provisions    C.................................  This provision is
                                     and scope.                                                currently
                                                                                               designated a
                                                                                               Compatibility
                                                                                               Category C.
                                                                                               However, since
                                                                                               the proposed
                                                                                               revisions address
                                                                                               areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction,
                                                                                               Agreement States
                                                                                               should not adopt
                                                                                               these amendments.
20.2203...........................  Reports of            C--paragraphs (a), (b)............  Portions of this
                                     exposures, etc.,     D--paragraph (d)..................   provision is
                                     exceeding the        NRC--paragraph (c)................   currently
                                     limits.                                                   designated a
                                                                                               Compatibility
                                                                                               Category C.
                                                                                               However, since
                                                                                               the proposed
                                                                                               revisions address
                                                                                               areas of
                                                                                               exclusive NRC
                                                                                               jurisdiction,
                                                                                               Agreement States
                                                                                               should not adopt
                                                                                               these amendments.
----------------------------------------------------------------------------------------------------------------
                             10 CFR Part 21--Reporting of Defects and Noncompliance
----------------------------------------------------------------------------------------------------------------
21.2..............................  Scope...............  N/A...............................  The provisions in
                                                                                               Part 21 are
                                                                                               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.

[[Page 12842]]

 
21.3..............................  Definitions.........  N/A...............................  The provisions in
                                                                                               Part 21 are
                                                                                               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.
21.5..............................  Communication.......  N/A...............................  The provisions in
                                                                                               Part 21 are
                                                                                               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.
21.21.............................  Notification of       N/A...............................  The provisions in
                                     failure to comply                                         Part 21 are
                                     or existence of a                                         derived from
                                     defect.                                                   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.
21.51.............................  Maintenance and       N/A...............................  The provisions in
                                     inspections of                                            Part 21 are
                                     records.                                                  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.
21.61.............................  Failure to notify...  N/A...............................  The provisions in
                                                                                               Part 21 are
                                                                                               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.
----------------------------------------------------------------------------------------------------------------

[[Page 12843]]

 
                                      10 CFR Part 25--Access Authorization
----------------------------------------------------------------------------------------------------------------
25.35.............................  Classified visits...  NRC...............................  This provision is
                                                                                               designated a
                                                                                               Compatibility
                                                                                               Category NRC
                                                                                               because it
                                                                                               addresses
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
----------------------------------------------------------------------------------------------------------------
                                    10 CFR Part 26--Fitness for Duty Programs
----------------------------------------------------------------------------------------------------------------
26.2..............................  Scope...............  [D]...............................  Agreement States
                                                                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
26.10.............................  General performance   [D]...............................  Agreement States
                                     objectives.                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
10 CFR Part 50....................  Domestic licensing    NRC for all sections..............  These provisions
                                     of production and                                         are designated a
                                     utilization                                               Compatibility
                                     facilities.                                               Category NRC
                                                                                               because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 51....................  Environmental         NRC for all sections..............  These provisions
                                     protection                                                are designated a
                                     regulation for                                            Compatibility
                                     domestic licensing                                        Category NRC
                                     and related                                               because they
                                     regulatory                                                address
                                     functions.                                                activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 52....................  Licenses,             NRC for all sections..............  These provisions
                                     certifications, and                                       are designated a
                                     approvals for                                             Compatibility
                                     nuclear power                                             Category NRC
                                     plants.                                                   because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 54....................  Requirements for      NRC for all sections..............  These provisions
                                     renewal of                                                are designated a
                                     operating licenses                                        Compatibility
                                     for nuclear power                                         Category NRC
                                     plants.                                                   because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 55....................  Operators' licenses.  NRC for all sections..............  These provisions
                                                                                               are designated a
                                                                                               Compatibility
                                                                                               Category NRC
                                                                                               because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 72....................  Licensing             NRC for all sections..............  These provisions
                                     requirements for                                          are designated a
                                     ISFSI, HLW, and                                           Compatibility
                                     greater than class                                        Category NRC
                                     C.                                                        because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 73....................  Physical protection   NRC for all sections..............  These provisions
                                     of plants and                                             are designated a
                                     materials.                                                Compatibility
                                                                                               Category NRC
                                                                                               because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 75....................  Safeguards on         NRC for all sections..............  These provisions
                                     nuclear material.                                         are designated a
                                                                                               Compatibility
                                                                                               Category NRC
                                                                                               because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 95....................  Facility security     NRC for all sections..............  These provisions
                                     clearance and                                             are designated a
                                     safeguarding of                                           Compatibility
                                     national security                                         Category NRC
                                     information and                                           because they
                                     restricted data.                                          address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 140...................  Financial protection  NRC for all sections..............  These provisions
                                     requirements and                                          are designated a
                                     indemnity                                                 Compatibility
                                     agreements.                                               Category NRC
                                                                                               because they
                                                                                               address
                                                                                               activities
                                                                                               reserved to the
                                                                                               Commission.
10 CFR Part 170...................  Annual fees.........  [D]...............................  Agreement States
                                                                                               adopt similar
                                                                                               provisions as a
                                                                                               part of their
                                                                                               regulatory
                                                                                               programs through
                                                                                               a mechanism that
                                                                                               is appropriate
                                                                                               under the State's
                                                                                               laws, but should
                                                                                               not address areas
                                                                                               of exclusive NRC
                                                                                               jurisdiction.
----------------------------------------------------------------------------------------------------------------

VIII. Plain Language

    The Presidential memorandum dated June 1, 1998, entitled ``Plain 
Language in Government Writing'' directed that the Government's writing 
be in plain language. This memorandum was published on June 10, 1998 
(63 FR 31883). In complying with this directive, the NRC made editorial 
changes to improve the organization and readability of the existing 
language of the paragraphs being revised. These types of changes are 
not discussed further in this document. The NRC requests comments on 
the proposed rule specifically with respect to the clarity and 
effectiveness of the language used. Comments should be submitted using 
one of the methods detailed under the ADDRESSES heading of the preamble 
to this proposed rule.

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 proposing to revise 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

[[Page 12844]]

certification reviews and on discussions with stakeholders on these 
licensing processes. This rulemaking does not establish standards or 
substantive the requirements with which all applicants and licensees 
must comply. In addition, this rule would amend 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 
would not constitute the establishment of a standard that contains 
generally applicable requirements.

X. Environmental Impact--Categorical Exclusion

    The NRC has determined that the changes made in this rule fall 
within the types of actions described in 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.\11\
---------------------------------------------------------------------------

    \11\ 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 2 is consistent with the original 
categorical exclusion determination. To ensure that future changes 
in part 52 are categorically excluded, the proposed rule contains an 
appropriate change to Sec.  51.22(c)(3).
---------------------------------------------------------------------------

XI. Paperwork Reduction Act Statement

    This proposed rule contains new or amended information collection 
requirements contained in 10 CFR parts 21, 25, 50, 52, and 54 that are 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). These information collection requirements have been submitted to 
the Office of Management and Budget for review and approval. The 
proposed changes to 10 CFR parts 19, 20, 26, 51, 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, 3150-0021, 3150-0018, 3150-0132, 3150-0002, 3150-0055, 3150-0047, 
and 3150-0039.
    Type of submission, new or revision: New.
    The title of the information collection: 10 CFR part 52 and 
Conforming Amendments to Parts 1, 2, 10, 19, 20, 21, 25, 26, 50, 51, 
54, 55, 72, 73, 75, 95, 140, and 170, ``Licenses, Certifications, and 
Approvals for Nuclear Power Plants,'' Revised Proposed Rule.
    The form number if applicable: N/A.
    How often the collection is required: On occasion and every 10 to 
20 years for applications for renewal.
    Who will be required or asked to report: Designers and 
manufacturers of commercial nuclear power plants, electric power 
companies, and any person eligible under the Atomic Energy Act to apply 
for a construction permit for a nuclear power plant.
    An estimate of the number of annual responses: 20.333.
    The estimated number of annual respondents: 4.33.
    An estimate of the total number of hours needed annually to 
complete the requirement or request: 452,416 (448,946 hours reporting 
and 3470 hours recordkeeping).
    Abstract: 10 CFR part 52 establishes requirements for the granting 
of early site permits, approvals and certifications of standard nuclear 
power plant designs, licenses which combine in a single license a 
construction permit and an operating license with conditions (combined 
licenses), and manufacturing licenses. Part 52 also establishes 
requirements for renewal of those approvals, permits, certifications, 
and licenses; amendments to them; and exemptions or variances from 
them.
    NRC uses the information collected to assess the adequacy and 
suitability of an applicant's site, plant design, training and 
experience, and plans and procedures for the protection of public 
health and safety. The NRC review of such information and the findings 
derived from that information form the basis of NRC decisions and 
actions concerning the issuance, modification, or revocation of site 
permits, design approvals and certifications, combined licenses, and 
manufacturing licenses for nuclear power plants.
    The U.S. Nuclear Regulatory Commission is seeking public comment on 
the potential impact of the information collections contained in this 
proposed rule (or proposed policy statement) and on the following 
issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Is the estimate of burden accurate?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected?
    4. How can the burden of the information collection be minimized, 
including the use of automated collection techniques?
    A copy of the OMB clearance package may be viewed free of charge at 
the NRC Public Document Room, One White Flint North, 11555 Rockville 
Pike, Room O-1 F21, Rockville, Maryland 20852. The OMB clearance 
package and rule are available at the NRC worldwide Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html for 60 days after 
the signature date of this notice and are also available at the rule 
forum site, http://ruleforum.llnl.gov.
    Send comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden and on the 
above issues, by April 12, 2006 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, John A. Asalone, Office 
of Information and Regulatory Affairs, NEOB-10202, (3150-0151), Office 
of Management and Budget, Washington, DC 20503. Comments received after 
this date will be considered if it is practical to do so, but assurance 
of consideration cannot be given to comments received after this date. 
You may also e-mail comments to [email protected] or 
comment by telephone at (202) 395-4650.

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 draft regulatory analysis on this 
proposed regulation. The analysis examines the costs and benefits of 
the alternatives considered by the Commission. The draft analysis can 
be viewed in NRC's ADAMS system, Accession Number ML052840320. The 
Commission

[[Page 12845]]

requests public comment on the draft regulatory analysis. Comments on 
the draft analysis may be submitted to the NRC as indicated under the 
ADDRESSES heading.

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, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This proposed 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 proposed 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 
proposed rule and, therefore, a backfit analysis is not required, 
because the proposed 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 early site permits, combined 
licenses, or manufacturing licenses that would be protected by the 
backfitting restrictions in Sec.  50.109. To the extent that the 
proposed rule would revise the requirements for future early site 
permits, standard design certifications, combined licenses, standard 
design approvals and manufacturing licenses for nuclear power plants, 
these revisions would not constitute backfits because they are 
prospective in nature and the backfit rule was not intended to apply to 
every NRC action which substantially changes the expectations of future 
applicants.
    Other provisions in the proposed rule would apply to currently-
approved standard design approvals and certifications, but these would 
not constitute backfitting 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 the proposed rule represent 
conforming changes throughout 10 CFR 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 or design certification itself. 
Accordingly, the Commission believes that these conforming changes with 
respect to design approvals and design certifications do not raise new 
backfitting considerations that must be addressed in this rulemaking.

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

[[Page 12846]]

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 
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.

    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. 553, the NRC is proposing to 
adopt 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

    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.

    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

    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 2.790 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).

    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.
    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; 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.
* * * * *
    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

    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.
    8. In Sec.  2.101, paragraphs (a)(1), (a)(2), the introductory text 
of paragraph (a)(3), paragraphs (a)(3)(ii), and paragraph (a)(4) are 
revised to read as follows:


Sec.  2.101  Filing of application.

    (a)(1) An application for a permit, license, a license transfer, a 
license amendment, a license renewal, and standard design approval, 
shall be filed with the 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

[[Page 12847]]

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 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 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 e-mail 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 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 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 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 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 
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 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.
* * * * *
    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.
* * * * *
    10. In Sec.  2.104, the introductory text of paragraph (a) is 
revised, current paragraphs (d) and (e) are redesignated as paragraphs 
(l) and (m), respectively, and revised, new paragraphs (d), (e), and 
(f) are added, and paragraphs (g) through (k) are added and reserved, 
and footnote 1 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 as required 
by law 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

[[Page 12848]]

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 an early site permit, construction permit or combined 
license for a facility of the type described in Sec.  50.22 of this 
chapter, or a testing facility, the notice (other than a notice under 
paragraph (d) of this section) shall be issued as soon as practicable 
after the application has been docketed; provided, that if the 
Commission, under Sec.  2.101(a)(2), decides to determine the 
acceptability of the application based on its technical adequacy as 
well as completeness, the notice shall be issued as soon as practicable 
after the application has been tendered. The notice will state:
---------------------------------------------------------------------------

    \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.  50.21(b) or 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.
---------------------------------------------------------------------------

* * * * *
    (d) In the case of an application for an early site permit under 
subpart A of part 52 of this chapter, the notice will, except as the 
Commission determines otherwise, state, in implementation of paragraph 
(a)(3) of this section:
    (1) If the proceeding is a contested proceeding, the presiding 
officer will consider the following issues:
    (i) Whether applicable standards and requirements of the Act and 
the Commission's regulations have been met;
    (ii) Whether any required notifications to other agencies or bodies 
have been duly made;
    (iii) If the applicant requests authorization to perform the 
activities under Sec.  52.17(c) of this chapter, whether there is 
reasonable assurance that the proposed site is a suitable location for 
a reactor of the general size and type described in the application 
from the standpoint of radiological health and safety considerations 
under the Act and regulations issued by the Commission.
    (iv) Whether there is reasonable assurance that the site is in 
conformity with the provisions of the Act, and the Commission's 
regulations;
    (v) Whether the applicant is technically qualified to engage in any 
activities authorized;
    (vi) Whether 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;
    (vii) Whether issuance of the early site permit will be inimical to 
the common defense and security or to the health and safety of the 
public; and
    (viii) Whether, in accordance with the requirements of subpart A of 
part 52 of this chapter and subpart A of part 51 of this chapter, the 
early site permit should be issued as proposed.
    (2) If the proceeding is not a contested proceeding, the presiding 
officer will determine, without conducting a de novo evaluation of the 
application, whether:
    (i) The application and the record of the proceeding contain 
sufficient information, and the review of the application by the NRC 
staff has been adequate to support affirmative findings on paragraphs 
(d)(1)(i) through (v), and (vii) of this section, and a negative 
finding on paragraph (d)(1)(vi) of this section; and
    (ii) The review conducted under part 51 of this chapter under the 
National Environmental Policy Act (NEPA) has been adequate.
    (3) Regardless of whether the proceeding is contested or 
uncontested, the presiding officer will, in accordance with subpart A 
of part 51 of this chapter:
    (i) Determine whether the requirements of section 102(2) (A), (C), 
and (E) of the NEPA and subpart A of part 51 of this chapter have been 
complied with in the proceeding;
    (ii) Independently consider the final balance among conflicting 
factors contained in the record of the proceeding with a view to 
determine the appropriate action to be taken; and
    (iii) If the applicant requests authorization to perform the 
activities under Sec.  52.17(c) of this chapter, whether there is 
reasonable assurance that the proposed site is a suitable location for 
a reactor of the general size and type described in the application 
from the standpoint of radiological health and safety considerations 
under the Act and regulations issued by the Commission.
    (iv) Determine whether the combined license should be issued, 
denied or appropriately conditioned to protect environmental values.
    (e) In the case of an application for a combined license under 
subpart C of part 52 of this chapter, the notice will, except as the 
Commission determines otherwise, state, in implementation of paragraph 
(a)(3) of this section:
    (1) If the proceeding is a contested proceeding, the presiding 
officer will consider the following issues:
    (i) Whether applicable standards and requirements of the Act and 
the Commission's regulations have been met;
    (ii) Whether any required notifications to other agencies or bodies 
have been duly made;
    (iii) Whether 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) Whether the applicant is technically and financially qualified 
to engage in the activities authorized;
    (v) Whether issuance of the license will not be inimical to the 
common defense and security or to the health and safety of the public.
    (vi) Whether the proposed inspections, tests, analyses, and 
acceptance criteria, including those applicable to emergency planning, 
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 
regulations;
    (vii) Whether any inspections, tests, or analyses have been 
successfully completed and the acceptance criteria in a referenced 
early site permit, standard design certification or for a manufactured 
reactor have been met, but only to the extent that the combined license 
application represents that those inspections, tests and analyses have 
been successfully completed and the acceptance criteria have been met;
    (viii) Whether the issuance of the combined license will be 
inimical to the common defense and security or to the health and safety 
of the public; and
    (ix) Whether, in accordance with the requirements of subpart C of 
part 52 of this chapter and subpart A of part 51 of this chapter, the 
combined license should be issued as proposed.
    (2) If the proceeding is not a contested proceeding, the presiding 
officer will determine, without conducting a de novo evaluation of the 
application, if:
    (i) The application and the record of the proceeding contain 
sufficient information, and the review of the application by the NRC 
staff has been adequate to support affirmative findings on paragraphs 
(e)(1)(i) through (vii), and (ix) of this section, and a negative 
finding on paragraph (e)(1)(viii) of this section; and
    (ii) The review conducted under part 51 of this chapter under NEPA 
has been adequate.

[[Page 12849]]

    (3) Regardless of whether the proceeding is contested or 
uncontested, the presiding officer will, in accordance with subpart A 
of part 51 of this chapter:
    (i) Determine whether the requirements of section 102(2) (A), (C), 
and (E) of the NEPA and subpart A of part 51 of this chapter have been 
complied with in the proceeding;
    (ii) Independently consider the final balance among conflicting 
factors contained in the record of the proceeding with a view to 
determine the appropriate action to be taken; and
    (iii) Determine whether the combined license should be issued, 
denied or appropriately conditioned to protect environmental values.
    (f) In the case of an application for a manufacturing license under 
subpart F of part 52 of this chapter, the issues stated in the notice 
of hearing under paragraph (a)(3) of this section will not involve 
consideration of the particular sites at which any of the nuclear power 
reactors to be manufactured may be located and operated. Except as the 
Commission determines otherwise, the notice of hearing will state:
    (1) If the proceeding is a contested proceeding, the presiding 
officer will consider the following issues:
    (i) Whether applicable standards and requirements of the Act and 
the Commission's regulations have been met;
    (ii) Whether 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 provisions of the Act, and the Commission's regulations;
    (iii) Whether the proposed reactor(s) to be manufactured can be 
incorporated into a nuclear power plant 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;
    (iv) Whether the applicant is technically qualified to design and 
manufacture the proposed nuclear power reactor(s);
    (v) Whether 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 
reactor has been manufactured and will be operated in conformity with 
the license, the provisions of the Act, and the Commission's 
regulations;
    (vi) Whether the issuance of a license for manufacture of the 
reactor(s) will be inimical to the common defense and security or to 
the health and safety of the public; and
    (vii) Whether, in accordance with the requirements of subpart F of 
part 52 and subpart A of part 51 of this chapter, the license should be 
issued as proposed.
    (2) If the proceeding is not a contested proceeding, the presiding 
officer will determine, without conducting a de novo evaluation of the 
application, whether:
    (i) The application and the record of the proceeding contain 
sufficient information, and the review of the application by the NRC 
staff has been adequate to support affirmative findings on paragraphs 
(f)(1)(i) through (v), and (vii) of this section proposed to be made 
and a negative finding on paragraph (f)(1)(vi) of this section; and
    (ii) The review conducted under part 51 of this chapter under NEPA 
has been adequate.
    (3) Regardless of whether the proceeding is contested or 
uncontested, the presiding officer will, in accordance with subpart A 
of part 51:
    (i) Determine whether the requirements of section 102(2) (A), (C), 
and (E) of the National Environmental Policy Act and subpart A of part 
51 of this chapter have been complied with in the proceeding;
    (ii) Independently consider the final balance among conflicting 
factors contained in the record of the proceeding with a view to 
determine the appropriate action to be taken; and
    (iii) Determine whether the manufacturing license should be issued, 
denied or appropriately conditioned to protect environmental values.
    (4) The place of hearing on an application for a manufacturing 
license will be Rockville, Maryland, or such other location as the 
Commission deems appropriate.
    (g)-(k) [Reserved]
    (l) In an application for a construction permit or an operating 
license for a facility 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 to consider the antitrust aspects of 
the application, the notice of hearing will, unless the Commission 
determines otherwise, state:
    (1) A time of the hearing, which will be as soon as practicable 
after the receipt of the Attorney General's advice and compliance with 
sections 105 and 189a of the Act and this part;
    (2) The presiding officer for the hearing who shall be either an 
administrative law judge or an atomic safety and licensing board 
established by the Commission or by the Chief Administrative Judge of 
the Atomic Safety and Licensing Board Panel;
    (3) That the presiding officer will consider and decide whether the 
activities under the proposed license would create or maintain a 
situation inconsistent with the antitrust laws described in section 
105a of the Act; and
    (4) That matters of radiological health and safety and common 
defense and security, and matters raised under NEPA, will be considered 
at another hearing if otherwise required or ordered to be held, for 
which a notice will be published under paragraphs (a) and (b) of this 
section, unless otherwise authorized by the Commission.
    (m)(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 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 a 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 opportunity for hearing 
under Sec.  52.103 of this chapter on whether the facility as 
constructed complies, or on completion will comply, with the acceptance 
criteria in the combined license, except for those ITAAC that the 
Commission found were met under Sec.  52.97, 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,

[[Page 12850]]

if it is to be located or conducted within an Indian reservation).
    (3) 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.
    11. In Sec.  2.105, the introductory text of paragraphs (a) and 
(a)(4) are revised, and paragraphs (a)(12) 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, a document under Sec.  52.103(a) of this chapter with 
respect to a finding that inspections, tests, analyses, and acceptance 
criteria for a combined license under subpart C of part 52 have been 
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).
    (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 copies of the safety analysis may be 
obtained and examined;
    (iii) A finding that the application for the license or amendment 
complies with the requirements of the Act and this chapter, including 
successful completion of all inspections, tests, analyses, and 
acceptance criteria; 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.
* * * * *
    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 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.
* * * * *
    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 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.
    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 part 50 of this chapter, a submittal for early review of 
site suitability issues under appendix Q to part 50 of this

[[Page 12851]]

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 part 50 
of this chapter, or for a standard design approval under subpart E of 
part 52 of this chapter, 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 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.
    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.
    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.
    17. In Sec.  2.340, the section heading and paragraphs (b) and (c) 
are revised, paragraph (h) is redesignated as paragraph (o), paragraph 
(a) is redesignated as paragraph (a)(1), and paragraphs (a)(2), (e), 
(h), and (i) are added, and paragraphs (j) through (n) are added and 
reserved to read as follows:


Sec.  2.340  Initial decisions; immediate effectiveness of certain 
decisions.

    (a)(1) * * *
    (2) Initial decisions on findings 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 acceptance criteria being met, 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 matters which have 
been determined to be the issues in the proceeding by the Commission or 
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).
    (b) Immediate effectiveness of certain decisions. Except as 
provided in paragraphs (d) through (i) of this section, or as otherwise 
ordered by the Commission in special circumstances, an initial decision 
directing the issuance or amendment of an early site permit, a 
construction permit, a construction authorization, an operating 
license, a combined license under part 52 of this chapter, or a license 
under 10 CFR part 72 to store spent fuel in an independent spent fuel 
storage installation (ISFSI) at a reactor site, or a decision making 
the finding under Sec.  52.103(g) that acceptance criteria have been 
met, is effective immediately 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. If any decision under 
this paragraph is not made by the Commission acting as the presiding 
officer, the decision is subject to review and further decision by the 
Commission upon petition for review filed by any party under Sec.  
2.341 or upon its own motion.
    (c) Except as provided in paragraphs (d) through (i) of this 
section, or as otherwise ordered by the Commission in special 
circumstances, the Director of Nuclear Reactor Regulation or Director 
of Nuclear Material Safety and Safeguards, as appropriate, 
notwithstanding the filing or granting of a petition for review, shall 
issue an early site permit, a construction permit, a construction 
authorization, an operating license, a combined license under part 52 
of this chapter, or a license under 10 CFR part 72 to store spent fuel 
in an independent spent fuel storage installation at a reactor site, or 
amendments thereto, authorized by an initial decision, within ten (10) 
days from the date of issuance of the decision.
* * * * *
    (e) Nuclear power reactor early site permits. (1) Presiding 
officers. Presiding officers shall hear and decide all issues that come 
before them, indicating in their decisions the type of licensing

[[Page 12852]]

action, if any, which their decision would authorize. The presiding 
officer's decisions concerning early site permits are not effective 
until the Commission actions outlined in paragraph (e)(2) of this 
section have taken place.
    (2) Commission. Within sixty (60) days of the service of any 
presiding officer decision that would otherwise authorize issuance of 
an early site permit, the Commission will seek to issue a decision on 
any stay motions that are timely filed. These motions must be filed as 
provided by Sec.  2.341. For the purpose of this paragraph, a stay 
motion is one that seeks to defer the effectiveness of a presiding 
officer decision beyond the period necessary for the Commission action 
described herein. If no stay papers are filed, the Commission will, 
within the same time period (or earlier if possible), analyze the 
record and early site permit decision below on its own motion and will 
seek to issue a decision on whether a stay is warranted. However, the 
Commission will not decide that a stay is warranted without giving the 
affected parties an opportunity to be heard. The initial decision will 
be considered stayed pending the Commission's decision. In deciding 
these stay questions, the Commission shall employ the procedures set 
out in Sec.  2.342.
* * * * *
    (h) Issuance of nuclear power reactor combined licenses under part 
52 of this chapter. (1) Presiding officers. Presiding officers shall 
hear and decide all issues that come before them, indicating in their 
decisions the type of licensing action, if any, which their decision 
would authorize. A presiding officer's decision authorizing issuance of 
a combined license is immediately effective, and the Director shall 
issue the appropriate license in accordance with paragraph (c) of this 
section.
    (2) The Commission. (i) Reserving the power to step in at an 
earlier time, the Commission will, upon receipt of the presiding 
officer's decision authorizing issuance of a combined license, review 
the matter on its own motion to determine whether to stay the 
effectiveness of the decision. A combined license decision will be 
stayed by the Commission only if it determines that it is in the public 
interest to do so, based on a consideration of the gravity of the 
substantive issue, the likelihood that it has been resolved incorrectly 
below, the degree to which correct resolution of the issue would be 
prejudiced by construction pending review, and other relevant public 
interest factors.
    (ii) The parties may file brief comments with the Commission 
pointing out matters which, in their view, pertain to the immediate 
effectiveness issue. To be considered, these comments must be received 
within ten (10) days of the presiding officer's decision. However, the 
Commission may dispense with comments by so advising the parties. An 
extensive stay will not be issued without giving the affected parties 
an opportunity to be heard.
    (iii) The Commission intends to issue a stay decision within thirty 
(30) days of receipt of the presiding officer's decision. The presiding 
officer's initial decision will be considered stayed pending the 
Commission's decision.
    (iv) In announcing a stay decision, the Commission may allow the 
proceeding to run its ordinary course or give instructions as to the 
future handling of the proceeding. Furthermore, the Commission may, in 
a particular case, determine that compliance with existing regulations 
and policies may no longer be sufficient to warrant approval of a 
license application and may alter those regulations and policies.
    (i) Findings under Sec.  52.103(g) of this chapter with respect to 
acceptance criteria in nuclear power reactor combined licenses. (1) 
Presiding officers. Presiding officers shall hear and decide all issues 
that come before them with respect to whether acceptance criteria in 
the combined license have been met, in accordance with Sec.  52.103(g) 
of this chapter. A presiding officer's decision may not become 
effective if it would otherwise allow operation at greater than five 
(5) percent of rated power until the Commission actions outlined in 
paragraph (i)(2) of this section have taken place. If a decision 
otherwise allows operation up to five (5) percent, the decision is 
immediately effective.
    (2) The Commission. (i) Reserving the power to step in at an 
earlier time, the Commission will, upon receipt of the presiding 
officer's finding under Sec.  52.103(g) with respect to whether 
acceptance criteria in the combined license have been met, other than a 
finding which would otherwise allow only fuel loading and low power (up 
to five (5) percent of rated power) testing, review the matter on its 
own motion to determine whether to stay the effectiveness of the 
finding. A presiding officer finding will be stayed by the Commission, 
insofar as it allows operations other than fuel loading and low power 
testing, if it determines that it is in the public interest to do so, 
based on a consideration of the gravity of the substantive issue, the 
likelihood that it has been resolved incorrectly below, the degree to 
which correct resolution of the issue would be prejudiced by operation 
pending review, and other relevant public interest factors.
    (ii) For findings other than those authorizing only fuel loading 
and low power testing consistent with the target schedule set forth 
below, the parties may file brief comments with the Commission pointing 
out matters which, in their view, pertain to the immediate 
effectiveness issue. To be considered, these comments must be received 
within ten (10) days of the presiding officer's findings. However, the 
Commission may dispense with comments by so advising the parties. An 
extensive stay will not be issued without giving the affected parties 
an opportunity to be heard.
    (iii) The Commission intends to issue a stay decision within thirty 
(30) days of receipt of the presiding officer's findings. The presiding 
officer's findings will be considered stayed pending the Commission's 
decision insofar as such findings may allow operations other than fuel 
loading and operation up to five (5) percent of rated power.
    (iv) In announcing a stay decision, the Commission may allow the 
proceeding to run its ordinary course or give instructions as to the 
future handling of the proceeding. Furthermore, the Commission may, in 
a particular case, determine that compliance with existing regulations 
and policies may no longer be sufficient to warrant a finding that the 
acceptance criteria in the combined license have been met and may alter 
those regulations and policies.
    (j)-(n) [Reserved]
* * * * *
    18. 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

[[Page 12853]]

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:
* * * * *
    19. 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.
    20. In Sec.  2.501, the section heading, the introductory language 
of paragraph (a), and paragraph (b) are revised to read as follows:


Sec.  2.501  Notice of hearing on application under subpart F of 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 and Reserved]

    21. Remove and reserve Sec.  2.502.


Sec.  2.503  [Removed and Reserved]

    22. Remove and reserve Sec.  2.503.


Sec.  2.504  [Removed and Reserved]

    23. Remove and reserve Sec.  2.504.
    24. 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 provisions of subpart B of part 52, apply to 
standard design certification rulemaking.
    25. 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.
    26. 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/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.
    (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 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 Chief, New Reactor Licensing Branch, 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 Rules and 
Directives 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

[[Page 12854]]

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) 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 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.

PART 10--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR 
ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN 
EMPLOYMENT CLEARANCE

    27. 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.

    28. 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.
* * * * *
    29. 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

    30. 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,

[[Page 12855]]

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).

    31. 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 
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.
    32. 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; 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.
    33. 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.
    34. In Sec.  19.11, paragraph (c) is removed and reserved, and the 
introductory text of paragraph (a), and 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) 415-5877, 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

[[Page 12856]]

posted for a minimum of 5 working days or until action correcting the 
violation has been completed, whichever is later.
    35. 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, applicant for 
an early site permit under subpart A of part 52, and applicant for a 
standard design certification under subpart B of part 52 shall afford 
to the Commission at all 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.
    36. 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 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, 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.
    37. 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.
    38. 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

    39. 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).

    40. 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.
    41. 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.
* * * * *
    42. In Sec.  20.2203, paragraphs (c) and (d) are revised to read as 
follows:

[[Page 12857]]

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/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. 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

    43. 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).

    44. 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 supplies 
basic components with respect to a regulatory 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 authorizes fuel load 
and operation under Sec.  52.103 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.
* * * * *
    45. 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
    (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.  50.34(a)(1), Sec.  50.67(b)(2), or Sec.  
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

[[Page 12858]]

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 regulatory 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 
ESP 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, 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 ESP information, a 
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.
* * * * *
    46. 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.
    47. In Sec.  21.21 paragraphs (a)(3) introductory text, (a)(3)(i), 
(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 regulatory approval under part 52 of this chapter--
    (i) Fails to comply with the Atomic Energy Act of 1954, as amended, 
or any applicable 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 regulatory 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 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.
* * * * *
    48. In Sec.  21.51 paragraph (a)(4) is added and paragraph (b) is 
revised to read as follows:


Sec.  21.51  Maintenance and inspection of records.

    (a) * * *

[[Page 12859]]

    (4) Applicants for standard design certification under subpart C 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.
    (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.
    49. In Sec.  21.61, paragraph (b) is revised to read as follows:


Sec.  21.61  Failure to notify.

* * * * *
    (b) Any NRC licensee (including a 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 
fail 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

    50. 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).

    51. The heading of Part 25 is revised to read as set forth above.
    52. 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

    53. 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).

    54. 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.
* * * * *
    55. In Sec.  26.10, paragraph (a) is revised to read as follows:


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;
* * * * *
    56. 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,

[[Page 12860]]

or transport formula quantities of strategic special nuclear 
material (SSNM).
* * * * *

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

    57. 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).

    58. In Sec.  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.
* * * * *
    59. 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 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, 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.
* * * * *

[[Page 12861]]

    60. 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.
    61. 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 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.
    62. 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) 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
    (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),\3\ as well as the 
plans of State governments wholly or partially within the ingestion 
pathway EPZ.\4\ 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

[[Page 12862]]

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.
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    \3\ 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.
    \4\ 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.
---------------------------------------------------------------------------

* * * * *
    (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.
* * * * *
    63. 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 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

[[Page 12863]]

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.
    (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.
* * * * *
    64. 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:
    (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) An estimate of 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) An estimate of the information required in paragraph (b)(2) of 
this section.
    65. In Sec.  50.36, current 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 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 application.
* * * * *
    66. In Sec.  50.36a, the introductory text of 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 will include technical 
specifications that, in addition to requiring compliance with 
applicable provisions of Sec.  20.1301 of this chapter, require that:
* * * * *
    67. 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

[[Page 12864]]

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, 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.
    68. The undesignated center heading before Sec.  50.40 is revised 
as follows:

Standards for Licenses, Certifications, and Regulatory Approvals

    69. 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 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.
    70. 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.
    71. 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) An applicant for a combined license or an amendment of a 
combined license under part 52 of this chapter who proposes to 
construct a utilization facility will be granted the combined license 
or amendment 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.
    (c) 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 either a 
construction permit or combined license 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.
    72. 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

[[Page 12865]]

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 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.
* * * * *
    73. In Sec.  50.47, paragraph (a)(1), the introductory text of 
paragraph (c)(1), paragraphs (c)(1)(i) and (c)(1)(iii)(B) are 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) 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. No finding under this section is necessary 
for issuance of a renewed combined license.
    (iii) For emergency plans submitted by an applicant under 10 CFR 
52.17(b)(2)(ii), no early site permit under subpart A of part 52 of 
this chapter 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. No finding under this section is necessary for 
issuance of a renewed early site permit.
* * * * *
    (c)(1) Failure to meet the applicable standards set forth in 
paragraph (b) of this section may result in the Commission declining to 
issue an operating license or combined license. However, the applicant 
will have an opportunity to demonstrate to the satisfaction of the 
Commission that deficiencies in the plans are not significant for the 
plant in question, that adequate interim compensating actions have been 
or will be taken promptly, or that there are other compelling reasons 
to permit plant operations. Where an applicant for an operating license 
or combined license asserts that its inability to demonstrate 
compliance with the requirements of paragraph (b) of this section 
results wholly or substantially from the decision of state and/or local 
governments not to participate further in emergency planning, or if an 
applicant cannot obtain the certifications required by Sec.  
52.79(a)(22) of this chapter, an operating license or combined license 
may be issued if the applicant demonstrates to the Commission's 
satisfaction that:
    (i) The applicant's inability to comply with the requirements of 
paragraph (b) of this section or Sec.  52.79(a)(22) of this chapter is 
wholly or substantially the result of the non-participation of state 
and/or local governments.
* * * * *
    (iii) * * *
    (B) The utility's measures designed to compensate for any 
deficiencies resulting from State and/or local non-participation. In 
making its determination on the adequacy of a utility plan, the NRC 
will recognize the reality that in an actual emergency, State and local 
government officials will exercise their best efforts to protect the 
health and safety of the public. The NRC will determine the adequacy of 
that expected response, in combination with the utility's compensating 
measures, on a case-by-case basis, subject to the following guidance. 
In addressing the circumstance where applicant's inability to comply 
with the requirements of paragraph (b) of this section or Sec.  
52.79(a)(22) of this chapter, is wholly or substantially the result of 
non-participation of state and/or local governments, it may be presumed 
that in the event of an actual radiological emergency State and local 
officials would generally follow the utility plan. However, this 
presumption may be rebutted by, for example, a good faith and timely 
proffer of an adequate and feasible State and/or local radiological 
emergency plan that would in fact be relied upon in a radiological 
emergency.
* * * * *
    (e) Notwithstanding the requirements of paragraphs (a) and (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).
    74. In Sec.  50.48, the introductory text of paragraph (a)(1) is 
revised 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

[[Page 12866]]

protection plan that satisfies Criterion 3 of appendix A to this part. 
This fire protection plan must:
* * * * *
    75. 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) 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.
* * * * *
    76. In Sec.  50.54, the introductory text, and paragraphs (a)(1), 
(i-1), and (o) 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 operating license issued 
under this part, and the following paragraphs with the exception of 
paragraph (s) of this section are conditions in every combined license 
issued under part 52 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) of this part 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 three (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.
* * * * *
    (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, FEMA identifies one or more deficiencies in the state of 
offsite emergency preparedness, the holder of a combined license under 
10 CFR 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 FEMA 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.
    77. In Sec.  50.55, the heading, the introductory text and 
paragraphs (a), (b), (c), 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 (a), (b), (c), (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 and combined license 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 permit or license 
expires 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.
    (c) Except as modified by this section and Sec.  50.55a, the 
construction permit or

[[Page 12867]]

combined license is subject to the same conditions to which a license 
is subject.
* * * * *
    (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)(10) of this section. 
The interim report should describe the deviation or failure to comply 
that it 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 Sec.  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 or combined license who obtains information reasonably indicating 
the existence of any defect found in the construction or any defect 
found in the final design of a facility as approved and released for 
construction must notify the Commission of the defect through a 
director or responsible officer or designated person as discussed in 
paragraph (e)(10) of this section.
    (iii) The holder of a facility construction permit subject to this 
part or combined 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 (e)(10) 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, 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 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

[[Page 12868]]

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 operating 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 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]
    78. In Section 50.55a, the introductory paragraph, 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),

[[Page 12869]]

(f)(3)(iv)(B), (f)(4)(i), the introductory text of paragraph (g)(3), 
paragraph (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 32, 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.
* * * * *
    (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 13, 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) 
referenced in paragraph (b)(3) of this section at the time the 
construction permit 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 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

[[Page 12870]]

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 13, 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.
    79. 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.
    80. 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.
---------------------------------------------------------------------------

* * * * *
    81. 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 [INSERT EFFECTIVE DATE OF FINAL RULE], 
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 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 [INSERT EFFECTIVE 
DATE OF FINAL RULE], 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.
    82. 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,

[[Page 12871]]

but before [INSERT EFFECTIVE DATE OF FINAL RULE], the licensee shall 
submit the information defined below 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 [INSERT EFFECTIVE DATE OF FINAL 
RULE], the applicant shall submit the information defined below in its 
final safety analysis report.
* * * * *
    83. In Sec.  50.65, paragraphs (a)(1) and (c) are 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), 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 only shall 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.
* * * * *
    (c) The requirements of this section shall be implemented by each 
licensee no later than July 10, 1996. For combined licenses under part 
52, the requirements of this section shall be implemented by the 
licensee no later than 30 days before the scheduled date for initial 
loading of fuel.
    84. 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 Nuclear Reactor Regulation. All furniture, supplies 
and communication equipment will be furnished by the Commission.
* * * * *
    85. In Sec.  50.71, paragraphs (a), (c), (d)(1), and the 
introductory text of paragraph (e) are revised, paragraph (f) is 
redesignated as paragraph (g) and revised, and new paragraph (f) is 
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 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 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 licensee or prepared by the 
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 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

[[Page 12872]]

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.
---------------------------------------------------------------------------

* * * * *
    (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.
    86. 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.
* * * * *
    87. In Sec.  50.75, paragraphs (a) and (b) are revised, 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 paragraphs (e)(3) 
and (f)(1) are 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.
    (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 use 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, following issuance of the combined license until the 
date that the Commission makes the finding under 10 CFR 52.103(g), 
submit a report to the NRC, by March 31 of each year, containing an 
update to the certification described under paragraph (b)(1) of this 
section. 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; and 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

[[Page 12873]]

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.
* * * * *
    88. 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.
    89. In Sec.  50.80, paragraph (a) is 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.
* * * * *
    90. 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.
    91. Section 50.90 is revised to read as follows:


Sec.  50.90  Application for amendment of license or construction 
permit.

    Whenever a holder of a license, including a construction permit and 
operating license under this part, and a 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.  50.4 or Sec.  52.3 of this 
chapter, as applicable, fully describing the changes desired, and 
following as far as applicable, the form prescribed for original 
applications.
    92. 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 licensed 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:
* * * * *
    93. 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 or 
construction permit will be issued to the applicant, the Commission 
will be guided by the considerations which govern the issuance of 
initial licenses or construction 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, combined license or manufacturing license for a facility or 
reactor licensed under Sec.  50.21(b) or Sec.  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:
* * * * *
    94. 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 part 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 that failure to make timely completion of 
the proposed construction or alteration of a facility under a 
construction permit 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.
    95. 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

[[Page 12874]]

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
    (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.
* * * * *
    96. 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.
    97. 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.

* * * * *
    98. 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 program to be 
applied to the design, 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 [INSERT DATE OF FINAL 
RULE], every applicant for an early site permit under part 52 of 
this chapter is required by the provisions of Sec.  52.17 to

[[Page 12875]]

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, 
design certification, or manufacturing license under part 52 of this 
chapter is required by the provisions of 10 CFR 52.137, 52.47, and 
52.157, respectively, to include in its final 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. Nuclear power plants and 
fuel reprocessing plants include structures, systems, and components 
that prevent or mitigate the consequences of 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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

* * * * *
    99. 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 apprise 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.
* * * * *

    100. 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, and 10 are redesignated as 
7, 8, 9, 10, and 11, 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.
* * * * *

III. The Final Safety Analysis Report or Early Site Permit Application

    The final safety analysis report shall contain the plans for 
coping with emergencies. Early site permit applications may contain 
plans for coping with emergencies under Sec.  52.17(b) of this 
chapter. 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.

[[Page 12876]]

    The plans submitted must include a description of the elements 
set out in Section IV for the emergency planning zones (EPZs) 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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    (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 one 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 
FEMA 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. of this appendix. If FEMA 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 date that the Commission makes the 
finding under Sec.  52.103 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.
* * * * *
    101. In Appendix I to Part 50, the first paragraphs of Sections I, 
II, IV, V, and the introductory paragraph of Sections A.3 of the 
Concluding Statement of Position of the Regulatory Staff (Docket-RM-50-
2) 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

[[Page 12877]]

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.
* * * * *
    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 construction permit under this part or a design 
certification, a combined license, or a manufacturing license under 
part 52 of this chapter.
* * * * *

Concluding Statement of Position of the Regulatory Staff (Docket-RM-50-
2) Guides on Design Objectives for Light-Water-Cooled Nuclear Power 
Reactors

    A.* * *
    3. Notwithstanding the guidance in paragraph A.2, for a 
particular site, if an applicant for a construction permit under 
this part or a design approval, a design certification, a combined 
license, or a manufacturing license under part 52 of this chapter 
has proposed baseline in-plant control measures \2\ to reduce the 
possible sources of radioactive material in liquid effluent releases 
and the calculated quantity exceeds the quantity set forth in 
paragraph A.2, the requirements for design objectives for 
radioactive material in liquid effluents may be deemed to have been 
met provided:
---------------------------------------------------------------------------

    \2\ These measures may include treatment of clear liquid waste 
streams (normally tritiated, nonaerated, low conductivity equipment 
drains and pump seal leakoff), dirty liquid waste streams (normally 
nontritiated, aerated, high conductivity building sumps, floor and 
sample station drains), steam generator blowdown streams, chemical 
waste streams, low purity and high purity liquid streams (resin 
regenerate and laboratory wastes), as appropriate for the type of 
reactor.
---------------------------------------------------------------------------

* * * * *
    102. 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.
    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 and Reserved]

    103. Appendix M to Part 50 is removed and reserved.

Appendix O to Part 50 [Removed and Reserved]

    104. Appendix O to Part 50 is removed and reserved.
    105. 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

[[Page 12878]]

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.
    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

    106. 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)).

    107. 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, 51.58, 51.60, 51.61, 51.62, 51.66, 51.68, and 51.69.
    108. In Sec.  51.20, paragraph (b)(6) is removed and reserved, and 
paragraphs (b)(1) and (b)(2) are revised to 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]
* * * * *
    109. In Sec.  51.22, the introductory text of paragraph (c)(3), 
paragraphs (c)(3)(i), (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.
* * * * *
    110. 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.

* * * * *

[[Page 12879]]

    (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 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.
    111. In Sec.  51.30, 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 or a manufacturing license 
under part 52 of this chapter, shall identify the proposed action and 
include:
    (1) A brief discussion of:
    (i) The need for the proposed action;
    (ii) Alternatives as required by section 102(2)(E) of NEPA;
    (iii) The environmental impacts of the proposed action and 
alternatives as appropriate; and
    (2) A list of agencies and persons consulted, and identification of 
sources used.
* * * * *
    (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 (SAMDAs) and 
the bases for not incorporating SAMDAs 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 SAMDA previously 
rejected in the earlier environmental assessment to become cost 
beneficial, or results in the identification of new SAMDAs, in which 
case the costs and benefits of new SAMDAs and the bases for not 
incorporating new SAMDAs 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 
SAMDAs and the bases for not incorporating SAMDAs in the manufacturing 
license. An environmental assessment for an amendment to a 
manufacturing license will be limited to consideration whether the 
design change which is the subject of the proposed amendment either 
renders a SAMDA previously rejected in an environmental assessment to 
become cost beneficial, or results in the identification of new SAMDAs, 
in which case the costs and benefits of new SAMDAs and the bases for 
not incorporating new SAMDAs 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.
    112. 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 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 that in Sec.  51.32 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) or (e), as 
applicable.
    (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 NRC's Director of Nuclear Reactor Regulation (staff director) will 
determine the costs and benefits of severe accident mitigation design 
alternatives (SAMDAs) and the bases for not incorporating SAMDAs 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 that in Sec.  51.32 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 SAMDAs as required by Sec.  51.30(d) or (e), as 
applicable.
    (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

[[Page 12880]]

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. 
The presiding officer will issue the final environmental assessment.
    (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), may take a position and 
offer evidence on the matters within the scope of the environmental 
assessment.
    113. 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.
    114. In Sec.  51.45 paragraph (c) is revised to read as follows:


Sec.  51.45  Environmental report.

* * * * *
    (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 to 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.
* * * * *
    115. 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 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. Environmental reports need not 
include an assessment of the economic, technical, and other benefits 
and costs of the proposed action or an analysis of other energy 
alternatives. 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. 
Environmental reports must include an evaluation of alternative sites 
to determine whether there is any obviously superior alternative to the 
site proposed. If the applicant seeks to perform the activities at the 
site allowed by Sec.  50.10(e)(1) of this chapter, the environmental 
report must include a plan for redress of the site that will achieve an 
environmentally stable and aesthetically acceptable site suitable for 
whatever non-nuclear use may conform with local zoning laws. 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.
    (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; 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. The applicant 
must have a reasonable process for identifying any new and significant 
information regarding the NRC's conclusions in the early site permit 
environmental impact statement. 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,'' 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

[[Page 12881]]

characteristics and design parameters specified in the early site 
permit;
    (ii) Information to resolve any other significant environmental 
issue not considered in the early site permit proceeding, either for 
the site or design; and
    (iii) Any new and significant information on the site or design to 
the extent that it differs from, or is in addition to, that discussed 
in the early site permit environmental impact statement.
    (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.
    (4) Application requesting authority to conduct activities under 
Sec.  50.10(e) of this chapter. If the applicant seeks to perform 
activities at the site allowed by Sec.  50.10(e) of this chapter, then 
the environmental report must include a plan for redress of the site 
that will achieve an environmentally stable and aesthetically 
acceptable site suitable for whatever non-nuclear use may conform with 
local zoning laws.
    116. 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.
* * * * *
    117. 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-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.
* * * * *
    118. 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 renewal 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:
* * * * *
    119. 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 (SAMDAs), and the 
bases for not incorporating SAMDAs 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.
    (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 SAMDA previously rejected in an environmental assessment to 
become cost beneficial, or results in the identification of new SAMDAs 
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.
    120. 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 or 
operate a

[[Page 12882]]

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 for the nuclear power 
plant shall submit a copy to the Director of the Office of Nuclear 
Reactor Regulation, or a copy to 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/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. 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 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.  50.4 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 Nuclear 
Reactor Regulation.
    121. Section 51.55 is added to read as follows:


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 (SAMDAs), and the basis for not incorporating SAMDAs 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 SAMDA previously rejected in an environmental assessment to 
become cost beneficial, or results in the identification of new SAMDAs 
that may be reasonably incorporated into the design certification.
    122. 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.  
51.61 or Sec.  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.
    123. 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, 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 
environmental impact statement prepared at the early site permit stage 
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, and will not include an assessment of 
the benefits (for example, need for power) of the proposed action or an 
evaluation of other alternative energy sources unless considered by the 
applicant, but must include an evaluation of alternative sites to 
determine whether there is any alternative to the site proposed. 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 significant new environmental 
issue or information not considered on the site or the design only

[[Page 12883]]

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. 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 the assessment of aquatic impacts is no longer 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.
---------------------------------------------------------------------------

* * * * *
    124. 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\ 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.
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    (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, 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.
    (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 and 
the design of

[[Page 12884]]

the facility falls within the site characteristics and design 
parameters specified in the early site permit, then the draft 
supplemental combined license environmental impact statement shall 
incorporate by reference the early site permit final environmental 
impact statement, and summarize the findings and conclusions of the 
early site permit final environmental impact statement.
    (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 falls 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 falls 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 SAMDAs. 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 and Reserved]

    125. Section 51.76 is removed and reserved.
    126. 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 for a nuclear power plant under parts 52 or 54 
of this chapter, the Commission shall prepare an EIS, 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 postoperating or post combined license stage 
or an environmental assessment, as appropriate, which will update the 
prior environmental review. 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.
    127. 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 section 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 
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.
    128. 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:
    (a) Determine, in an uncontested proceeding, whether the NEPA 
review conducted by the NRC staff has been

[[Page 12885]]

adequate to identify all reasonable SAMDAs for the design of the 
reactor to be manufactured and evaluate the environmental, technical, 
economic, and other benefits and costs of each SAMDA; and
    (b) Determine, in a contested proceeding, whether in accordance 
with the regulations in this subpart, the manufacturing license should 
be issued as proposed by the NRC's Director of Nuclear Reactor 
Regulation.
    129. Section 51.107 is added to read as follows:


Sec.  51.107  Public hearings in proceedings for issuance of combined 
licenses.

    (a) In addition to complying with applicable requirements of Sec.  
51.104, in a proceeding for the issuance of a combined license for a 
nuclear power reactor, the presiding officer will:
    (1) Determine whether the requirements of section 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 Nuclear Reactor Regulation.
    (b) If the combined license application references an early site 
permit, then the presiding officer in a combined license hearing shall 
not admit contentions proffered by any party on environmental issues 
which have been accorded finality under Sec.  52.39 of this chapter, 
unless this contention--
    (1) Demonstrates that the design of the facility falls outside the 
design parameters specified in the early site permit;
    (2) Demonstrates that the site no longer falls within the site 
characteristics specified in the early site permit; or
    (3) Raises any other significant environmental issue not considered 
which is material to the site or the design only to the extent that it 
differs from those discussed or it reflects significant new information 
in addition to that discussed in the final environmental impact 
statement prepared by the Commission in connection with the early site 
permit.
    (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.
    130. Section 51.108 is added under the undesignated center heading 
``Production and Utilization Facilities,'' to read as follows:


Sec.  51.108  Public hearings on a Commission findings that 
inspections, tests, 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).
    131. 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.
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.
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.

[[Page 12886]]

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.
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 applications.
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

    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).

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 50.5.
    (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, or a combined license 
application.
    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.
    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.
    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 a 
manufacturing license.
    Standard design means a design which is sufficiently detailed and 
complete to support certification 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.
    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.

[[Page 12887]]

    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, 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, 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 
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, 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.  50.34 of this chapter;
    (ii) Change to an emergency plan under Sec.  50.54(q) of this 
chapter;
    (iii) Emergency implementing procedures under appendix E, Section V 
of this part.
    (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 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)(3) 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)(3) of this chapter, must be submitted to 
the NRC's Document Control Desk, with a copy to

[[Page 12888]]

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) Applicant for a standard design certification;
    (3) Applicant for a license;
    (4) Applicant for a standard design approval;
    (5) Employee of a licensee.
    (6) Employee of an applicant for a license, a standard design 
certification, or a standard design approval;
    (7) Any contractor (including a supplier or consultant), 
subcontractor, or employee of a contractor or subcontractor of any 
licensee; or
    (8) 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.
    License means a license issued under this part, including an early 
site permit.
    Licensee means any person holding a license issued under this part, 
including an early site permit.
    (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 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 (a)(1) or (a)(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 ederal 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
    (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

[[Page 12889]]

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 rule;
    (3) Imposition of a civil penalty on the licensee, holder of a 
standard design approval, or applicant; or
    (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(c). 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 part 52, and for 30 days following license termination or the 
expiration or termination of the standard design certification or 
standard design approval under 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) 415-5877, 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 
a construction permit holder, and a combined 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, or license 
conditions 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 governed by Sec.  
50.12. 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.

    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 license, standard design approval, or standard design 
certification under this part shall be deemed to have been

[[Page 12890]]

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 standard design certification or 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, and D.

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 
Nuclear Reactor Regulation. 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) 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 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 
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;
    (x) For nuclear power facilities to be sited on multi-unit sites, 
an evaluation of the potential hazards to the

[[Page 12891]]

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;
    (xi) Information demonstrating that site characteristics are such 
that adequate security plans and measures can be developed;
    (xii) For applications submitted after [insert date of final rule], 
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 of this part will be satisfied; 
and
    (xiii) 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 was issued to 
establish criteria that the NRC staff intends to use in evaluating 
whether an applicant/licensee meets the Commission's regulations. 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 application 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 application may also:
    (i) Propose major features of the emergency plans in the site 
safety analysis report, 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, that can be reviewed and approved by NRC in consultation with 
the Federal Emergency Management Agency (FEMA) in the absence of 
complete and integrated emergency plans; or
    (ii) Propose complete and integrated emergency plans in the site 
safety analysis report for review and approval by the NRC, in 
consultation with FEMA, 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, and each major feature of an emergency plan, 
submitted under paragraph (b)(2) 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 operate in conformity with the license, the 
provisions of the Atomic Energy Act, and the NRC's regulations.
    (4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the 
application must include a description of contacts and arrangements 
made with Federal, State, and local governmental agencies with 
emergency planning responsibilities. The application must contain any 
certifications that have been obtained. If these certifications cannot 
be obtained, 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. 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 and describe in the site 
safety analysis report the activities that are requested, and propose a 
plan in the environmental report 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.
    (d) The NRC staff will advise the applicant on whether any 
information beyond that required by this section must be submitted.


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 consultation with FEMA, 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

[[Page 12892]]

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, and L 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 shall 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.
    (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)(1) 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 renew the permit.
    (2) An early site permit also continues to be valid beyond the date 
of expiration in any proceeding on an operating license application 
which is based on a construction permit that references the early site 
permit, and in any hearing held under Sec.  52.103 before operation 
begins under a combined license which references the early site 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.


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 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

[[Page 12893]]

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.


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 Nuclear Reactor Regulation 
(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, operating license, or combined license, or the findings 
required by Sec.  52.103, if the application for the construction 
permit, operating license, 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. If the early site permit approved an emergency 
plan (or major features thereof) that are 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.
    (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 
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; 
or
    (v) Any significant environmental issue not considered which is 
material to the site or the design to the extent that it differs from 
those discussed or it reflects significant new information in addition 
to that discussed in the final environmental impact statement prepared 
by the Commission in connection with the early site permits.
    (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 one or more 
site characteristics, design parameters, or terms and conditions of the 
early site permit. 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. A variance will not be issued once the construction permit, 
operating license, or combined license is issued.
    (e) 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 
certification for nuclear power facilities separate from the filing of 
an application for a

[[Page 12894]]

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.


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 
or combined license for such a facility has not been filed.
    (b) The application must comply with the applicable filing 
requirements of 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 
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
    (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

[[Page 12895]]

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) 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;
    (6) 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;
    (7) An analysis and description of the equipment and systems for 
combustible gas control as required by 10 CFR 50.44;
    (8) A coping analysis, and any design features necessary to address 
station blackout, as required by 10 CFR 50.63;
    (9) A description of the kinds and quantities of radioactive 
materials expected to be produced and used in the construction and 
operation and the design features for controlling and limiting 
radioactive effluents and radiation exposures within the limits set 
forth in 10 CFR part 20;
    (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 on electric equipment important to safety that 
is required by 10 CFR 50.49(d);
    (12) Information demonstrating how the applicant will comply with 
requirements for reduction of risk from anticipated transients without 
scram (ATWS) events in Sec.  50.62;
    (13) Information demonstrating how the applicant will comply with 
requirements for criticality accidents in Sec.  50.68(b)(2) through 
(b)(4);
    (14) through (15) [Reserved]
    (16) The information necessary to demonstrate that SSCs important 
to safety comply with the earthquake engineering criteria in 10 CFR 
part 50, appendix S;
    (17) 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);
    (18) The information necessary to demonstrate technical resolutions 
of those unresolved safety issues and medium- and high-priority generic 
safety issues that are identified in the version of NUREG-0933 current 
on the date 6 months before the docket date of the application and that 
are technically relevant to the standard plant design;
    (19) The information necessary to demonstrate how operating 
experience insights from generic letters and bulletins issued up to six 
months before the docket date of the application, or comparable 
international operating experience, have been incorporated into the 
plant design;
    (20) A description and analysis of design features for the 
prevention and mitigation of severe accidents (core-melt accidents), 
including challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen detonation, and containment bypass;
    (21) A description of the quality assurance program to be 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 will be satisfied;
    (22) Proposed technical specifications prepared in accordance with 
the requirements of Sec. Sec.  50.36 and 50.36a of this chapter;
    (23) The technical qualifications of the applicant to engage in the 
proposed activities in accordance with the regulations in this chapter;
    (24) A description of the design features that will provide 
physical protection of the standard plant design in accordance with the 
requirements of 10 CFR part 73;
    (25) A representative conceptual design for those portions of the 
standard plant for which the application does not seek certification, 
to aid the NRC in its review of the final safety analysis and 
probabilistic risk assessment, and to permit assessment of the adequacy 
of the interface requirements in paragraph (b)(3) of this section;
    (26) 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 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 was issued to establish criteria that the 
NRC staff intends to use in evaluating whether an applicant meets the 
Commission's regulations. The SRP is not a substitute for the 
regulations, and compliance is not a requirement; and
    (27) The NRC staff will advise the applicant on whether any 
technical information beyond that required by this section must be 
submitted.
    (b) The application must also contain:
    (1) A design-specific probabilistic risk assessment (PRA);
    (2) The proposed inspections, tests, analyses, and acceptance 
criteria (ITAAC) that are necessary and sufficient to provide 
reasonable assurance that, if the inspections, tests, and analyses are 
performed and the acceptance criteria met, a plant that incorporates 
the design certification is

[[Page 12896]]

built and will operate in accordance with the design certification, the 
provisions of the Act, and the Commission's regulations;
    (3) 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 
final safety analysis and design-specific PRA required by this section;
    (4) Justification that compliance with the interface requirements 
of paragraph (b)(3) of this section is verifiable through inspection, 
testing (either in the plant or elsewhere), or analysis. The method to 
be used for verification of interface requirements must be included as 
part of the proposed ITAAC required by paragraph (b)(2) of this 
section; and
    (5) An evaluation of severe accident mitigation design alternatives 
to the plant design under 10 CFR 51.30, and a description of how cost-
beneficial design alternatives are included in the standard plant 
design.
    (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 the various options for the configuration 
of the plant and site, including variations in, or sharing of, common 
systems, interface requirements, and system interactions. The final 
safety analysis and the PRA must also account for differences among the 
various options, 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;
    (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 shall 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 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.


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

[[Page 12897]]

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; or
    (iii) Reduces unnecessary regulatory burden and maintains 
protection to public health and safety and the common defense and 
security.
    (2) The rulemaking procedures 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 or operating 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 standard design 
certification rule may request an exemption from one or more elements 
of the design 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 must 
be subject to litigation in the same manner as other issues in the 
operating license or combined license hearing.

[[Page 12898]]

    (2) Subject to Sec.  50.59 of this chapter, a licensee who 
references a standard design certification rule may make changes to the 
design of the nuclear power facility, without prior Commission 
approval, unless the proposed change involves a change to the design as 
described in the rule certifying the design. The licensee shall 
maintain records of all changes to 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, or operating license which references a 
standard 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 Nuclear Reactor Regulation.
    (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. The application must also state the earliest and latest 
dates for completion of construction.


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) 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:
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    (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).
---------------------------------------------------------------------------

    \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.
---------------------------------------------------------------------------

    (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

[[Page 12899]]

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 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.
---------------------------------------------------------------------------

    (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) The analyses and the descriptions of the equipment and systems 
required by Sec.  50.44 of this chapter for combustible gas control;
    (9) The coping analyses required, and any necessary design features 
necessary to address station blackout, as described in Sec.  50.63 of 
this chapter;
    (10) A description of the program 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) necessary to ensure that the 
systems and components meet the requirements of the ASME Boiler and 
Pressure Vessel Code in accordance with Sec.  50.55a of this chapter;
    (12) A description of the primary containment leakage rate testing 
program 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;
    (14) A description of the operator training program necessary to 
meet the requirements of 10 CFR part 55;
    (15) A description of the program for monitoring the effectiveness 
of maintenance necessary to meet the requirements of Sec.  50.65 of 
this chapter;
    (16) 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;
    (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 SSCs important 
to safety comply 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 that are 
identified in the version of NUREG-0933 current on the date 6 months 
before application and that 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;

[[Page 12900]]

    (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) 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;
    (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 to be applied 
to the design, 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 shall include a discussion of how the 
applicable requirements of appendix B to 10 CFR part 50 will be 
satisfied;
    (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) Plans for conduct of normal operations, including maintenance, 
surveillance, and periodic testing of structures, systems, and 
components;
    (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;
    (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) 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;
    (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.)
---------------------------------------------------------------------------

    \8\ A physical security plan that contains all the information 
required in both Sec. Sec.  73.55 of this chapter and appendix C to 
10 CFR part 73 satisfies the requirement for a contingency plan.
---------------------------------------------------------------------------

    (ii) 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 which demonstrates how operating experience 
insights from generic letters and bulletins issued up to 6 months 
before the docket date of the application, or comparable international 
operating experience, have been incorporated into the plant design;
    (38) A description and analysis of design features for the 
prevention and mitigation of severe accidents (core-melt accidents), 
including challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen detonation, and containment bypass;
    (39) The earliest and latest dates for completion of the 
construction;
    (40) [Reserved]
    (41) For applications for light-water cooled nuclear power plant 
combined licenses, an evaluation of the facility against the Standard 
Review Plan (SRP) 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 was issued to establish criteria that the NRC staff intends to use 
in evaluating whether an applicant/licensee meets the Commission's 
regulations. 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) The NRC staff will advise the applicant on whether any 
information beyond that required by this section must be submitted.
    (b) If the application for a final safety analysis report 
references an early site

[[Page 12901]]

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, but 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 
will be satisfied by the date of issuance 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. If the proposed facility emergency plans incorporate 
existing emergency plans or major features of emergency plans, 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 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, but 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.
    (2) The final safety analysis report must demonstrate that the 
interface requirements established for the design under Sec.  52.137 
have been met.
    (3) 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, but 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.
    (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 must be satisfied by the date of issuance of the 
combined license.
    (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, but must contain, in addition to the information 
and analyses otherwise required, information sufficient to demonstrate 
that the site parameters for the manufactured reactor are bounded by 
the site where the manufactured reactor is to be installed and used.
    (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 will be satisfied by the date of issuance of the combined 
license.


Sec.  52.80  Contents of applications; additional technical 
information.

    The application must contain:
    (a) A plant-specific probabilistic risk assessment (PRA). If the 
application references a standard design certification or standard 
design approval, or if the application proposes to use a nuclear power 
reactor manufactured under a manufacturing license under subpart F of 
this part, the plant-specific PRA must use the PRA for the design 
certification, design approval, or manufactured reactor, as applicable, 
and must be updated to account for site-specific design information and 
any design changes, departures, or variances.
    (b) The proposed inspections, tests, and analyses, including those 
applicable to emergency planning, that the licensee shall perform, and 
the acceptance criteria which 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 combined license, 
the provisions of the Atomic Energy Act, and the NRC's 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.
    (c) A complete environmental report as required by 10 CFR 51.50(c).


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.

    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 licensed issued 
are governed by the relevant provisions addressing finality, including 
Sec. Sec.  52.39, 52.63, 52.98, 52.145, and 52.171.

[[Page 12902]]

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) of this section, the application 
for the license is withdrawn or denied, and the early site permit 
referenced by the application expires, 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 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 this subpart 
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. 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.
    (c) Issuance of the variance 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 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.  50.54, Sec.  50.59, or 
Sec.  50.90).
    (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

[[Page 12903]]

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 variances 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 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) Holders of combined licenses shall comply with the provisions 
of 10 CFR 50.70 and 50.71.
    (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 particular ITAAC has been 
met.
    (c) The licensee shall notify the NRC that the inspections, tests, 
or analyses in the ITAAC have been successfully completed and that the 
corresponding acceptance criteria have been met. For those inspections, 
tests, or analyses that are completed within 180 days prior to the 
scheduled date for initial loading of fuel, the licensee shall notify 
the NRC within 10 days of the successful completion of ITAAC.
    (d)(1) 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. A request for a variance or 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 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 or request a license amendment under Sec.  
52.98(f).
    (e) The NRC shall ensure that the required inspections, tests, and 
analyses in the ITAAC are performed. At appropriate intervals, the NRC 
shall publish notices in the Federal Register of the NRC staff's 
determination of the successful completion of inspections, tests, and 
analyses.


Sec.  52.103  Operation under a combined license.

    (a) 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 subpart C of 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) After receiving a request for a hearing, the Commission 
expeditiously shall either deny or grant the request. If the request is 
granted, the Commission shall determine, after considering petitioners' 
prima facie showing and any answers thereto, whether during a period of 
interim operation, there will be reasonable assurance of adequate 
protection of the public health and safety. If the Commission 
determines that there is reasonable assurance, it shall allow operation 
during an interim period under the combined license.
    (d) The Commission shall determine appropriate hearing procedures 
in accordance with 10 CFR part 2 for any hearing under paragraph (a) of 
this section.
    (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.
    (g) The licensee shall not load fuel into the reactor and shall not 
operate the facility until the Commission makes a finding that the 
acceptance criteria in

[[Page 12904]]

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--
    (a) 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
    (b) 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 [insert the 
effective date of this rule], 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 12905]]

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 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 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 12906]]

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) 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;
    (6) 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;
    (7) An analysis and description of the equipment and systems for 
combustible gas control as required by 10 CFR 50.44;
    (8) A coping analysis, and any design features necessary to address 
station blackout, as required by 10 CFR 50.63;
    (9) A description of the kinds and quantities of radioactive 
materials expected to be produced and used in the construction and 
operation and the design features for controlling and limiting 
radioactive effluents and radiation exposures within the limits set 
forth in 10 CFR part 20;
    (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 on electric equipment important to safety that 
is required by 10 CFR 50.49(d);
    (12) Information demonstrating how the applicant will comply with 
requirements for reduction of risk from anticipated transients without 
scram (ATWS) events in Sec.  50.62;
    (13) Information demonstrating how the applicant will comply with 
requirements for criticality accidents in Sec.  50.68(b)(2) through 
(b)(4);
    (14)-(15) [Reserved]
    (16) The information necessary to demonstrate that SSCs important 
to safety comply with the earthquake engineering criteria in 10 CFR 
part 50, appendix S;
    (17) 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);
    (18) The information necessary to demonstrate technical resolutions 
of those unresolved safety issues and medium- and high-priority generic 
safety issues that are identified in the version of NUREG-0933 current 
on the date 6 months before the docket date of the application and that 
are technically relevant to the standard plant design;
    (19) The information necessary to demonstrate how operating 
experience insights from generic letters and bulletins issued up to 6 
months before the docket date of the application, or comparable 
international operating experience, has been incorporated into the 
plant design;

[[Page 12907]]

    (20) A description and analysis of design features for the 
prevention and mitigation of severe accidents (core-melt accidents), 
including challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen detonation, and containment bypass;
    (21) A description of the quality assurance program to be 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 will be satisfied;
    (22) 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;
    (23) The technical qualifications of the applicant to engage in the 
proposed activities in accordance with the regulations in this chapter;
    (24) A description of the design features that will provide 
physical protection of the standard plant design in accordance with the 
requirements of 10 CFR part 73;
    (25) [Reserved]
    (26) An evaluation of the standard 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 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 alternative proposed provides an 
acceptable method of complying with Commission's regulations, or 
portions thereof, that underlie the corresponding SRP acceptance 
criteria. The SRP was issued to establish criteria that the NRC staff 
intends to use in evaluating whether an applicant meets the 
Commission's regulations. The SRP is not a substitute for the 
regulations, and compliance is not a requirement; and
    (27) The NRC staff will advise the applicant on whether any 
technical information beyond that required by this section must be 
submitted.
    (b) The application must also contain:
    (1) A design-specific probabilistic risk assessment (PRA);
    (2) [Reserved]
    (3) A description, analysis, and evaluation of the interfaces 
between the standard design and the balance of the nuclear power plant.
    (c) 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, combined 
license, or an operating license which references the standard 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 preliminary or final 
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 Nuclear Reactor Regulation.
    (b) The application must comply with the applicable filing 
requirements of Sec. Sec.  52.3 and 50.30 of this chapter.

[[Page 12908]]

    (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 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 to be 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
    (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.
    (e) Information necessary to establish that the design of the 
reactor to be manufactured complies with the technical requirements in 
part 50 of this chapter, 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 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) The analyses and the descriptions of the equipment and systems 
required by Sec.  50.44 of this chapter for combustible gas control;
    (5) The coping analyses required, and any design features necessary 
to address station blackout, as described in Sec.  50.63 of this 
chapter;
    (6) The information on 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

[[Page 12909]]

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) 
through(b)(4);
    (9) through (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;
    (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 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 to be applied 
to the design and 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 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;
    (f) The site parameters postulated for the design, and an analysis 
and evaluation of the reactor design in terms of those site parameters;
    (g) 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 and probabilistic risk assessment required by Sec.  
52.158(a);
    (h) Justification that compliance with the interface requirements 
of paragraph (a)(18) of this section is verifiable through inspection, 
testing (either in the plant or elsewhere), or analysis;
    (i) 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 the probabilistic 
risk assessment required by Sec.  52.158(a), and to permit assessment 
of the adequacy of the interface requirements in paragraph (g) of this 
section;
    (j) A description and analysis of design features for the 
prevention and mitigation of severe accidents (core-melt accidents), 
including challenges to containment integrity caused by core-concrete 
interaction, steam explosion, high-pressure core melt ejection, 
hydrogen detonation, and containment bypass;
    (k) [Reserved]
    (l) If the reactor is to be used in modular plant design, the 
various options for the configuration of the plant and site, including 
variations in, or sharing of, common systems, interface requirements, 
and system interactions must be described. The final safety analysis 
and the probabilistic risk assessment must account for differences 
among the various options, including any restrictions which will be 
necessary during the construction and startup of a given module to 
ensure the safe operation of any module already operating;
    (m) A description of the management plan for design and 
manufacturing activities, including:
    (1) The organizational and management structure singularly 
responsible for direction of design and manufacture of the reactor;
    (2) Technical resources directed by the applicant, and the 
qualifications requirements;
    (3) 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;
    (4) 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
    (5) 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;
    (n) Necessary parameters to be used in developing plans for 
preoperational testing and initial operation;
    (o) 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 application and which are technically relevant to the 
design;
    (p) A description of how operating experience insights from generic 
letters and bulletins issued up to six months before the docket date of 
the application, or comparable international operating experience, has 
been incorporated into the design of the reactor to be manufactured;
    (q) An evaluation of the site against applicable sections of the 
Standard Review Plan 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 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 was issued to 
establish criteria that the NRC staff intends to use in evaluating 
whether an applicant/licensee meets the Commission's regulations. The 
SRP is not a substitute for the regulations, and compliance is not a 
requirement; and
    (r) The NRC staff shall advise the applicant if any information 
beyond that required by this section must be submitted.


Sec.  52.158  Contents of application; additional technical 
information.

    The application must contain:
    (a) Probabilistic risk assessment (PRA). A design-specific PRA for 
the reactor. If the application references a certified design, the PRA 
for the certified design must be updated to reflect any additional 
portions of the reactor to be manufactured which are not within the 
scope of the certified design.
    (b)(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

[[Page 12910]]

acceptance criteria which 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 conformance with the 
manufacturing license; the provisions of the Atomic Energy Act, and the 
NRC's regulations; and
    (ii) The reactor will operate in conformity with design 
characteristics in the manufacturing license, any license authorizing 
operation of the reactor as part of a nuclear power plant, the 
provisions of the Act, and the NRC's regulations.
    (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 criterion 
has been met. The Federal Register notification required by Sec.  
52.163 must indicate that the application includes this notification.
    (c)(1) An environmental report as required by 10 CFR 51.54. The 
report must address the costs and benefits of severe accident 
mitigation design alternatives (SAMDAs), and the bases for not 
incorporating SAMDAs 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 related environmental assessment prepared by the NRC will 
be similarly directed.
    (2) If the 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 
hearing in Sec.  2.104 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 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 and L.


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 conducting a hearing in accordance with 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) The number of nuclear power reactors authorized to be 
manufactured, and the latest date for completion of the manufacturing 
of all the reactors. The number of reactors to be specified in the 
manufacturing license may be no more than the number of reactors whose 
start of manufacture can practically begin within a 10-year period 
commencing on the date of issuance of the manufacturing license;
    (4) Site parameters and design characteristics for the manufactured 
reactor; and
    (5) 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) 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).


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

[[Page 12911]]

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)(1) of this section.
    (3) In making the findings required for issuance of a construction 
permit, operating license, combined license, and for any hearing under 
Sec.  52.103, 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 SAMDAs, and the bases for not incorporating SAMDAs 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 
through 50.92.
    (2) An applicant or licensee who references or uses a nuclear power 
reactor manufactured under a manufacturing license under this subpart 
may request a variance 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 variance will comply with the requirements of 10 
CFR 50.12(a), and that the special circumstances outweigh any decrease 
in safety that may result from the reduction in standardization caused 
by the exemption. The granting of a variance on request of an applicant 
must be subject to litigation in the same manner as other issues in the 
construction permit, operating license, 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 filed 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 filed 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 site permit 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 valid for not less than 5, 
nor more than 15 years from the date of renewal, and 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.

[[Page 12912]]

    (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 this 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 this 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, 52.63, 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, 52.99, 52.103, 52.104, 52.105, 52.107, 52.109, 52.131, 
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, 52.159, 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 (hereinafter 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 10 CFR 52.47, with the exception of 
generic technical specifications and conceptual design information;
    2. Information required for a final safety analysis report under 
10 CFR 50.34;
    3. 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
    4. 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 license that wishes to reference this 
appendix shall, in addition to complying with the requirements of 10 
CFR 52.77, 52.78, and 52.79, 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;

[[Page 12913]]

    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. Physically 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.

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 52.97(b). 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

[[Page 12914]]

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 50.7 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 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 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 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 a 
severe accident issue identified in the plant-specific DCD, requires 
a license amendment if:
    (1) There is a substantial increase in the probability of a 
severe accident such that a particular 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 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 technical specifications and other 
operational requirements 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 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 50.12(a). The grant of an exemption must be 
subject to litigation in the same manner as other issues material to 
the license hearing.

[[Page 12915]]

    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 and Tier 2. 
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 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), 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.
---------------------------------------------------------------------------

    \1\ ``System 80+'' is a trademark of Westinghouse Electric 
Company LLC.
---------------------------------------------------------------------------

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 (hereinafter 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

[[Page 12916]]

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 10 CFR 52.47, with the exception of 
generic technical specifications and conceptual design information;
    2. Information required for a final safety analysis report under 
10 CFR 50.34;
    3. 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
    4. 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 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 license that wishes to reference this 
appendix shall, in addition to complying with the requirements of 10 
CFR 52.77, 52.78, and 52.79, 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(a) that is not within 
the scope of this appendix.
    3. Physically 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

[[Page 12917]]

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 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 52.97(b). 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 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 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 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 a 
severe accident issue identified in the plant-specific DCD, requires 
a license amendment if--
    (1) There is a substantial increase in the probability of a 
severe accident such that a particular 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 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.

[[Page 12918]]

    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) 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 technical specifications and other 
operational requirements 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 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 50.12(a). 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.

[[Page 12919]]

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 and Tier 2. 
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)(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.
---------------------------------------------------------------------------

    \1\ AP600 is a trademark of Westinghouse Electric Company LLC.
---------------------------------------------------------------------------

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 (hereinafter 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 10 CFR 52.47, with the exception of 
generic technical specifications and conceptual design information;
    2. Information required for a final safety analysis report under 
10 CFR 50.34;
    3. 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
    4. Combined license (COL) action items (combined 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.
    5. 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

[[Page 12920]]

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 license that wishes to reference this 
appendix shall, in addition to complying with the requirements of 10 
CFR 52.77, 52.78, and 52.79, 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
    f. Information required by 10 CFR 52.47(a) that is not within 
the scope of this appendix.
    3. Physically 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 (SAMDAs) 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 SAMDA 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.

[[Page 12921]]

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 Sec.  52.97(b). 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 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 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 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 a 
severe accident issue identified in the plant-specific DCD, requires 
a license amendment if:
    (1) There is a substantial increase in the probability of a 
severe accident such that a particular 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 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.
    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.
    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

[[Page 12922]]

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 technical specifications and other 
operational requirements 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 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 50.12(a). 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.

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 and Tier 2. 
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

[[Page 12923]]

B. The applicant for certification of the AP1000 design is 
Westinghouse Electric Company LLC.
---------------------------------------------------------------------------

    \1\ AP1000 is a trademark of Westinghouse Electric Company LLC.
---------------------------------------------------------------------------

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 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 10 CFR 52.47, with the exception of 
generic TS, the design-specific PRA, the evaluation of SAMDAs, and 
conceptual design information;
    2. Information required for a final safety analysis report under 
10 CFR 50.34;
    3. 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
    4. COL action items (COL information), which identify certain 
matters that must be addressed in the site-specific portion of the 
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.
    5. 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 SAMDAs 
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 license that wishes to reference this 
appendix shall, in addition to complying with the requirements of 10 
CFR 52.77, 52.78, and 52.79, 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. Physically 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,

[[Page 12924]]

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 SAMDAs associated with 
the information in the NRC's EA for the AP1000 design and Appendix 
1B of the generic DCD, for plants referencing this appendix whose 
site parameters are within those specified in the SAMDA 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 52.97(b). 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;

[[Page 12925]]

    (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 a 
severe accident issue identified in the plant-specific DCD, requires 
a license amendment if:
    (1) There is a substantial increase in the probability of a 
severe accident such that a particular 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 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 
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 TS 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 50.12(a). 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.

[[Page 12926]]

    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 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.

PART 54--REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR 
POWER PLANTS

    132. 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, 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.

    133. 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).
    134. 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.
* * * * *
    135. 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.
* * * * *
    136. 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

[[Page 12927]]

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.
    137. In Sec.  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.
* * * * *
    138. 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.
    139. 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

    140. 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).

    141. 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,
* * * * *
    142. 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,
* * * * *
    143. 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 Nuclear Reactor Regulation has delegated to the Regional 
Administrators of Regions I, II, III, and IV authority and 
responsibility under the regulations in this part for the 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 Nuclear Reactor Regulation 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

    144. 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).

    145. 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.
    146. 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

[[Page 12928]]

general license will be removed from the reactor site.
* * * * *

PART 73--PHYSICAL PROTECTION OF PLANTS AND MATERIALS

    147. 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).

    148. 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,
* * * * *
    149. 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.
* * * * *
    150. 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:
* * * * *
    151. 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.
* * * * *
    152. 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 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.
* * * * *
    153. 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

    154. 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).

    155. 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 the Director, Office of Nuclear Reactor 
Regulation. For all other installations, the cognizant Director is the 
Director, Office of Nuclear Material Safety and Safeguards.
* * * * *

[[Page 12929]]

PART 95--FACILITY SECURITY CLEARANCE AND SAFEGUARDING OF NATIONAL 
SECURITY INFORMATION AND RESTRICTED DATA

    156. 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.

    157. 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.
* * * * *
    158. 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.
    159. 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--
* * * * *
    160. 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.
    161. 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.
    162. Section 95.31 is revised to read as follows:


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.
    163. 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.
* * * * *
    164. 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.
    165. 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 matters 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.
* * * * *
    166. 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

[[Page 12930]]

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.
    167. 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 
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.
* * * * *
    168. 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.
* * * * *
    169. 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.
* * * * *
    170. 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.
    171. 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.
    172. 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

[[Page 12931]]

individual and take the action necessary to preclude that individual 
having further access to the information.
    173. 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.
    174. 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:
* * * * *
    175. 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

    176. 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).

    177. 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.
* * * * *
    178. 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 to possess 
and use plutonium in a plutonium processing and fuel fabrication plant.
    179. 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.
    180. 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.
* * * * *
    181. 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 parts 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.
    182. 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
* * * * *
    183. 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

[[Page 12932]]

indemnified with respect to such facilities.
* * * * *
    184. 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.
* * * * *

    185. 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

    186. 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).

    187. 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 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 MATERIAL LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF 
COMPLIANCE, REGISTRATIONS, AND QUALITY ASSURANCE PROGRAM APPROVALS 
AND GOVERNMENT AGENCIES LICENSED BY NRC

    188. 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).

    189. 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 on-site; 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 22nd day of February, 2006.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
[FR Doc. 06-1856 Filed 3-10-06; 8:45 am]
BILLING CODE 7590-01-P