[Federal Register Volume 62, Number 91 (Monday, May 12, 1997)]
[Rules and Regulations]
[Pages 25800-25831]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-11968]
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NUCLEAR REGULATORY COMMISSION
10 CFR Part 52
RIN 3150--AE87
Standard Design Certification for the U.S. Advanced Boiling Water
Reactor Design
AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule.
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SUMMARY: The Nuclear Regulatory Commission (NRC or Commission) is
amending its regulations to certify the U.S. Advanced Boiling Water
Reactor (ABWR) design. The NRC is adding a new provision to its
regulations that approves the U.S. ABWR design by rulemaking. This
action is necessary so that applicants for a combined license that
intend to construct and operate the U.S. ABWR design may do so by
appropriately referencing this regulation. The applicant for
certification of the U.S. ABWR design was GE Nuclear Energy.
EFFECTIVE DATE: The effective date of this rule is June 11, 1997. The
incorporation by reference of certain publications listed in the
regulations is approved by the Director of the Federal Register as of
June 11, 1997.
FOR FURTHER INFORMATION CONTACT: Jerry N. Wilson, Office of Nuclear
Reactor Regulation, telephone (301) 415-3145 or Geary S. Mizuno, Office
of the General Counsel, telephone (301) 415-1639, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. Background.
II. Public comment summary and resolution.
A. Principal Issues.
1. Finality.
2. Tier 2 Change Process.
3. Need for Additional Applicable Regulations.
B. Responses to specific requests for comment from proposed
rule.
C. Other Issues.
1. NRC Verification of ITAAC Determinations.
2. DCD Introduction.
3. Duplicate documentation in design certification rule.
4-7. OCRE comments.
III. Section-by-section discussion.
A. Introduction (Section I).
B. Definitions (Section II).
C. Scope and contents (Section III).
D. Additional requirements and restrictions (Section IV).
E. Applicable regulations (Section V).
F. Issue resolution (Section VI).
G. Duration of this appendix (Section VII).
H. Processes for changes and departures (Section VIII).
I. Inspections, tests, analyses, and acceptance criteria
(Section IX).
J. Records and Reporting (Section X).
IV. Finding of no significant environmental impact: availability.
V. Paperwork Reduction Act statement.
VI. Regulatory analysis.
VII. Regulatory Flexibility Act certification.
VIII. Backfit analysis.
I. Background
On September 29, 1987, General Electric Company applied for
certification of the U.S. ABWR standard design with the NRC. The
application was made in accordance with the procedures specified in 10
CFR Part 50, Appendix O, and the Policy Statement on Nuclear Power
Plant Standardization, dated September 15, 1987. The application was
docketed on February 22, 1988 (Docket No. STN 50-605).
The NRC added 10 CFR Part 52 to its regulations to provide for the
issuance of early site permits, standard design certifications, and
combined licenses for nuclear power reactors. Subpart B of 10 CFR Part
52 established the process for obtaining design certifications. A major
purpose of this rule was to achieve early resolution of licensing
issues and to enhance the safety and reliability of nuclear power
plants.
[[Page 25801]]
On December 20, 1991, GE Nuclear Energy (GE), an operating
component of General Electric Company's power systems business,
requested that its application, originally submitted pursuant to 10 CFR
Part 50, Appendix O, be considered as an application for design
approval and subsequent design certification pursuant to Subpart B of
10 CFR Part 52. Notice of receipt of this request was published in the
Federal Register on March 20, 1992 (57 FR 9749), and a new docket
number (52-001) was assigned.
The NRC staff issued a final safety evaluation report (FSER)
related to the certification of the U.S. ABWR design in July 1994
(NUREG-1503). The FSER documents the results of the NRC staff's safety
review of the U.S. ABWR design against the requirements of 10 CFR Part
52, Subpart B, and delineates the scope of the technical details
considered in evaluating the proposed design. Subsequently, the
applicant submitted changes to the U.S. ABWR design and the NRC staff
evaluated these design changes in a supplement to the FSER (NUREG-1503,
Supplement No. 1). A copy of the FSER and Supplement No. 1 may be
obtained from the Superintendent of Documents, U. S. Government
Printing Office, Mail Stop SSOP, Washington, DC 20402-9328 or the
National Technical Information Service, Springfield, VA 22161. A final
design approval (FDA) was issued for the U.S. ABWR design on July 13,
1994 and revised on November 23, 1994 to provide a 15 year duration. An
FDA, which incorporates the design changes, will be issued to supersede
the current FDA after issuance of this final design certification rule.
The NRC staff originally proposed a conceptual design certification
rule for evolutionary standard plant designs in SECY-92-287, ``Form and
Content for a Design Certification Rule.'' Subsequently, the NRC staff
modified the draft rule language proposed in SECY-92-287 to incorporate
Commission guidance and published a draft-proposed design certification
rule in the Federal Register on November 3, 1993 (58 FR 58665), as an
Advanced Notice of Proposed Rulemaking (ANPR) for public comment. In
accordance with the Administrative Procedure Act of 1947 (APA), as
amended, 10 CFR Part 52 provides the opportunity for the public to
submit written comments on proposed design certification rules.
However, Part 52 went beyond the requirements of the APA by providing
the public with an opportunity to request a hearing before an Atomic
Safety and Licensing Board in a design certification rulemaking.
Therefore, on April 7, 1995 (60 FR 17902), the NRC published a proposed
rule in the Federal Register which invited public comment and provided
the public with the opportunity to request an informal hearing before
an Atomic Safety and Licensing Board. The period within which an
informal hearing could be requested expired on August 7, 1995. The NRC
did not receive any requests for an informal hearing during this
period. The NRC staff conducted public meetings on the development of
this design certification rule on November 23, 1993, May 11 and
December 4, 1995, and May 2 and July 15, 1996, in order to enhance
public participation.
The Commission has considered the comments received and made
appropriate modifications to this design certification rule, as
discussed in Sections II and III, and revised the numbering system used
in the proposed rule. With these modifications, the Commission adopts
as final this design certification rule, Appendix A to 10 CFR Part 52,
for the U.S. ABWR design.
II. Public Comment Summary and Resolution
The public comment period for the proposed design certification
rule, the design control document, and the environmental assessment for
the U.S. ABWR design expired on August 7, 1995. The NRC received twenty
letters containing public comments on the proposed rule. The most
extensive comments were provided by the Nuclear Energy Institute (NEI),
in a letter dated August 4, 1995, which provided comments on behalf of
the nuclear industry. In general, NEI commended the NRC for its efforts
to provide standard design certifications but expressed serious
concerns about aspects of the proposed rule that would, in NEI's view,
undermine the goals of design certification. These concerns are
addressed in the following responses to the public comments. Fourteen
utilities and three vendors also provided comments. All of these
comment letters endorsed the NEI comments of August 4, 1995, and some
provided additional comments. The Department of Energy and the Ohio
Citizens for Responsible Energy, Inc. (OCRE) also submitted comment
letters. OCRE provided two sets of comments, the first addressed the
NRC's specific requests for comment and the second addressed OCRE's
concerns about certain aspects of the U.S. ABWR design.
The NRC received other letters that were entered into the docket
and are part of the record of the rulemaking proceeding, including an
August 4, 1995 letter from NEI to the Chairman of the NRC, which
submitted a copy of the Executive Summary of their public comment
letter, and a May 11, 1995 letter, which provided suggestions on
finality, secondary references, and other explanatory material. Also,
the NRC received a second letter from the General Electric Company,
which commented on the comments provided by OCRE.
On February 6, 1996, the NRC staff issued SECY-96-028, ``Two Issues
for Design Certification Rules,'' which requested the Commission's
approval of the staff's position on two major issues raised by NEI in
its comments on the proposed design certification rules. The NRC staff
issued this paper because of fundamental disagreements with the nuclear
industry on the need for applicable regulations and the matters to be
considered in verifying inspections, tests, analyses, and acceptance
criteria (ITAAC). Both NEI and DOE commented on SECY-96-028 in letters
dated March 5 and 13, 1996, respectively.
On March 8, 1996, the Commission conducted a public meeting in
which industry representatives and NRC staff presented their views on
SECY-96-028. During this meeting, NEI and the NRC staff both indicated
agreement on the ITAAC verification issue. Subsequently, in a staff
requirements memorandum (SRM) dated March 21, 1996, the Commission
requested the NRC staff to meet again with industry to try to resolve
the issue of applicable regulations. The NRC staff met with
representatives of Combustion Engineering, Inc. (ABB-CE), GE, and NEI
in a public meeting on March 25, 1996 and were unable to reach
agreement. As a result, the NRC staff provided revised resolutions of
applicable regulations and ITAAC determinations in SECY-96-077,
``Certification of Two Evolutionary Designs,'' dated April 15, 1996,
that superseded the proposals in SECY-96-028. SECY-96-077 addressed the
comments on the proposed design certification rules and provided final
design certification rules for the Commission's consideration.
Subsequently, notice of a 30 day comment period for SECY-96-077 was
published in the Federal Register (61 FR 18099), and the comment period
was extended for an additional 60 days (61 FR 27027) at the request of
NEI.
In response to the supplementary comment period, ABB-CE, GE Nuclear
Energy, and NEI submitted additional comments on the final design
certification rules in letters dated July 23, 1996. Westinghouse also
submitted comments in a letter dated July 24,
[[Page 25802]]
1996. NEI sent an unsolicited letter, dated September 23, 1996, to the
Director of the Office of Nuclear Reactor Regulation on three design
certification issues. NEI also sent a letter, dated September 16, 1996,
to Chairman Jackson that provided additional information in response to
questions that were asked by the Commission in its August 27, 1996
briefing on design certification rulemaking.
The following discussion is separated into three groups: (1)
Resolution of the principal issues raised by the commenters, (2)
resolution of the NRC's specific requests for comment from the proposed
rule, and (3) resolution of other issues raised by the commenters.
A. Principal Issues
1. Finality
Comment Summary. The applicant and NEI submitted extensive comments
on the scope of issues that were proposed to be accorded finality under
10 CFR 52.63(a)(4), i.e. are not subject to re-review by the NRC or re-
litigation in hearings. In summary, both commenters argued that:
The scope of issues accorded finality is too narrow;
Changes made in accordance with the change process are not
accorded finality;
Changes approved by the NRC should have protection under
10 CFR 52.63(a)(4);
The rule does not provide finality in all subsequent
proceedings;
The rule should be clarified regarding finality of SAMDA
evaluations;
A de novo review is not required for design certification
renewal;
Finality for Technical Specifications; and
Finality for Operational Requirements.
These comments are found in GE Comments dated August 3, 1995,
Attachment A, pp. 2-4; NEI Comments dated August 4, 1995, Attachment B,
pp. 1-23; NEI Comments dated July 23, 1996, pp. 1-21; and NEI letter
dated September 16, 1996.
Response: Scope of issues accorded finality. The applicant and NEI
took issue with the proposed rule's language limiting the scope of
nuclear safety issues resolved to those issues ``associated with'' the
information in the FSER or Design Control Document (DCD). Each argued
that there were many other documents which included and/or addressed
issues whose status should be regarded as ``resolved in connection
with'' this design certification rulemaking. These additional documents
include ``secondary references'' (i.e., DCD references to documents and
information which are not contained in the DCD, including secondary
references containing proprietary and safeguards information), docketed
material, and the entire rulemaking record (refer to GE Comments,
Attachment A, pp. 2-3; NEI Comments dated August 4, 1995, Attachment B,
pp. 6-9).
The Commission has reconsidered its position and decided that the
ambit of issues resolved by this rulemaking should be the information
that is reviewed and approved in the design certification rulemaking,
which includes the rulemaking record for the standard design. This
position reflects the Commission's SRM on SECY-90-377, dated February
15, 1991. Also, the Commission concludes that the set of issues
resolved should be those that were addressed (or could have been
addressed if they were considered significant) as part of the design
certification rulemaking process. However, the Commission does not
agree that all matters submitted on the docket for design certification
should be accorded finality under 10 CFR 52.63(a)(4). Some of this
information was neither reviewed nor approved and some was not directly
related to the scope of issues resolved by this rulemaking. Therefore,
the final rule provides finality for all nuclear safety issues
associated with the information in the FSER and Supplement No. 1, the
generic DCD, including referenced information that is intended as
requirements, and the rulemaking record.
In adopting this final design certification rulemaking, the
Commission also finds that the design certification does not require
any additional or alternative design criteria, design features,
structures, systems, components, testing, analyses, acceptance
criteria, or additional justifications in support of these matters.
Inherent in the concept of design certification by rulemaking is that
all these issues which were addressed, or could have been addressed, in
this rulemaking are resolved and therefore, may not be raised in a
subsequent NRC proceeding. If this were not the case and one could
always argue in a subsequent proceeding that an additional,
alternative, or modified system, structure or component of a
previously-certified design was needed, or additional justification was
necessary, or a modification to the testing and acceptance criteria is
necessary, there would be little regulatory certainty and stability
associated with a design certification. The underlying benefits of
certification of individual designs by rulemaking, e.g., early
Commission consideration and resolution of design issues and early
Commission consideration and agreement on the methods and criteria for
demonstrating completion of detailed design and construction in
compliance with the certified design, would be virtually negated. Thus,
in accord with the views of the applicant and NEI, the Commission
clarifies and makes explicit its previously implicit determination that
the scope of issues resolved in connection with the design
certification rulemaking includes the lack of need for alternative,
additional or modified design criteria, design features, structures,
systems, components, or inspections, tests, analyses, acceptance
criteria or justifications, and such matters may not be raised in
subsequent NRC proceedings.
In the statements of consideration (SOC) for the proposed rule, the
Commission proposed that issues associated with ``requirements'' in
secondary references, not specifically approved for incorporation by
reference by the Office of the Federal Register (OFR) because they
contained proprietary or safeguards information, would not be
considered resolved in the design certification rulemaking within the
meaning of 10 CFR 52.63(a)(4) (See 60 FR 17902, 17911). Both GE and NEI
took exception to this position, arguing that issues arising from
secondary references should be included in the set of issues resolved
(See GE Comments, Attachment A, pp. 2-3; NEI Comments dated August 4,
1995, Attachment B, pp. 6-9). The Commission has determined that the
set of issues resolved by this rulemaking embraces those issues arising
from secondary references that are requirements for the certified
design, including those containing proprietary and safeguards
information. This is consistent with the intent of 10 CFR Part 52 that
issues related to the design certification should be considered and
resolved in the design certification rulemaking. However, since OFR
does not approve of ``incorporation by reference'' of proprietary and
safeguards information, even though it was available to potential
commenters on this proposed design certification rule (see 60 FR 17902
at 17920-21; April 7, 1995), the Commission has included in VI.E of
this appendix, a process for obtaining proprietary and safeguards
information at the time that notice of a hearing in connection with
issuance of
[[Page 25803]]
a combined license is published in the Federal Register. Such persons
will have actual notice of the requirements contained in the
proprietary and safeguards information and, therefore, will be subject
to the issue finality provisions of Section VI of this appendix.
Changes made in accordance with the ``50.59-like'' change process.
The proposed design certification rule included a change process
similar to that provided in 10 CFR 50.59. Specifically, proposed
Section 8(b)(5) provided ``that such changes open the possibility for
challenge in a hearing'' for Tier 2 changes in accordance with the
Commission's guidance in its SRM on SECY-90-377, dated February 15,
1991. The NRC also believed that providing an opportunity for a hearing
would serve to discourage changes that could erode the benefits of
standardization. The applicant and NEI argued that Tier 2 departures
under the ``Sec. 50.59-like'' process should not be subject to any
opportunity for hearing but may only be challenged via a 10 CFR 2.206
petition; and, therefore, should be subject to the special backfit
restrictions of 10 CFR 52.63(a). For purposes of brevity, this
discussion refers to both generic changes and plant-specific departures
as ``changes.''
The Commission has reconsidered and revised its position on issue
resolution in connection with Tier 2 departures under the ``Sec. 50.59-
like'' process. Section 50.59 was originally adopted by the Commission
to afford a Part 50 operating license holder greater flexibility in
changing the facility as described in the FSAR while still assuring
that safety-significant changes of the facility would be subject to
prior NRC review and approval [refer to 27 FR 5491, 5492 (first
column); June 9, 1962]. The ``unreviewed safety question'' definition
was intended by the Commission to exclude from prior regulatory
consideration those licensee-initiated changes from the previously NRC-
approved FSAR that could not be viewed as having safety significance
sufficient to warrant prior NRC licensing review and approval. To put
it another way, any change properly implemented pursuant to Sec. 50.59
should continue to be regarded as within the envelope of the original
safety finding by the NRC. Moreover, the departure process for Tier 2
information, as specified in VIII.B of this appendix, includes
additional restrictions derived from 10 CFR 52.63(b)(2), viz., the Tier
2 change must not involve a change to Tier 1 information. Thus, the
departure process (VIII.B.5), if properly implemented by an applicant
or licensee, must logically result in departures which are both
``within the envelope'' of the Commission's safety finding for the
design certification rule and for which the Commission has no safety
concern. Therefore, it follows that properly implemented departures
from Tier 2 should continue to be accorded the same extent of issue
resolution as that of the original Tier 2 information from which it was
``derived.'' As a result, Section VI of this appendix has been amended
to reflect the Commission's determination on issue resolution for Tier
2 changes made in accordance with the departure process and to provide
backfit protection for changes made in accordance with the processes of
Section VIII of this appendix.
However, the converse of this reasoning leads the Commission to
reject the applicant's and NEI's contention that no part of the
applicant's or licensee's implementation of the departure process
(VIII.B.5) should be open to challenge in a subsequent licensing
proceeding, but instead should be raised as a petition for enforcement
action under 10 CFR 2.206. Because Sec. 2.206 applies to holders of
licenses and is considered a request for enforcement action (thereby
presenting some potential difficulties when attempting to apply this in
the context of a combined license applicant), it is unclear why an
applicant or licensee who departs from the design certification rule in
noncompliance with the process (VIII.B.5) should nonetheless reap the
benefits of issue resolution stemming from the design certification
rule. An incorrect departure from the requirements of this appendix
essentially places the departure outside of the scope of the
Commission's safety finding in the design certification rulemaking. It
follows that properly-founded contentions alleging such incorrectly-
implemented departures cannot be considered ``resolved'' by this
rulemaking. The industry also appears to oppose an opportunity for a
hearing on the basis that there is no ``remedy'' available to the
Commission in a licensing proceeding that would not also constitute a
violation of the Tier 2 backfitting restrictions applicable to the
Commission and that in a comparable situation with an operating plant
the proper remedy is enforcement action. However, for purposes of issue
finality the focus should be on the initial licensing proceeding where
the result of an improper change evaluation would simply be that the
change is not considered resolved and no enforcement action is needed.
Neither the applicant nor NEI provided compelling reasons why
contentions alleging that applicants or licensees have not properly
implemented the departure process (VIII.B.5) should be entirely
precluded from consideration in an appropriate licensing proceeding
where they are relevant to the subject of the proceeding.
Although the Commission disagrees with the applicant and NEI over
the admissibility of contentions alleging incorrect implementation of
the departure process, the Commission acknowledges that they have a
valid concern regarding whether the scope of the contentions will
incorrectly focus on the substance of correctly-performed departures
and the possible lengthened time necessary to litigate such matters in
a hearing (See, e.g., Transcript of December 4, 1995, Public Meeting,
p. 47). Therefore, the Commission has included an expedited review
process (VIII.B.5.f), similar to that provided in 10 CFR 2.758, for
considering the admissibility of such contentions. Persons who seek a
hearing on whether an applicant has departed from Tier 2 information in
noncompliance with the applicable requirements must submit a petition,
together with information required by 10 CFR 2.714(b)(2), to the
presiding officer. If the presiding officer concludes that a prima
facie case has been presented, he or she shall certify the petition and
the responses to the Commission for final determination as to
admissibility.
Subsequently, in its comments dated July 23, 1996, NEI requested
the Commission to modify VIII.B.5.f to clarify that a ``50.59-like''
change is not subject to a hearing under Sec. 52.103 or Sec. 50.90
unless the change bears directly on an asserted ITAAC noncompliance or
the requested amendment, respectively. The Commission determined that
NEI's proposed wording correctly stated its intention regarding the
opportunity for a hearing on ``50.59-like'' departures after a license
is issued and, therefore, VIII.B.5.f of this appendix has been
appropriately modified.
Changes approved by the NRC should have protection under
Sec. 52.63. NEI, in its comments dated July 23, 1996, requested the
Commission to provide the special backfit protection of Sec. 52.63 to
all changes to Tier 1, Tier 2*, and changes to Tier 2 that involve an
unreviewed safety question or a change in the technical specifications.
The special provision in Sec. 52.63(a)(4) states that ``* * * the
Commission shall treat as resolved those matters resolved in connection
with the issuance or renewal of a design certification.'' The
Commission stated, in its SRM on
[[Page 25804]]
SECY-90-377, that ``* * * the process provides issue finality on all
information provided in the application that is reviewed and approved
in the design certification rulemaking.'' The Commission also stated
that ``* * * changes to the design reviewed and approved by the staff
should be minimized * * *.'' Based on this guidance, the Commission
decided that the special backfit provision should be extended to
generic changes made to the DCD that are approved by rulemaking. Also,
for departures that are approved by license amendment or exemption, the
Commission decided that the licensee of that plant should receive the
special backfit protection. However, any other licensee that references
the same DCD should not have finality for that plant-specific
departure, unless it was again approved by license amendment or
exemption for that licensee.
Finality in all subsequent proceedings. GE and NEI requested that
Section 6 of the proposed rule be expanded to include a more detailed
statement regarding the findings, issues resolved, and restrictions on
the Commission's ability to ``backfit'' this appendix. The Commission
agrees that the industry's proposal has some merit, and has revised
Section VI of this appendix, beginning with the general subjects
embodied in NEI's proposed redraft, but restructured the NEI proposal
into three sections to reflect the scope of issues resolved, change
process, and rulemaking findings, thereby conforming the language to
reflect the conventions of the appendix (e.g., generic changes versus
plant-specific departures), and making minor editorial changes for
clarity and consistency. However, one area in which the Commission
declines to adopt the industry's proposal is the inclusion of a
statement that extends issue finality to all subsequent proceedings.
Section 52.63(a)(4) explicitly states that issues resolved in a
design certification rulemaking have finality in combined license
proceedings, proceedings under Sec. 52.103, and operating license
proceedings. There are other NRC proceedings not mentioned in
Sec. 52.63(a)(4), e.g., combined license amendment proceedings and
enforcement proceedings, in which the design certification should
logically be afforded issue resolution and, therefore, are included in
Section VI of this appendix. However, NEI listed NRC proceedings such
as design certification renewal proceedings, for which issue finality
would not be appropriate. Moreover, it should be understood that to say
that this design certification rule is accorded ``issue finality'' does
not eliminate changes properly made under the change restrictions in
Section VIII of this appendix. Therefore, the Commission declines to
adopt in its entirety the industry proposal that issue finality should
extend to all subsequent NRC proceedings.
In its comments dated July 23, 1996, NEI requested the Commission
to modify the last phrase of Section 6(b), of SECY-96-077, to reflect
the NRC staff's intent regarding finality in enforcement proceedings.
Section 6(b) stated that the DCD has finality in enforcement
proceedings ``where these proceedings reference this appendix.'' NEI
was concerned that this phrase could be construed as depriving finality
to plants that reference the design certification rules in enforcement
proceedings that do not explicitly reference the design certification
rule. The intent of the phrase was to limit finality of the information
in the design certification rule to enforcement proceedings involving a
plant referencing the rule. Therefore, the Commission replaced the
wording, ``where these proceedings reference this appendix,'' with
``involving plants referencing this appendix'' in Section VI.B of the
final rules.
Finality regarding SAMDA evaluations. In its comments dated July
23, 1996, NEI requested the Commission to extend finality for the SAMDA
evaluation when an exemption from a site parameter specified in the
evaluation has been approved. Section VI.B.7 of this appendix accords
finality to severe accident mitigation design alternatives (SAMDAs) for
plants referencing the design certification rules ``whose site
parameters are within those specified in the Technical Support
Document'' (TSD). NEI is concerned that the last phrase could open all
SAMDAs to re-review and re-litigation during a subsequent proceeding
where the licensee has requested an exemption from a site parameter
specified in the DCD, even though the exemption has no impact on the
SAMDAs. NEI also stated that a clarification to the SOC was not
sufficient and believed that a modification to the rule language was
needed.
The NRC staff agrees that it was not the intent to re-litigate
SAMDA issues under such circumstances. The intent was that an
intervenor in any subsequent proceeding could challenge a SAMDA based
on an exemption to a TSD site parameter only after bringing forward
evidence demonstrating that the SAMDA analysis was invalidated.
However, the NRC staff does not agree that the wording should be
changed. NEI's proposed modification would shift the burden of
demonstrating the acceptability of the exemption from the licensee.
Moreover, it would be difficult to extend the NEPA review to all
available sites without any qualification. Therefore, the Commission
decided not to change Section VI.B.7 of this appendix but did explain
in section III.F of this SOC that requests for litigation must meet
Sec. 2.714 requirements.
A de novo review is not required for design certification renewal.
In its comments dated July 23, 1996, NEI requested the Commission to
extend finality to design certification renewal proceedings and to
define a review procedure for renewal applications that would limit the
scope of review. Subsequently, NEI stated in a letter dated September
23, 1996, that principles for renewal reviews can and should be
established in the design certification rules. The extension of
finality to a renewal proceeding would produce the illogical result
that the NRC's conclusion in the original design certification
rulemaking, that the design provided adequate protection and was in
compliance with the applicable regulations, would also apply to the
renewal review even though the regulations in Part 52 require another
review and finding at the renewal stage 15 years later. The effect of
this extension would be to extend the design certification for another
15 years (for a total of 30 years) instead of the intended 15 years.
The NRC staff agrees with NEI that the renewal review must be
conducted against the Commission's regulations applicable and in effect
at the time of the original certification, and that the backfit
limitations in Sec. 52.59 must be satisfied in order to require a
change to the certified design. However, the NRC staff disagrees with
NEI's position that the information to be considered in the renewal
review is limited to ``an evaluation of experience between the time of
certification and the renewal application,'' as well as NEI's
implication that the scope of the design for which new information can
be considered is limited to those areas which the design certification
applicant concedes there is new information or proposes a modification.
The effect of NEI's position would be to preclude the NRC from
considering new information which could have altered the Commission's
consideration and approval of the design had it been known at the time
of the original certification review, and to cede control of the scope
of the renewal review to the design certification applicant.
Furthermore, the review procedure for a
[[Page 25805]]
renewal application is not dependent on whether the applicant proposed
changes to the previously certified design. The underlying philosophy
was that new safety requirements and issues that arose during the
duration of the design certification rule could not be applied to the
certified design (unless the adequate protection standard was met).
However, these issues could be raised for consideration at the renewal
stage and applied to the application for renewal if the backfit
standard in Sec. 52.59 was met. Therefore, any portion of the certified
design could be reviewed (subject to Sec. 52.59) to ensure that the
applicable regulations for the certified design are being met based on
consideration of new information (e.g. operating experience, research,
or analysis) resulting from the previous 15 years of experience with
the design.
The Commission rejects NEI's proposal to apply the finality
provision of Sec. 52.63 to the review of renewal applications because
this would suggest improperly that NRC, in its renewal review, is bound
by previous safety conclusions in the initial certification review. The
type of renewal review was resolved by the Commission during the
development of 10 CFR Part 52. At that time, the Commission determined
that the backfit standard in Sec. 52.59(a) controls the development of
new requirements during the review of applications for renewal.
Therefore, the Commission disagrees with NEI's proposed revision to
Section 6(b), in its letter dated September 23, 1996, and NEI's
proposal for a new Section 6(e) is unnecessary because this process is
already correctly covered in Sec. 52.59.
The Commission does not plan or expect to be able to conduct a de-
novo review of the entire design if a certification renewal application
is filed under Sec. 52.59. It expects that the review focus would be on
changes to the design that are proposed by the applicant and insights
from relevant operating experience with the certified design or other
designs, or other material new information arising after the NRC
staff's review of the design certification. The Commission will defer
consideration of specific design certification renewal procedures until
after it has issued this appendix.
Finality for Technical Specifications. In its comments dated August
4, 1995, Attachment B (pp. 124-129), NEI requested that the NRC
establish a single set of integrated technical specifications governing
the operation of each plant that references this design certification
and that the technical specifications be controlled by a single change
process. In the proposed rule, the NRC included the technical
specifications for the standard designs in the generic DCD in order to
maximize the standardization of the technical specifications for plants
that reference this design certification. As a result, a plant that
references this design certification would have two sets of technical
specifications associated with its license: (1) Technical
specifications from Chapter 16 of Tier 2 of the generic DCD and
applicable to the standardized portion of the plant, and (2) those
technical specifications applicable to the site-specific portion for
the plant. While each portion of the technical specifications would be
subject to a different change process, the substantive aspects of the
change processes would be essentially the same.
In the design certification rule that was attached to SECY-96-077,
the technical specifications were removed from Tier 2 for two reasons.
First, the removal from Tier 2 responded to NEI's comment regarding a
single change process. NEI's proposal to include the technical
specifications in Tier 2 prior to issuance of a combined license (COL),
and then remove them after COL issuance is not acceptable. If the
technical specifications are included in Tier 2 by the design
certification rulemaking, they would remain there and be controlled by
the Tier 2 change process for the life of the facility. Second, the NRC
staff wanted the ability to impose future operational requirements and
standards (distinct from design matters) on the technical
specifications for a plant that referenced the certified design and
Section 4(c) of the rule in SECY-96-077 provided that ability. However,
Section 4(c) would not be used to backfit design features (i.e.
hardware changes) unless the criteria of Sec. 52.63 were met.
In its comments dated July 23, 1996, NEI requested the Commission
to extend finality to the technical specifications in Chapter 16 of the
DCD. NEI stated that the technical specifications in the DCDs should
remain part of the design certification and be accorded finality
because they have been reviewed and approved by the NRC. NEI also
proposed that, after the license is granted, the technical
specifications in the DCD would no longer have any relevance to the
license and there would be a single set of technical specifications
that will be controlled by the 10 CFR 50.90 license amendment process
and subject to the backfit provisions in 10 CFR 50.109.
The Commission does not support extension of the special backfit
provisions of Sec. 52.63 to technical specifications and other
operational requirements as requested by NEI, rather the Commission
supports the proposal to treat the technical specifications in Chapter
16 of the DCD as a special category of information, as described in the
NRC staff's comment analyses dated August 13 and October 21, 1996. The
purpose of design certification is to review and approve design
information. There is no provision in Subpart B of 10 CFR Part 52 for
review and approval of purely operational matters. The Commission
approves a revised Section VIII.C of this appendix that would apply to
the technical specifications, bases for the technical specifications,
and other operational requirements in the DCD; that would provide for
use of Sec. 52.63 only to the extent the design is changed; and that
would use Sec. 2.758 and Sec. 50.109 to the extent an NRC safety
conclusion is being modified or changed but no design change is
required. In applying Sec. 2.758 and Sec. 50.109, it will be necessary
to determine from the certification rulemaking record what safety
issues were considered and resolved. This is because Sec. 2.758 will
not bar review of a safety matter that was not considered and resolved
in the design certification rulemaking. There would be no backfit
restriction under Sec. 50.109 because no prior position was taken on
this safety matter. After the COL is issued, the set of technical
specifications for the COL (the combination of plant-specific and DCD
derived) would be subject to the backfit provisions in Sec. 50.109
(assuming no Tier 1 or Tier 2 changes are involved).
Finality for operational requirements. A new provision was included
in the design certification rules, set forth in Section 4(c), that were
attached to SECY-96-077. The reason for this provision was that the
operational requirements in the DCD had not received a complete and
comprehensive review. Therefore, the new Section 4(c) was needed to
reserve the right of the Commission to impose operational requirements
on plants referencing this appendix, such as license conditions for
portions of the plant within the scope of this design certification,
e.g. start-up and power ascension testing. NEI claimed, in its comments
dated July 23, 1996, that the backfit provisions in Section 4(c)
contradicted 10 CFR 52.63 and were incompatible with the purpose of 10
CFR Part 52.
NEI's claim that Section 4(c) contradicts 10 CFR 52.63 and enables
the NRC to impose changes to the design information in the DCD without
regard to the special backfit provisions of Sec. 52.63 is wrong.
Section 4(c) clearly referred to ``facility operation'' not ``facility
design.'' The purpose of
[[Page 25806]]
Section 4(c) was to ensure that any necessary operational requirements
could be applied to plants that reference these certified designs
because plant operational matters were not finalized in the design
certification review. It was also clear that the NRC staff considered
resolved design matters to be final. Refer to SECY-96-077 which states:
``Most importantly, a provision has been included in Section 4 to
provide that the final rules do not resolve any issues regarding
conditions needed for safe operation (as opposed to safe design).''
This is consistent with the goal of design certification, which is to
preserve the resolution of design features, which are explicitly
discussed or inferred from the DCD. The backfit provisions in Sections
VIII.A and VIII.B of this appendix control design changes.
Subsequently, in its comments of September 23, 1996, NEI requested
that all DCD requirements, including operational-related and other non-
hardware requirements, be accorded finality under Sec. 52.63. The
Commission has determined that NEI's proposal to assign finality to
operational requirements is unacceptable, because operational matters
were not comprehensively reviewed and finalized for design
certification (refer to section III.F of this SOC). Although the
information in the DCD that is related to operational requirements was
necessary to support the NRC's safety review of the standard designs,
the review of this information was not sufficient to conclude that the
operational requirements are fully resolved and ready to be assigned
finality under Sec. 52.63. Therefore, the Commission retained the
former Section 4(c), but reworded this provision on operational
requirements and placed it in Section VI.C of this appendix with the
other provisions on finality (also refer to Section VIII.C of this
appendix).
2. Tier 2 Change Process
Comment Summary. NEI submitted many comments on the following
aspects of the Tier 2 change process:
Scope of the change process in VIII.B.5;
Post-design certification rulemaking changes to Tier 2
information;
Restrictions on Tier 2* information; and
Additional aspects of the change process.
Response. The proposed design certification rule provided a change
process for Tier 2 information that had the same elements as the Tier 1
change process in order to implement the two-tiered rule structure that
was requested by industry. Specifically, the Tier 2 change process in
Section 8(b) of the proposed rule provided for generic changes, plant-
specific changes, and exemptions similar to the provisions in 10 CFR
52.63, except that some of the standards for plant-specific orders and
exemptions are different. Section 8(b) also had a provision similar to
10 CFR 50.59 that allows for departures from Tier 2 information by an
applicant or licensee, without prior NRC approval, subject to certain
restrictions, in accordance with the Commission's SRM on SECY-90-377,
dated February 15, 1991.
Scope of the change process in VIII.B.5. In its comments dated
August 4, 1995, Attachment B, pp. 67-82, NEI raised a concern regarding
application of the Sec. 50.59-like change process to severe accident
information, and stated:
Instead of applying the Sec. 50.59-like process to all of
Chapter 19, we propose (1) that the process be applied only to those
sections that identify features that contribute significantly to the
mitigation or prevention of severe accidents (i.e., Section 19.8 for
the ABWR and Section 19.15 for the System 80+), and (2) that changes
in these sections should constitute unreviewed safety questions only
if they would result in a substantial increase in the probability or
consequences of a severe accident.
The Commission agrees that departures from Tier 2 information that
describe the resolution of severe accident issues should use criteria
that is different from the criteria in 10 CFR 50.59 for determining if
a departure constitutes an unreviewed safety question (USQ). Because of
the increased uncertainty in severe accident issue resolutions, the NRC
has included ``substantial increase'' criteria in VIII.B.5.c of this
appendix for Tier 2 information that is associated with the resolution
of severe accident issues. The (Sec. 50.59-like) criteria in VIII.B.5.b
of this appendix, for determining if a departure constitutes a USQ,
will apply to the remaining Tier 2 information. If the proposed
departure from Tier 2 information involves the resolution of other
safety issues in addition to the severe accident issues, then the USQ
determination must be based on the criteria in VIII.B.5.b of this
appendix.
However, NEI misidentified the sections of the DCD that describe
the resolutions of the severe accident issues. Section 19.8 for the
U.S. ABWR and Section 19.15 for the System 80+ design identify
important features that were derived from various analyses of the
design, such as seismic analyses, fire analyses, and the probabilistic
risk assessment. This information was used in preparation of the Tier 1
information and, as stated in the proposed rule, it should be used to
ensure that departures from Tier 2 information do not impact Tier 1
information. For these reasons, the Commission rejects the contention
that the severe accident resolutions are contained in Section 19.8 of
the generic DCD.
Subsequently, in its comments dated July 23, 1996, NEI requested
the Commission to expand the scope of design information that is
controlled by the special change process for severe accident issues to
all of the information in Chapter 19 of the DCD. The NRC staff intended
that this special change process be limited to severe accident design
features, where the intended function of the design feature is relied
upon to resolve postulated accidents when the reactor core has melted
and exited the reactor vessel and the containment is being challenged
(severe accidents). These design features are identified in Section
19.11 of the System 80+ DCD and Section 19E of the ABWR DCD. This
special change process was not intended for design features that are
discussed in Chapter 19 for other reasons, such as resolution of
generic safety issues. However, the NRC staff recognizes that the
severe accident design features identified in Section 19E are described
in other areas of the DCD, i.e. the Lower Drywell Flooder is described
in Section 9.5.12 of the ABWR DCD. Therefore, the location of design
information is not important to the application of the special change
process for severe accident issues and it is not specified in Section
VIII.B.5. The importance of this provision is that it be limited to the
severe accident design features. In addition, the Commission is
cognizant of certain design features that have intended functions to
meet ``design basis'' requirements and to resolve ``severe accidents.''
These design features will be reviewed under either VIII.B.5.b or
VIII.B.5.c depending upon the design function being changed. Finally,
the Commission rejects NEI's request to expand the scope of design
information that is controlled by the special change process for severe
accident issues.
Post-design certification rulemaking changes to Tier 2 information.
In its comments dated August 4, 1995, Attachment B, pp. 83-89, NEI
requested that the NRC add a Sec. 50.59-like provision to the change
process that would allow design certification applicants to make
generic changes to Tier 2 information prior to the first license
application. These applicant-initiated, post-certification Tier 2
changes would be binding upon all referencing applicants and licensees
(i.e., referencing applicants and
[[Page 25807]]
licensees must comply with all such changes) and would continue to
enjoy ``issue preclusion'' (i.e., issues with respect to the adequacy
of the change could not be raised in a subsequent proceeding as a
matter of right). However, the changes would not be subject to public
notice and comment. Instead NEI proposed that the changes would be
considered resolved and final (not subject to further NRC review) six
months after submission, unless the NRC staff informs the design
certification applicant that it disagrees with the determination that
no unreviewed safety question exists.
The Commission declines to adopt the NEI proposal. The applicant-
initiated Tier 2 changes proposed by NEI have the essential attributes
of a ``rule,'' and the process of NRC review and ``approval'' (negative
consent) would appear to be ``rulemaking,'' as these terms are defined
in Section 551 of the APA. Section 553(b) of the APA requires public
notice in the Federal Register and an opportunity for public comment
for all rulemakings, except in certain situations delineated in Section
553(b)(A) and (B) which are not applicable to applicant-initiated
changes. The NEI proposal conflicts with the rulemaking requirements of
the APA. If the NEI proposal is based upon a desire to permit the
applicant to disseminate worthwhile Tier 2 changes, there are three
alternatives already afforded by Part 52 and this appendix. The
applicant (as any member of the public) may submit a petition for
rulemaking pursuant to Subpart H of 10 CFR Part 2, to modify this
design certification rule to incorporate the proposed changes to Tier
2. If the Commission grants the petition and adopts a final rule, the
change is binding on all referencing applicants and licensees in
accordance with VIII.B.2 of this appendix. Also, the applicant could
develop acceptable documentation to support a Tier 2 departure in
accordance with VIII.B of this appendix. This documentation could be
submitted for NRC staff review and approval, similar to the manner in
which the NRC staff reviews topical reports. 1 Finally, the
applicant could provide its proposed changes to a COL applicant who
could seek approval as part of its COL application review. The
Commission regards these regulatory approaches to be preferable to the
NEI proposal. However, if NEI is requesting that the Commission change
its preliminary determination, as set forth in its February 15, 1991
SRM on SECY-90-377, that generic Tier 2 rulemaking changes be subject
to the same restrictive standard as generic Tier 1 changes, the
Commission declines to do so. The Commission believes that maintaining
a high standard for generic changes to both Tier 1 and Tier 2 will
ensure that the benefits of standardization are appropriately achieved.
---------------------------------------------------------------------------
\1\ Topical reports, which are usually submitted by vendors such
as GE, Westinghouse, and Combustion Engineering, request NRC staff
review and approval of generic information and approaches for
addressing one or more of the Commission's requirements. If the
topical report is approved by the NRC staff, it issues a safety
evaluation setting forth the bases for the staff's approval together
with any limitations on referencing by individual applicants and
licensees. Applicants and licensees may incorporate by reference
topical reports in their applications, in order to facilitate timely
review and approval of their applications or responses to requests
for information. However, limitations in NRC resources may affect
review schedules for these topical reports.
---------------------------------------------------------------------------
Subsequently, in its comments dated July 23, 1996, NEI requested
the Commission to modify this SOC to reflect NRC openness to discuss a
post-design certification change process and related issues after the
design certification rules are completed. The Commission has determined
that vendors who submit a design, which is subsequently certified by
rulemaking, may not make changes under a ``50.59-like'' process and
that NEI's request is outside the scope of this rulemaking. The
Commission believes that vendors should be limited in making changes to
rulemaking to amend the certification and that this appendix provides
an appropriate process for making generic changes to the DCD (refer to
the SRM on SECY-90-377 and the SOC for 10 CFR Part 52, Section II.1.h).
This process is available to everyone and the standard for changes is
the same for NRC, the applicant, and the public. This restrictive
change process is consistent with the NRC's goal of achieving and
preserving resolutions of safety issues to provide a stable and
predictable licensing process.
Restrictions on Tier 2* information. In its comments dated August
4, 1995, Attachment B, pp. 119-123, and in subsequent comments dated
July 23, 1996, pp. 50-54, NEI requested that the restriction on
departures from all Tier 2* information expire at first full power and,
in any event, the expiration of the restrictions should be consistent
for both the U.S. ABWR and System 80+ designs. The Commission stated in
the proposed design certification rule that the restriction on changing
Tier 2* information resulted from the development of the Tier 1
information in the generic DCD. During the development of the Tier 1
information, the applicant for design certification requested that the
amount of information in Tier 1 be minimized to provide additional
flexibility for an applicant or licensee who references this design
certification. Also, many codes, standards, and design processes, which
were not specified in Tier 1, that are acceptable for meeting ITAAC
were specified in Tier 2. The result of these actions is that certain
significant information only exists in Tier 2 and the Commission does
not want this significant information to be changed without prior NRC
approval. This Tier 2* information is identified in the generic DCD
with italicized text and brackets.
Although the Tier 2* designation was originally intended to last
for the lifetime of the facility, like Tier 1 information, the NRC
staff reevaluated the duration of the change restriction for Tier 2*
information during the preparation of the proposed rule. The NRC staff
determined that some of the Tier 2* information could expire when the
plant first achieves full (100%) power, after the finding required by
10 CFR 52.103(g), while other Tier 2* information must remain in effect
throughout the life of the plant that references this rule. The
determining factors were the Tier 1 information that would govern these
areas after first full power and the NRC staff's judgement on whether
prior approval was required before implementation of the change due to
the significance of the information.
As a result of NEI's comments, the NRC again reevaluated the
duration of the Tier 2* change restrictions. The NRC agrees with NEI
that expiration of Tier 2* information for the two evolutionary designs
should be consistent, unless there is a design-specific reason for a
different treatment. The NRC decided that the Tier 2* restrictions for
equipment seismic qualification methods and piping design acceptance
criteria could expire at first full power, because the approved
versions of the ASME code provide sufficient control of Tier 2* changes
for these two areas. Also, the Tier 2* restriction for the ABWR human
factors engineering design and implementation process can expire at
first full power because the NRC staff concluded that step 6 of the
Tier 1 implementation process requires that any changes made to the
Main Control Room and Remote Shutdown System conform with the Human-
System Design Implementation Process. However, the fuel design
evaluation information and the licensing acceptance criteria for fuel
must remain
[[Page 25808]]
designated as Tier 2* in the U.S. ABWR DCD in order to clarify the
acceptance criteria for reviewing changes to the current fuel design.
As discussed in Section 4.2 of the U.S. ABWR FSER (NUREG-1503), the
criteria were based on previous work with GE Nuclear Energy to define
the licensing acceptance criteria for core reload calculations.
Recent industry proposals for currently operating core fuel designs
have indicated a desire to modify the fuel burnup limit design
parameter. However, operational experience with fuel with extended fuel
burnup has indicated that cores should not be allowed to operate beyond
the burnup limits specified in the generic DCDs without NRC approval.
This experience is summarized in a Commission memorandum from James M.
Taylor, ``Reactivity Transients and High Burnup Fuel,'' dated September
13, 1994, including Information Notice (IN) 94-64, ``Reactivity
Insertion Transient and Accident Limits for High Burnup Fuel,'' dated
August 31, 1994. Experimental data on the performance of high burnup
fuel under reactivity insertion conditions became available in mid-
1993. The NRC issued IN 94-64 and IN 94-64, Supplement 1, on April 6,
1995, to inform industry of the data. The unexpectedly low energy
deposition to initiation of fuel failure in the first test rod (at 62
GWd/MTU) led to a re-evaluation of the licensing basis assumptions in
the NRC's standard review plan (SRP). The NRC performed a preliminary
safety assessment and concluded that there was no immediate safety
issue for currently operating cores because of the low to medium burnup
status of the fuel (refer to Commission Memorandum from James M.
Taylor, ``Reactivity Transients and Fuel Damage Criteria for High
Burnup Fuel,'' dated November 9, 1994, including an NRR safety
assessment and the joint NRR/RES action plan). Therefore, the NRC has
determined that additional actions by industry are not needed to
justify current burnup limits for operating reactor fuel designs.
However, the NRC has determined that it needs to carefully consider any
proposed changes to the fuel burnup parameter in the generic DCDs for
these fuel designs until further experience is gained with extended
fuel burnup characteristics. Requests for extension of these burnup
limits will be evaluated based on supporting experimental data and
analyses, as appropriate, for current and advanced fuel designs.
Therefore, the NRC has determined that the Tier 2* designation for the
fuel burnup parameters should not expire for the lifetime of a
referencing facility.
NEI also stated in its comments dated July 23, 1996, that to the
extent the Commission does not adopt its recommendation that all Tier
2* restrictions expire at first full power, the SOC should be modified
to reflect the NRC staff's intent that Tier 2* material in the DCD may
be superseded by information submitted with a license application or
amendment. The Commission decided that, if certain Tier 2* information
is changed in a generic rulemaking, the category of the new information
(Tier 1, 2*, or 2) would also be determined in the rulemaking and the
appropriate process for future changes would apply. If certain Tier 2*
information is changed on a plant-specific basis, then the appropriate
modification to the change process would apply only to that plant.
Additional aspects of the change process. In its comments dated
August 4, 1995, Attachment B, pp. 109-118, NEI raised some additional
concerns with the Tier 2 change process. The first concern was with the
process for determining if a departure from Tier 2 information
constituted an unreviewed safety question. Specifically, NEI identified
the following statement in section III.H of the SOC for the proposed
rule. ``* * * if the change involves an issue that the NRC staff has
not previously approved, then NRC approval is required.'' A
clarification of this statement was provided in the May 11, 1995 public
meeting on design certification (pp. 12-14 of meeting transcript), when
the NRC staff stated that the NRC was not creating a new criterion for
determining unreviewed safety questions but was explaining existing
criteria. A further discussion of this statement took place between the
staff and counsel to GE Nuclear Energy at the December 4, 1995 public
meeting on design certification (pp. 53-56 of meeting transcript), in
which counsel for GE Nuclear Energy agreed that a departure which
creates an issue that was not previously reviewed by the NRC would be
evaluated against the existing criteria for determining whether there
was an unreviewed safety question. The Commission does not believe
there is a need for a change to the language of this appendix. The
statement above was not included in section III.H of this SOC.
NEI also requested that Section 8(b) of the proposed rule be
revised to state that exemptions are not required for changes to the
technical specifications or Tier 2* information that do not involve an
unreviewed safety question. The Commission has determined that this is
consistent with the Commission's intent that permitted departures from
Tier 2* under VIII.B of this appendix should not also require an
exemption, unless otherwise required by, or implied by 10 CFR Part 52,
Subpart B and, accordingly, has revised paragraph VIII.B.6 of this
appendix. As discussed above, the technical specifications in Chapter
16 of the generic DCD are not in Tier 2 and, in its comments dated
September 23, 1996, NEI proposed that requested departures from Chapter
16 by an applicant for a COL require an exemption. The Commission
agrees with NEI's new position and included this provision in Section
VIII.C of this appendix. NEI also raised a concern with the requirement
for quarterly reporting of design changes during the construction
period. This issue is discussed in section III.J of this SOC.
Finally, NEI raised a concern with the status of 10 CFR 52.63(b)(2)
in the two-tiered rule structure that has been implemented in this
appendix and claimed that 10 CFR 52.63(b) clearly embodies a two-tier
structure. NEI's claim is not correct. The Commission adopted a two-
tiered design certification rule structure (Commission SRM on SECY-90-
377, dated February 15, 1991) and created a change process for Tier 2
information that has the same elements as the Tier 1 change process. In
addition, the Tier 2 change process includes a provision that is
similar to 10 CFR 50.59, namely VIII.B.5 of this appendix. Therefore,
as stated in section II (Topic 6) of the proposed rule, there is no
need for 10 CFR 52.63(b)(2) in the two-tiered change process that has
been implemented for this appendix.
Subsequently, in its comments dated July 23, 1996, NEI requested
the Commission to modify Section VIII.B.4 of this appendix so that
exemption requests are only subject to an opportunity for a hearing.
The Commission decided that NEI's proposal was consistent with the
intent of this appendix and modified Section VIII.B.4, accordingly.
Also, NEI requested the Commission to modify Section VIII.B.6.b of this
appendix to restrict the need for a license amendment and an
opportunity for a hearing to those Tier 2* changes involving unreviewed
safety questions. NEI claimed that a hearing opportunity for Tier 2*
changes was unnecessary and should be provided only if the change
involves an unreviewed safety question. The Commission disagrees with
NEI because of the safety significance of the Tier 2* information. The
safety significance of the Tier 2* information was determined at the
time that the Tier 1 information was selected.
[[Page 25809]]
Any changes to Tier 2* information will require a license amendment
with the appropriate hearing opportunity.
3. Need for Additional Applicable Regulations
Comment Summary. NEI and the other industry commenters criticized
Section 5(c) of the proposed design certification rule, which
designated additional applicable regulations for the purposes of 10 CFR
52.48, 52.54, 52.59, and 52.63 (refer to NEI Comments dated August 4,
1995, Attachment B, pp. 24-57; NEI Comments dated July 23, 1996, pp.
27-34; and NEI letter dated September 16, 1996).
Response. NEI raised many issues in its comments. These comments
have been consolidated into the following groups to facilitate
documentation of the NRC staff's responses.
NEI stated that there is no requirement in 10 CFR Part 52 that
compels the Commission to adopt these new applicable regulations, that
the new applicable regulations are not necessary for adequate
protection or to improve the safety of the standard designs, and that
the applicable regulations are inconsistent with the Commission's SRM,
dated September 14, 1993. NEI also stated that the adoption of new
applicable regulations is contrary to the purpose of design
certification and Commission policy. The NRC staff developed the new
applicable regulations in accordance with the goals of 10 CFR Part 52,
Commission guidance, and to achieve the purposes of 10 CFR 52.48,
52.54, 52.59, and 52.63 (refer to SECY-96-028, dated February 6, 1996,
and the History of Applicable Regulations in Attachment 9 to SECY-96-
077, dated April 15, 1996). The Commission chose design-specific
rulemaking rather than generic rulemaking for the new technical and
severe accident issues. The Commission adopted this approach early in
the design certification review process because it was concerned that
generic rulemakings would cause significant delay in the design
certification reviews and it was thought that the new requirements
would be design-specific (refer to SRMs on SECY-91-262 and SECY-93-
226). Furthermore, the SOC discussion for Part 52, Section II.1.e,
``Applicability of Existing Standards,'' states that new standards may
be required and that these new standards may be developed in a design-
specific rulemaking.
NEI stated that the applicable regulations are unnecessary because
the NRC staff has applied these technical positions in reviewing and
approving the standard designs. In addition, each of these positions
has corresponding NRC staff approved provisions in the respective
design control documents (DCD) and these provisions already serve the
purpose of applicable regulations for all of the situations identified
by the NRC staff. In response, the NRC staff stated that NEI's
statement that information in the DCD will constitute an applicable
regulation confuses the difference between design descriptions approved
by rulemaking and the regulations (safety standards) that are used as
the basis to approve the design. Furthermore, during a meeting on April
25, 1994, and in a letter from Mr. Dennis Crutchfield (NRC) to Mr.
William Rasin (NEI), dated July 25, 1994, the NRC staff stated that
design information cannot function as a surrogate for the new (design-
specific) applicable regulations because this information describes
only one method for meeting the regulation and would not provide a
basis for evaluating proposed changes to the previously approved design
descriptions.
NEI was also concerned that ``broadly stated'' applicable
regulations could be used in the future by the NRC staff to impose
backfits on applicants and licensees that could not otherwise be
justified on the basis of adequate protection of public health and
safety, thereby eroding licensing stability. However, NEI acknowledged
in its comments that the NRC staff did not intend to reinterpret the
applicable regulations to impose compliance backfits and because
implementation of the applicable regulations was approved in the DCD,
the NRC staff could not impose a backfit on the approved implementation
without meeting the standards in the change process. Also, NEI claimed
that the additional applicable regulations were vague and, in some
cases, inconsistent with previous Commission directions. In response to
NEI's comments, the NRC staff proposed revised wording and a special
provision for compliance backfits to the additional applicable
regulations (refer to SECY-96-077). However, in subsequent comments,
NEI stated that the proposed wording changes and backfit provision did
not mitigate its concerns.
NEI commented in 1995 that some of the additional applicable
regulations are requirements on an applicant or licensee who references
this appendix, and requested in 1996 that these requirements be deleted
from the final rule. The NRC staff moved these requirements from
Section 5 of the proposed rules to Section 4 of the rules set forth in
SECY-96-077, in response to NEI's 1995 comment (refer to pp. 46-47 of
Attachment 1 to SECY-96-077). The Commission has removed those
requirements from Section IV and has reserved the right to impose these
operational requirements on applicants and licensees who reference this
appendix (refer to VI.C of this appendix). The additional applicable
regulations that are applicable to applicants or licensees who
reference this appendix are specified in the generic DCD as COL license
information.
NEI stated that the proposed additional applicable regulations were
viewed as penalizing advanced plants for incorporating design features
that enhance safety and could impact the regulatory threshold for
currently operating plants. NEI also stated that applicable regulations
are not needed to permit the NRC to deny an exemption request for a
design feature that is subject to an applicable regulation. The
Commission decided not to codify the additional applicable regulations
that were identified in section 5(c) of the proposed rule. Instead, the
Commission adopted the following position relative to the proposed
additional applicable regulations.
Although it is the Commission's intent in 10 CFR Part 52 to promote
standardization and design stability of power reactor designs,
standardization and design stability are not exclusive goals. The
Commission recognized that there may be special circumstances when it
would be appropriate for applicants or licensees to depart from the
referenced certified designs. However, there is a desire of the
Commission to maintain standardization across a group of reactors of a
given design. Nevertheless, Part 52 provides for changes to a certified
design in carefully defined circumstances, and one of these
circumstances is the option provided to applicants and licensees
referencing certified designs to request an exemption from one or more
elements of the certified design, e.g., 10 CFR 52.63(b)(1). The final
design certification rule references this provision for Tier 1 and
includes a similar provision for Tier 2. The criteria for NRC review of
requests for an exemption from Tier 1 and Tier 2 in the final rule are
the same as those for NRC review of rule exemption requests under 10
CFR Part 50 directed at non-certified designs, except that the final
rule requires consideration of an additional factor for Tier 1
exemptions--whether special circumstances outweigh any decrease in
safety that may result from the reduction in standardization caused by
the exemption. It has been the
[[Page 25810]]
practice of the Commission to require that there be no significant
decrease in the level of safety provided by the regulations when
exemptions from the regulations in Part 50 are requested. The
Commission believes that a similar practice should be followed when
exemptions from one or more elements of a certified design are
requested, that is, the granting of an exemption under 10 CFR 50.12 or
52.63(b)(1) should not result in any significant decrease in the level
of safety provided by the design (Tier 1 and Tier 2). The exemption
standards in sections VIII.A.4 and VIII.B.4 of the final rule have been
modified from the proposed rule to codify this practice.
In adopting this policy the Commission recognizes that the ABWR
design not only meets the Commission's safety goals for internal
events, but also offers a substantial overall enhancement in safety as
compared, generally, with the current generation of operating power
reactors. See, e.g. NUREG-1503 at Section 19.1. The Commission
recognizes that the safety enhancement is the result of many elements
of the design, and that much but not all of it is reflected in the
results of the probabilistic risk assessment (PRA) performed and
documented for them. In adopting a rule that the safety enhancement
should not be eroded significantly by exemption requests, the
Commission recognizes and expects that this will require both careful
analysis and sound judgment, especially considering uncertainties in
the PRA and the lack of a precise, quantified definition of the
enhancement which would be used as the standard. Also, in some cases
scientific proof that a safety margin has or has not been eroded may be
difficult or even impossible. For this reason, it is appropriate to
express the Commission's policy preference regarding the grant of
exemptions in the form of a qualitative, risk informed standard, in
section VIII of the final rule, and inappropriate to express the policy
in a quantitative legal standard as part of the additional applicable
regulations.
There are three other circumstances where the enhanced safety
associated with the ABWR design could be eroded: by design changes
introduced by GE at the certification renewal stage; by operational
experience or other new information suggesting that safety margins
believed to be achieved are not in fact present; and by applicant or
licensee design changes under section VIII.B.5 of the final rule (for
changes to Tier 2 only). In the first two cases Part 52 limits NRC's
ability to require that the safety enhancement be restored, unless a
question of adequate protection or compliance would be presented or, in
the case of renewals, unless the restoration offers cost-justified,
substantive additional protection. Thus, unlike the case of exemptions
where a policy of maintaining enhanced safety can be enforced
consistent with the basic structure of Part 52, in the case of renewals
and new information, implementation of such a policy over industry
objections would require changes to the basic structure of Part 52. The
Commission has been and still is unwilling to make fundamental changes
to Part 52 because this would introduce great uncertainty and defeat
industry's reasonable expectation of a stable regulatory framework.
Nevertheless, the Commission on its part also has a reasonable
expectation that vendors and utilities will cooperate with the
Commission in assuring that the level of enhanced safety believed to be
achieved with this design will be reasonably maintained for the period
of the certification (including renewal).
This expectation that industry will cooperate with NRC in
maintaining the safety level of the certified designs applies to design
changes suggested by new information, to renewals, and to changes under
section VIII.B.5 of the final rule. If this reasonable expectation is
not realized, the Commission would carefully review the underlying
reasons and, if the circumstances were sufficiently persuasive,
consider the need to reexamine the backfitting and renewal standards in
Part 52 and the criteria for Tier 2 changes under section VIII.B.5. At
this time there is no reason to believe that cooperation will not be
forthcoming and, therefore, no reason to change the regulations. With
this belief and stated Commission policy (and the exemption standard
discussed above), there is no need for the proposed additional
applicable regulations to be embedded in the final rule because the
objective of the additional applicable regulations--maintaining the
enhanced level of safety--should be achieved without them.
B. Responses to Specific Requests for Comment
Only two commenters addressed the specific requests for comments
that were set forth in section IV of the SOC for the proposed rule.
These commenters were NEI and the Ohio Citizens for Responsible Energy,
Inc. (OCRE). The following discussion provides a summary of the
comments and the Commission's response.
1. Should the requirements of 10 CFR 52.63(c) be added to a new 10
CFR 52.79(e)?
Comment Summary. OCRE agreed that the requirements of 10 CFR
52.63(c) should be added to a new 10 CFR 52.79(e) and NEI had no
objection, as long as the substantive requirements in Sec. 52.63(c)
were not changed.
Response. Because there is no objection to adding the requirements
of 10 CFR 52.63(c) to Subpart C of Part 52, as 10 CFR 52.79(e), the
Commission will consider this amendment as part of a future review of
Part 52. This future review will also consider lessons learned from
this rulemaking and will determine if 10 CFR 52.63(c) should be deleted
from Subpart B of Part 52.
2. Are there other words or phrases that should be defined in
Section 2 of the proposed rule?
Comment Summary. Neither NEI nor OCRE suggested other words or
phrases that need to be added to the definition section. However, NEI
recommended expanded definitions for specific terms in Section 2 of the
proposed rule.
Response. The Commission has revised Section II of this appendix as
a result of comments from NEI and DOE. A discussion of these changes is
provided in sections II.C.2 and II.C.3 of this SOC.
3. What change process should apply to design-related information
developed by a combined license (COL) applicant or holder that
references this design certification rule?
Comment Summary. OCRE recommended the change process in Section
8(b)(5)(i) of the proposed rule and stated that it is essential that
any design-related COL information including the plant-specific PRA
(and changes thereto) developed by the COL applicant or holder not have
issue preclusion and be subject to litigation in any COL hearing. NEI
recommended that the COL information be controlled by 10 CFR 50.54 and
50.59 but recognized that the COL applicant or holder must also
consider impacts on Tier 1 and Tier 2 information. Subsequently, in its
comments dated July 23, 1996, NEI requested the Commission to modify
the response to this question that was set forth in SECY-96-077.
Specifically, NEI stated that plant-specific changes should be
implemented under Sec. 50.59 or Sec. 50.90, as appropriate. The
Commission did not significantly modify its former response because the
change process must consider the effect on information in the DCD, as
NEI previously acknowledged.
Response. The Commission will develop a change process for the
plant-specific information submitted in a COL application that
references this appendix as part of a future review of Part 52. The
Commission expects that
[[Page 25811]]
the change process for the plant-specific portion of the COL
application will be similar to VIII.B.5 of this appendix. This approach
is generally consistent with the recommendations of OCRE and NEI.
The Commission agrees with OCRE that the plant-specific portion of
the COL application will not have issue preclusion in the licensing
hearing. A discussion of the information that will have issue
preclusion is provided in sections II.A.1 and III.F of this SOC.
4. Are each of the applicable regulations set forth in Section 5(c)
of the proposed rule justified?
Comment Summary. OCRE found each of the applicable regulations to
be justified and stated that these requirements are responsive to
issues arising from operating experience and will greatly reduce the
risk of severe accidents for plants using these standard designs. NEI
believes that none of the applicable regulations are justified and
stated that they are legally and technically unnecessary, could give
rise to unwarranted backfits, are destabilizing and, therefore,
contrary to the purpose of 10 CFR Part 52.
Response. The Commission has determined that it is not necessary to
codify the new applicable regulations, as explained in section II.A.3
of this SOC.
5. Section 8(b)(5)(i) of the proposed rule authorizes an applicant
or licensee who references the design certification to depart from Tier
2 information without prior NRC approval if the applicant or licensee
makes a determination that the change does not involve a change to Tier
1 or Tier 2 * information, as identified in the DCD; the technical
specifications; or an unreviewed safety question, as defined in
Sections 8(b)(5)(ii) and (iii). Where Section 8(b)(5)(i) states that a
change made pursuant to that paragraph will no longer be considered as
a matter resolved in connection with the issuance or renewal of a
design certification within the meaning of 10 CFR 52.63(a)(4), should
this mean that the determination may be challenged as not demonstrating
that the change may be made without prior NRC approval or that the
change itself may be challenged as not complying with the Commission's
requirements?
Comment Summary. OCRE believes that the process for plant-specific
departures from Tier 2, as well as the substantive aspect of the change
itself, should be open to challenge, although OCRE believes that the
second aspect is the more important. By contrast, NEI argued that
neither the departure process nor the change should be subject to
litigation in any licensing hearing. Rather, NEI argued that any person
who wished to challenge the change should raise the matter in a
petition for an enforcement action under 10 CFR 2.206.
Response. The Commission has determined that an interested person
should be provided the opportunity to challenge, in an appropriate
licensing proceeding, whether the applicant or licensee properly
complied with the Tier 2 departure process. Therefore, VIII.B.5 of this
appendix has been modified to include a provision for challenging Tier
2 departures. The scope of finality for plant-specific departures is
discussed in greater detail in section II.A.1 of this SOC.
6. How should the determinations made by an applicant or licensee
that changes may be made under Section 8(b)(5)(i) of the proposed rule,
without prior NRC approval, be made available to the public in order
for those determinations to be challenged or for the changes themselves
to be challenged?
Comment Summary. OCRE recommends that the determinations and
descriptions of the changes be set forth in the COL application and
that they should be submitted to the NRC after COL issuance. Any person
wishing to challenge the determinations or changes should file a
petition pursuant to 10 CFR 2.206. NEI recommends submitting periodic
reports that summarize departures made under Section 8(b)(5) to the NRC
pursuant to Section 9(b) of the proposed design certification rules,
consistent with the existing process for NRC notifications by licensees
under 10 CFR 50.59. These reports will be available in the NRC's Public
Document Room.
Response. The Tier 2 departure process in Section 8(b)(5) and the
respective reporting requirements in Section 9(b) of the proposed
design certification rule (VIII.B.5 and X.B of this appendix) were
based on 10 CFR 50.59. It therefore seems reasonable that the
information collection and reporting requirements that should be used
to control Tier 2 departures made in accordance with VIII.B.5 of this
appendix should generally follow the regulatory scheme in 10 CFR 50.59
(except that the requirements should also be applied to COL
applicants), absent countervailing considerations unique to the design
certification and combined license regulatory scheme in Part 52. OCRE's
proposal raises policy considerations which are not unique to this
design certification, but are equally applicable to the Part 50
licensing scheme. In fact, OCRE has submitted a petition (see 59 FR
30308; June 13, 1994) which raises the generic matter of public access
to licensee-held information. In view of the generic nature of OCRE's
concern and the pendency of OCRE's petition, which independently raises
this matter, the Commission concludes that this rulemaking should not
address this matter.
7. What is the preferred regulatory process (including
opportunities for public participation) for NRC review of proposed
changes to Tier 2 * information and the commenter's basis for
recommending a particular process?
Comment Summary. OCRE recommends either an amendment to the license
application or an amendment to the license, with the requisite hearing
rights. NEI recommends NRC approval by letter with an opportunity for
public hearing only for those Tier 2 * changes that also involve either
a change in Tier 1 or technical specifications, or an unreviewed safety
question.
Response. The Commission has developed a change process for Tier 2
* information, as described in sections II.A.2 and III.H of this SOC,
which essentially treats the proposed departure as a request for a
license amendment with an opportunity for hearing. Since Tier 2 *
departures require NRC review and approval, and involve a licensee
departing from the requirements of this appendix, the Commission
regards such requests for departures as analogous to license
amendments. Accordingly, VIII.B.6 of this appendix specifies that such
requests will be treated as requests for license amendments after the
license is issued, and that the Tier 2 * departure shall not be
considered to be matters resolved by this rulemaking prior to a license
being issued.
8. Should determinations of whether proposed changes to severe
accident issues constitute an unreviewed safety question use different
criteria than for other safety issues resolved in the design
certification review and, if so, what should those criteria be?
Comment Summary. OCRE supports the concept behind the criteria in
the proposed rule for determining if a proposed change to severe
accident issues constitutes an unreviewed safety question, but proposes
changes to the criteria. NEI agrees with the criteria in the proposed
rule but recommends an expansion of the scope of information that would
come under the special criteria for determining an unreviewed safety
question.
Response. The Commission disagrees with the recommendations of both
NEI and OCRE. The Commission has decided to retain the special change
[[Page 25812]]
process for severe accident information, as described in sections
II.A.2 and III.H of this SOC.
9. (a) (1) Should construction permit applicants under 10 CFR Part
50 be allowed to reference design certification rules to satisfy the
relevant requirements of 10 CFR Part 50?
(2) What, if any, issue preclusion exists in a subsequent operating
license stage and NRC enforcement, after the Commission authorizes a
construction permit applicant to reference a design certification rule?
(3) Should construction permit applicants referencing a design
certification rule be either permitted or required to reference the
ITAAC? If so, what are the legal consequences, in terms of the scope of
NRC review and approval and the scope of admissible contentions, at the
subsequent operating license proceeding?
(4) What would distinguish the ``old'' 10 CFR Part 50 2-step
process from the 10 CFR Part 52 combined license process if a
construction permit applicant is permitted to reference a design
certification rule and the final design and ITAAC are given full issue
preclusion in the operating license proceeding? To the extent this
circumstance approximates a combined license, without being one, is it
inconsistent with Section 189(b) of the Atomic Energy Act (added by the
Energy Policy Act of 1992) providing specifically for combined
licenses?
(b)(1) Should operating license applicants under 10 CFR Part 50 be
allowed to reference design certification rules to satisfy the relevant
requirements of 10 CFR Part 50?
(2) What should be the legal consequences, from the standpoints of
issue resolution in the operating license proceeding, NRC enforcement,
and licensee operation if a design certification rule is referenced by
an applicant for an operating license under 10 CFR Part 50?
(c) Is it necessary to resolve these issues as part of this design
certification, or may resolution of these issues be deferred without
adverse consequence (e.g., without foreclosing alternatives for future
resolution).
Comment Summary. OCRE proposed that a construction permit applicant
should be allowed to reference design certifications and that the
applicant be required to reference ITAAC because they are Tier 1. OCRE
indicated that in a construction permit hearing, those issues
representing a challenge to the design certification rule would be
prohibited pursuant to 10 CFR 2.758. At the operating license stage,
only an applicant whose construction permit referenced a design
certification rule should be allowed to reference the design
certification. In the operating license hearing, issues would be
limited to whether the ITAAC have been met. Requiring a construction
permit applicant to reference the ITAAC would not be the same as a
combined license applicant under 10 CFR Part 52, in OCRE's view,
apparently because the specific hearing provisions of 10 CFR 52.103
would not be employed. Finally, OCRE argued that resolution of these
issues could be safely deferred because the circumstances with which
these issues attend are not likely to be faced.
NEI also argued that a construction permit applicant should be
allowed to reference design certifications. However, NEI believed that
the applicant should be permitted, but not required, to reference the
ITAAC. If the applicant did not reference the ITAAC, then
``construction-related issues'' would be subject to both NRC review and
an opportunity for hearing at the operating license stage in the same
manner as construction-related issues in current Part 50 operating
license proceedings. NEI reiterated its view that design certification
issues should be considered resolved in all subsequent NRC proceedings.
With respect to deferring a Commission decision on the matter, NEI
suggested that these issues be resolved now because the industry wishes
to ``reinforce'' the permissibility of using a design certification in
a Part 50 proceeding. Further, NEI argues that deletion of all mention
of construction permits and operating licenses in the design
certification rule could be construed as indicating the Commission's
desire to preclude a construction permit or operating license applicant
from referencing a design certification.
Response. Although 10 CFR Part 52 provides for referencing of
design certification rules in Part 50 applications and licenses, the
Commission wishes to reserve for future consideration the manner in
which a Part 50 applicant could be permitted to reference this design
certification and whether it should be permitted or required to
reference the ITAAC. This decision is due to the manner in which ITAAC
were developed for this appendix and recognition of the lack of
experience with design certifications in combined licenses, in
particular the implementation of ITAAC. Therefore, the Commission has
decided that it is appropriate for the final rule to have some
uncertainty regarding the manner in which this appendix could be
referenced in a Part 50 proceeding, as set forth in Section IV.B of
this appendix.
C. Other Issues
1. NRC Verification of ITAAC Determinations
Comment Summary. In Attachment B of its comments dated August 4,
1995 (pp. 58-66), NEI raised an industry concern regarding the matters
to be considered by the NRC in verifying inspections, tests, analyses,
and acceptance criteria (ITAAC) determinations pursuant to 10 CFR
52.99, specifically citing quality assurance and quality control (QA/
QC) deficiencies. Although this issue was not specifically addressed in
the proposed rule, the following response is provided because of its
importance relative to future considerations of the successful
performance of ITAAC for a nuclear power facility. Subsequently, in its
comments dated July 23, 1996, NEI requested the Commission to delete
significant portions of the NRC's response, which was originally set
forth in SECY-96-077 (refer to pages 33-36 of Attachment 1).
Response. The Commission decided to delete the responses in SECY-
96-077 on licensee documentation of ITAAC verification; NRC inspection;
and facility ITAAC verification; because they do not directly relate to
the design certification rulemakings. However, the NRC disagrees with
NEI's assertion that QA/QC deficiencies have no relevance to the NRC
determination of whether ITAAC have been successfully completed. Simply
confirming that an ITAAC had been performed in some manner and a result
obtained apparently showing that the acceptance criteria had been met
would not be sufficient to support a determination that the ITAAC had
been successfully completed. The manner in which an ITAAC is performed
can be relevant and material to the results of the ITAAC. For example,
in conducting an ITAAC to verify a pump's flow rate, it is logical,
even if not explicitly specified in the ITAAC, that the gauge used to
verify the pump flow rate must be calibrated in accordance with
relevant QA/QC requirements and that the test configuration is
representative of the final as-built plant conditions (i.e. valve or
system line-ups, gauge locations, system pressures or temperatures).
Otherwise, the acceptance criteria for pump flow rate in the ITAAC
could apparently be met while the actual flow rate in the system could
be much less than that required by the approved design.
[[Page 25813]]
The NRC has determined that a QA/QC deficiency may be considered in
determining whether an ITAAC has been successfully completed if: (1)
The QA/QC deficiency is directly and materially related to one or more
aspects of the relevant ITAAC (or supporting Tier 2 information); and
(2) the deficiency (considered by itself, with other deficiencies, or
with other information known to the NRC) leads the NRC to question
whether there is a reasonable basis for concluding that the relevant
aspect of the ITAAC has been successfully completed. This approach is
consistent with the NRC's current methods for verifying initial test
programs. The NRC recognizes that there may be programmatic QA/QC
deficiencies that are not relevant to one or more aspects of a given
ITAAC under review and, therefore, should not be relevant to or
considered in the NRC's determination as to whether an ITAAC has been
successfully completed. Similarly, individual QA/QC deficiencies
unrelated to an aspect of the ITAAC in question would not form the
basis for an NRC determination that an ITAAC has not been met. Using
the ITAAC for pump flow rate example, a specific QA deficiency in the
calibration of pump gauges would not preclude an NRC determination of
successful ITAAC completion if the licensee could demonstrate that the
original deficiency was properly corrected (e.g., analysis, scope of
effect, root cause determination, and corrective actions as
appropriate), or that the deficiency could not have materially affected
the test in question.
Furthermore, although Tier 1 information was developed to focus on
the performance of the structures, systems, and components of the
design, the information contains implicit quality standards. For
example, the design descriptions for reactor and fluid systems describe
which systems are ``safety-related;'' important piping systems are
classified as ``Seismic Category I'' and identify the ASME Code Class;
and important electrical and instrumentation and control systems are
classified as ``Class 1E.'' The use of these terms by the evolutionary
plant designers was meant to ensure that the systems would be built and
maintained to the appropriate standards. Quality assurance deficiencies
for these systems would be assessed for their impact on the performance
of the ITAAC, based on their safety significance to the system. The QA
requirements of 10 CFR Part 50, Appendix B, apply to safety-related
activities. Therefore, the Commission anticipates that, because of the
special significance of ITAAC related to verification of the facility,
the licensee will implement similar QA processes for ITAAC activities
that are not safety-related.
During the ITAAC development, the design certification applicants
determined that it was impossible (or extremely burdensome) to provide
all details relevant to verifying all aspects of ITAAC (e.g., QA/QC) in
Tier 1 or Tier 2. Therefore, the NRC staff accepted the applicants'
proposal that top-level design information be stated in the ITAAC to
ensure that it was verified, with an emphasis on verification of the
design and construction details in the ``as-built'' facility. To argue
that consideration of underlying information which is relevant and
material to determining whether ITAAC have been successfully completed,
ignores the history of ITAAC development. In summary, the Commission
concludes that information such as QA/QC deficiencies which are
relevant and material to ITAAC may be considered by the NRC in
determining whether the ITAAC have been successfully completed. Despite
this conclusion, the Commission has decided to add a provision to this
appendix (IX.B.1), which was requested by NEI. This provision requires
the NRC's findings (that the prescribed acceptance criteria have been
met) to be based solely on the inspections, tests, and analyses. The
Commission has added this provision, which is fully consistent with 10
CFR Part 52, with the understanding that it does not affect the manner
in which the NRC intends to implement 10 CFR 52.99 and 52.103(g), as
described above.
2. DCD Introduction
Comment Summary. The proposed rule incorporated Tier 1 and Tier 2
information into the DCD but did not include the introduction to the
DCD. The SOC for the proposed rule indicated that this was a deliberate
decision, stating:
The introduction to the DCD is neither Tier 1 nor Tier 2
information, and is not part of the information in the DCD that is
incorporated by reference into this design certification rule.
Rather, the DCD introduction constitutes an explanation of
requirements and other provisions of this design certification rule.
If there is a conflict between the explanations in the DCD
introduction and the explanations of this design certification rule
in these statements of consideration (SOC), then this SOC is
controlling.
Both the applicant and NEI took strong exception to this statement.
They both argued that the language of the DCD introduction was the
subject of careful discussion and negotiation between the NRC staff,
NRC's Office of the General Counsel, and representatives of the
applicant and NEI. They, therefore, suggested that the definition of
the DCD in Section 2(a) of the proposed rule be amended to explicitly
include the DCD Introduction and that Section 4(a) of the proposed rule
be amended to generally require that applicants or licensees comply
with the entire DCD. However, in the event that the Commission rejected
their suggestion, NEI alternatively argued that the substantive
provisions of the DCD Introduction be directly incorporated into the
design certification rule's language (refer to NEI Comments dated
August 4, 1995, Attachment B, pp. 90-108, and July 23, 1996, pp. 43-49;
GE Comments, Attachment A, pp. 10-11).
Response. The DCD Introduction was created to be a convenient
explanation of some provisions of the design certification rule and was
not intended to become rule language itself. Therefore, the Commission
declines the suggestion to incorporate the DCD introduction, but
adopted NEI's alternative suggestion of incorporating substantive
procedural and administrative requirements into the design
certification rule. It is the Commission's view that the procedural and
administrative provisions described in the DCD Introduction should be
included in, and be an integrated part of, the design certification
rule. As a result, Sections II, III, IV, VI, VIII, and X of this
appendix have been revised and Section IX was created to adopt
appropriate provisions from the DCD Introduction. In some cases, the
wording of these provisions has been modified, as appropriate, to
achieve clarity or to conform with the final design certification rule
language.
3. Duplicate Documentation in Design Certification Rule
Comment Summary. On page 4 of its comments, dated August 7, 1995,
the Department of Energy (DOE) recommended that the process for
preparing the design certification rule be simplified by eliminating
the DCD, which DOE claims is essentially a repetition of the Standard
Safety Analysis Report (SSAR). DOE's concern, which was further
clarified during a public meeting on December 4, 1995, is that the NRC
will require separate copies of the DCD and SSAR to be maintained.
During the public meeting, DOE also expressed a concern that
Sec. 52.79(b) could be confusing to an applicant for a combined license
because it currently states: ``The final safety analysis report and
other required
[[Page 25814]]
information may incorporate by reference the final safety analysis
report for a certified standard design.''
Response. The NRC does not require duplicate documentation for this
design certification rule. The DCD is the only document that is
incorporated by reference into this appendix in order to meet the
requirements of Subpart B of Part 52. The SSAR supports the final
design approval (FDA) that was issued under Appendix O to 10 CFR Part
52. The DCD was developed to meet the requirements for incorporation by
reference and to conform with requests from the industry such as
deletion of the quantitative portions of the design-specific
probabilistic risk assessment. Because the DCD terminology was not
envisioned at the time that Part 52 was developed, the Commission will
consider modifying Sec. 52.79(b), as part of its future review of Part
52, in order to clarify the use of the term ``final safety analysis
report.'' In the records and reporting requirements in Section X of
this appendix, additional terms were used to distinguish between the
documents to be maintained by the applicant for this design
certification rule and the document to be maintained by an applicant or
licensee who references this appendix. These new terms are defined in
Section II of this appendix and further described in the section-by-
section discussion on records and reporting in section III.J of this
SOC. The applicant chose to continue to reference the SSAR as the
supporting document for its FDA. As a result, the applicant must
maintain the SSAR for the duration of the FDA.
4. In its Comments, Dated August 12, 1995, OCRE Stated
Although the ABWR will use the same type of Main Steam Isolation
Valves as are used in operating BWRs, it will not have a MSIV
Leakage Control System. Instead, GE is taking credit for fission
product retention in the main steam lines and main condenser.
However, in a main steam line break outside of containment, a design
basis event, such fission product retention will not occur. Given
the excessive leakage experience of MSIVs in operating BWRs, it
would be prudent to incorporate a MSIVLCS into the ABWR design. OCRE
would recommend a positive pressure MSIVLCS, which would pressurize
the main steam lines between the inboard and outboard MSIVs after
MSIV closure to a pressure above that in the reactor pressure
vessel. Thus, any leakage through the inboard MSIV will be into the
reactor.
Response. The NRC had concerns with the effectiveness of the main
steam isolation valve leakage collection system (MSIVLCS) to perform
its intended function under conditions of high MSIV leakage. NRC
classified this concern as a generic issue (C-8). An NRC study of
Generic Issue C-8 showed that neither the installation or removal of
the MSIVLCS could be justified. Operating experience with these systems
has shown that the MSIVLCS has required substantial maintenance and
resulted in substantial worker radiation exposure. The BWR Owners Group
subsequently proposed a resolution that would eliminate the safety-
related MSIVLCS and take cognizance of the fact that plate-out and
holdup of fission products leaking past the main steam isolation valves
will occur in the main steam lines and condenser. For the purpose of
giving credit to iodine holdup and plate-out in the main steam lines
and condensers, the NRC requires that the main steam piping (including
its associated piping to the condenser) and the condenser remain
structurally intact following a safe shutdown earthquake (Refer to NRC
Commission paper, SECY-93-087, ``Policy, Technical, and Licensing
Issues Pertaining to Evolutionary and Advanced Light-Water Reactor
(ALWR) Designs,'' dated April 2, 1993). The BWR Owners Group submitted
a topical report that proposed to eliminate the MSIVLCS and increase
the allowable MSIV leakage rates by taking credit for the holdup and
plate-out of fission products. The NRC has already approved plant
specific technical specification changes to eliminate the MSIVLCS for
the Hatch, Duane Arnold, and Limerick plants.
The U.S. ABWR design was evaluated against a number of design basis
accidents and was approved without a MSIVLCS. For the U.S. ABWR,
fission product holdup and plate-out in components of the main steam
system was justified and, therefore, was assumed in NRC's design basis
analyses. However, for the main steam line break, the NRC assumed that
one of the four main steam lines ruptured between the outer isolation
valve and turbine control valves, and did not take credit for retention
of iodine and noble gases in the coolant released through the break.
Any leakage through the MSIV after isolation was also assumed to be
released directly to the atmosphere. The contribution of this leakage
is insignificant when compared to the amount of reactor coolant lost
through the break prior to automatic isolation of the MSIV. In summary,
the U.S. ABWR represents an improved boiling water reactor design that
reduces worker radiation exposure, and meets the requirements of 10 CFR
Part 100 without the need for a MSIVLCS. Inclusion of an MSIVLCS would
result in substantial occupational exposures with little safety
benefit. Therefore, the Commission declines to adopt OCRE's
recommendation that a positive-pressure MSIVLCS be incorporated into
the U.S. ABWR design.
5. In its Comments, Dated August 12, 1995, OCRE Stated
The ABWR Standby Liquid Control System requires simultaneous
parallel, two-pump operation to achieve 100 gpm flow rate, necessary
to comply with 10 CFR 50.62(c)(4). However, a single failure
rendering one train inoperable would only yield a flow of 50 gpm,
which does not comply with the ATWS rule. OCRE recommends increasing
the capacity of each SLCS train to 100 gpm, so that the SLCS can
perform its ATWS mitigation function even with a single failure.
Response. The ATWS rule (10 CFR 50.62) requires the following with
regard to the SLCS for a boiling water reactor: ``Each boiling water
reactor must have a standby liquid control system (SLCS) with the
capability of injecting into the reactor pressure vessel a borated
water solution at such a flow rate, level of boron concentration and
boron-10 isotope enrichment, and accounting for reactor pressure vessel
volume, that the resulting reactivity control is at least equivalent to
that resulting from injection of 86 gallons per minute of 13 weight
percent sodium pentaborate decahydrate solution at the natural boron-10
isotope abundance into a 251-inch inside diameter reactor pressure
vessel for a given core design.'' For the U.S. ABWR design with a 278
inch inside diameter vessel, the ATWS rule is satisfied with injection
of 100 gpm of 13.4 weight percent of natural boron solution.
The Commission has previously concluded, as part of the ATWS
rulemaking, that a single-failure need not be assumed in the evaluation
of the SLCS. The statements of consideration for the ATWS rule 10 CFR
50.62 (49 FR 26036; June 26, 1984), under the heading ``Considerations
Regarding System and Equipment Criteria,'' states: ``In view of the
redundancy provided in existing reactor trip systems, the equipment
required by this amendment does not have to be redundant within
itself.'' OCRE presented no information which would lead the Commission
to reconsider and change its previous determination with respect to a
single-failure and the Commission declines to adopt OCRE's proposal.
[[Page 25815]]
6. In its Comments, Dated August 12, 1995, OCRE Stated
In the ABWR, the drywell to wetwell vacuum breakers consist of a
single vacuum breaker valve in each line. In operating BWRs, there
are two vacuum breaker valves in series in each line. The ABWR
design thus is vulnerable to a single failure, a stuck-open vacuum
breaker, which would result in suppression pool bypass, which can
overpressurize the containment in both design basis and severe
accidents. Having the containment function vulnerable to a single
failure is unacceptable. OCRE recommends the addition of a second
vacuum breaker valve in series with the one proposed in the design.
Response. The wetwell to drywell vacuum breaker system of operating
BWRs varies. Some operating BWRs have a single check valve per line
(typically Mark I's), others have two check valves in series (typically
Mark II's), and still others have a check valve in series with a motor
operated valve (typically Mark III's). The main concern with the number
of valves per vacuum breaker line focuses on the suppression pool
bypass capability of the containment design. In the evaluation of the
suppression pool bypass capability, a number of factors other than the
number of valves in each line must be considered to determine the
acceptability of the design. These factors are specified in the
Standard Review Plan Section 6.2.1.1.C, Appendix A (NUREG-0800) and
include the capability of containment sprays, periodic bypass leakage
testing and surveillance, and vacuum relief valve position indication.
A complete discussion of all these factors is included in the NRC's
NUREG-1503, Volume 1, ``Final Safety Evaluation Report Related to the
Certification of the Advanced Boiling Water Reactor Design,'' Sections
6.2.1.5, 6.2.1.8, 19.1.3.5.3, 19.2.3.3.5, and 20.5.1.
The U.S. ABWR wetwell to drywell vacuum breaker system consists of
eight lines, with a single check valve per line. For design basis
accidents, a single failure of the vacuum breaker in the stuck-open
position is not required to be considered for the U.S. ABWR. The U.S.
ABWR vacuum breakers are biased closed due to gravity and have
redundant position indication and alarm in the control room. Operating
plants have experienced stuck-open vacuum breakers as a result of
monthly stroke testing of the vacuum breakers. Most of these failures
have been related to the motor-operators installed for the purpose of
surveillance testing. The U.S. ABWR vacuum breakers do not have motor
operators and are subject to functional testing every 18 months.
Therefore, they are not subject to the motor operator failure mode and
due to the reduced frequency of surveillance testing and position
indication, these check valves are less likely to be stuck open when
needed during an accident.
A single failure of the vacuum breaker in the stuck-open position
is, however, considered in the evaluation of severe accident mitigation
capability. The analysis performed by GE indicates that the various
containment spray systems are capable of mitigating the consequences of
this scenario. In addition to the normal containment spray system, the
containment spray header can be supplied with water from the AC
independent water addition system (fire system) to mitigate bypass for
severe accidents.
GE performed an evaluation of many potential enhancements,
including adding a second vacuum breaker valve in series (Technical
Support Document for the ABWR). This evaluation concludes that the
potential safety enhancement of a second vacuum breaker valve in series
is minimal due to the existing design features. The NRC evaluated GE's
analysis of various design alternatives and concurs with GE's
conclusion. Although OCRE's suggested design change (the addition of a
second vacuum breaker valve in series) could minimally enhance safety,
the costs of such a change are not justified in view of the marginal
increase in safety (refer to section IV of this SOC). Accordingly, the
Commission declines to adopt OCRE's proposal.
7. In its comments, dated August 12, 1995, OCRE referred to
additional remarks made in a letter from the Advisory Committee on
Reactor Safeguards (ACRS), dated July 18, 1989, on proposed NRC staff
actions regarding the fire risk scoping study (NUREG/CR-5088). OCRE
believes that the recommendation, from two ACRS members, that the NRC
staff require the use of armored electrical cable in advanced light-
water reactors is sound advice. OCRE recommended that the NRC require
the use of armored cable in the U.S. ABWR and in all future nuclear
power plants.
Response. In reviewing the U.S. ABWR design, the NRC staff used the
enhanced guidance described in SECY-90-016, ``Evolutionary Light Water
Reactor (LWR) Certification Issues and Their Relationships to Current
Regulatory Requirements,'' dated January 12, 1990. The Commission
approved the NRC staff's position in SECY-90-016. This guidance was
used to resolve fire protection issues to minimize fire as a
significant contributor to the likelihood of a severe accident. The NRC
staff required that the U.S. ABWR design must be able to ensure that
safe shutdown can be achieved assuming that all equipment in any one
fire area will be rendered inoperable by fire and that reentry into the
fire area for repairs and operator actions is not possible. Because of
its physical configuration, the control room is excluded from this
approach and the U.S. ABWR is provided with an independent alternative
shutdown capability that is physically and electrically independent of
the control room. In the reactor containment building, the safety
divisions are widely separated around containment so that a single fire
will not cause the failure of any combination of active components that
could prevent safe shutdown. Additionally, the U.S. ABWR containment is
inerted with nitrogen during power operation which will prevent
propagation of any potential fire inside containment.
Evaluation of fire protection using this guidance assures an
acceptable level of safety for the U.S. ABWR. Instead of trying to
protect equipment in the fire area, the enhanced guidance requires that
equipment needed for safe shutdown be located in separate areas of the
plant so that one fire will not damage enough equipment to jeopardize
safe shutdown. While the use of armored electrical cable may provide
some protection to the electrical cables in the fire area, it does not
ensure that the cables will not be affected by the heat generated by
the fire. In addition, following a fire or other event that could
affect the cables, it would be impossible to inspect the cables to
determine if they were damaged by the event. Therefore, the NRC staff
does not agree that the ABWR should be required to use armored
electrical cables.
III. Section-by-Section Discussion
A. Introduction
The purpose of Section I of Appendix A to 10 CFR Part 52 (``this
appendix'') is to identify the standard plant design that is approved
by this design certification rule and the applicant for certification
of the standard design. Identification of the design certification
applicant is necessary to implement this appendix, for two reasons.
First, the implementation of 10 CFR 52.63(c) depends on whether an
applicant for a combined license (COL) contracts with the design
certification applicant to provide the generic DCD and supporting
design information. If the COL applicant does not use the design
certification applicant to provide this information, then the COL
applicant must meet the
[[Page 25816]]
requirements in 10 CFR 52.63(c). Also, X.A.1 of this appendix imposes a
requirement on the design certification applicant to maintain the
generic DCD throughout the time period in which this appendix may be
referenced.
B. Definitions
The terms Tier 1, Tier 2, Tier 2*, and COL action items (license
information) are defined in this appendix because these concepts were
not envisioned when 10 CFR Part 52 was developed. The design
certification applicants and the NRC staff used these terms in
implementing the two-tiered rule structure that was proposed by
industry after the issuance of 10 CFR Part 52. In addition, during
consideration of the comments received on the proposed rule, the
Commission determined that it would be useful to distinguish between
the ``plant-specific DCD'' and the ``generic DCD,'' the latter of which
is incorporated by reference into this appendix and remains unaffected
by plant-specific departures. This distinction is necessary in order to
clarify the obligations of applicants and licensees that reference this
appendix. Also, the technical specifications that are located in
Chapter 16 of the generic DCD were designated as ``generic technical
specifications'' to facilitate the special treatment of this
information in the final rule (refer to section II.A.1 of this SOC).
Therefore, appropriate definitions for these additional terms are
included in the final rule.
The Tier 1 portion of the design-related information contained in
the DCD is certified by this appendix and, therefore, subject to the
special backfit provisions in VIII.A of this appendix. An applicant who
references this appendix is required to incorporate by reference and
comply with Tier 1, under III.B and IV.A.1 of this appendix. This
information consists of an introduction to Tier 1, the design
descriptions and corresponding ITAAC for systems and structures of the
design, design material applicable to multiple systems of the design,
significant interface requirements, and significant site parameters for
the design. The design descriptions, interface requirements, and site
parameters in Tier 1 were derived entirely from Tier 2, but may be more
general than the Tier 2 information. The NRC staff's evaluation of the
Tier 1 information, including a description of how this information was
developed is provided in Section 14.3 of the FSER. Changes to or
departures from the Tier 1 information must comply with VIII.A of this
appendix.
The Tier 1 design descriptions serve as design commitments for the
lifetime of a facility referencing the design certification. The ITAAC
verify that the as-built facility conforms with the approved design and
applicable regulations. In accordance with 10 CFR 52.103(g), the
Commission must find that the acceptance criteria in the ITAAC are met
before operation. After the Commission has made the finding required by
10 CFR 52.103(g), the ITAAC do not constitute regulatory requirements
for licensees or for renewal of the COL. However, subsequent
modifications to the facility must comply with the design descriptions
in the plant-specific DCD unless changes are made in accordance with
the change process in Section VIII of this appendix. The Tier 1
interface requirements are the most significant of the interface
requirements for systems that are wholly or partially outside the scope
of the standard design, which were submitted in response to 10 CFR
52.47(a)(1)(vii) and must be met by the site-specific design features
of a facility that references the design certification. The Tier 1 site
parameters are the most significant site parameters, which were
submitted in response to 10 CFR 52.47(a)(1)(iii). An application that
references this appendix must demonstrate that the site parameters
(both Tier 1 and Tier 2) are met at the proposed site (refer to
discussion in III.D of this SOC).
Tier 2 is the portion of the design-related information contained
in the DCD that is approved by this appendix but is not certified. Tier
2 information is subject to the backfit provisions in VIII.B of this
appendix. Tier 2 includes the information required by 10 CFR 52.47,
with the exception of generic technical specifications and conceptual
design information, and supporting information on the inspections,
tests, and analyses that will be performed to demonstrate that the
acceptance criteria in the ITAAC have been met. As with Tier 1, III.B
and IV.A.1 of this appendix require an applicant who references this
appendix to incorporate Tier 2 by reference and to comply with Tier 2
(except for the COL action items and conceptual design information).
The definition of Tier 2 makes clear that Tier 2 information has been
determined by the Commission, by virtue of its inclusion in this
appendix and its designation as Tier 2 information, to be an approved
(``sufficient'') method for meeting Tier 1 requirements. However, there
may be other acceptable ways of complying with Tier 1. The appropriate
criteria for departing from Tier 2 information are set forth in Section
VIII of this appendix. Departures from Tier 2 do not negate the
requirement in Section III.B to reference Tier 2. NEI requested the
Commission, in its comments dated July 23, 1996, to include several
statements on compliance with Tier 2 in the definitions of Tier 1 and
Tier 2. The Commission determined that inclusion of those statements in
the Tier 2 definition was appropriate, but to also include them in the
Tier 1 definition would be unnecessarily redundant.
Certain Tier 2 information has been designated in the generic DCD
with brackets and italicized text as ``Tier 2*'' information and, as
discussed in greater detail in the section-by-section explanation for
Section VIII, a plant-specific departure from Tier 2* information
requires prior NRC approval. However, the Tier 2* designation expires
for some of this information when the facility first achieves full
power after the finding required by 10 CFR 52.103(g). The process for
changing Tier 2* information and the time at which its status as Tier
2* expires is set forth in VIII.B.6 of this appendix.
A definition of ``combined license (COL) action items'' (COL
license information) has been added to clarify that COL applicants are
required to address these matters in their license application, but the
COL action items are not the only acceptable set of information. 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 its
FSAR.
In developing the proposed design certification rule, the
Commission contemplated that there would be both generic (master) DCDs
maintained by the NRC and the design certification applicant, as well
as individual plant-specific DCDs, maintained by each applicant and
licensee who references this design certification rule. The generic
DCDs (identical to each other) would reflect generic changes to the
version of the DCD approved in this design certification rulemaking.
The generic changes would occur as the result of generic rulemaking by
the Commission (subject to the change criteria in Section VIII of this
appendix). In addition, the Commission understood that each applicant
and licensee referencing this Appendix would be required to submit and
maintain a plant-specific DCD. This plant-specific DCD would contain
(not just incorporate by reference) the information in the generic DCD.
The plant-specific DCD would be
[[Page 25817]]
updated as necessary to reflect the generic changes to the DCD that the
Commission may adopt through rulemaking, any plant-specific departures
from the generic DCD that the Commission imposed on the licensee by
order, and any plant-specific departures that the licensee chose to
make in accordance with the relevant processes in Section VIII of this
appendix. Thus, the plant-specific DCD would function akin to an
updated Final Safety Analysis Report, in the since that it would
provide the most complete and accurate information on a plant's
licensing basis for that part of the plant within the scope of this
appendix. However, the proposed rule defined only the concept of the
``master'' DCD. The Commission continues to believe that there should
be both a generic DCD and plant-specific DCDs. To clarify this matter,
the proposed rule's definition of DCD has been redesignated as the
``generic DCD,'' a new definition of ``plant-specific DCD'' has been
added, and conforming changes have been made to the remainder of the
rule. Further information on exemptions or departures from information
in the DCD is provided in section III.H below. The Final Safety
Analysis Report (FSAR) that is required by Sec. 52.79(b) will consist
of the plant-specific DCD, the site-specific portion of the FSAR, and
the plant-specific technical specifications.
During the resolution of comments on the final rules in SECY-96-
077, the Commission decided to treat the technical specifications in
Chapter 16 of the DCD as a special category of information and to
designate them as generic technical specifications (refer to II.A.1 of
SOC). A COL applicant must submit plant-specific technical
specifications that consist of the generic technical specifications,
which may be modified under Section VIII.C of this appendix, and the
remaining plant-specific information needed to complete the technical
specifications, including bracketed values.
C. Scope and Contents
The purpose of Section III of this appendix is to describe and
define the scope and contents of this design certification and to set
forth how documentation discrepancies or inconsistencies are to be
resolved. Paragraph A is the required statement of the Office of the
Federal Register (OFR) for approval of the incorporation by reference
of Tier 1, Tier 2, and the generic technical specifications into this
appendix and paragraph B requires COL applicants and licensees to
comply with the requirements of this appendix. The legal effect of
incorporation by reference is that the material is treated as if it
were published in the Federal Register. This material, like any other
properly-issued regulation, has the force and effect of law. Tier 1 and
Tier 2 information, as well as the generic technical specifications
have been combined into a single document, called the generic design
control document (DCD), in order to effectively control this
information and facilitate its incorporation by reference into the
rule. The generic DCD was prepared to meet the requirements of the OFR
for incorporation by reference (1 CFR Part 51). One of the requirements
of OFR for incorporation by reference is that the design certification
applicant must make the DCD available upon request after the final rule
becomes effective. The applicant requested the National Technical
Information Service (NTIS) to distribute the generic DCD for them.
Therefore, paragraph A states that copies of the DCD can be obtained
from NTIS, 5285 Port Royal Road, Springfield, VA 22161. The NTIS order
numbers for paper or CD-ROM copies of the ABWR DCD are PB97-147847 or
PB97-502090, respectively.
The generic DCD (master copy) for this design certification will be
archived at NRC's central file with a matching copy at OFR. Copies of
the up-to-date DCD will also be available at the NRC's Public Document
Room. Questions concerning the accuracy of information in an
application that references this appendix will be resolved by checking
the generic DCD in NRC's central file. If a generic change (rulemaking)
is made to the DCD pursuant to the change process in Section VIII of
this appendix, then at the completion of the rulemaking the NRC will
request approval of the Director, OFR for the changed incorporation by
reference and change its copies of the generic DCD and notify the OFR
and the design certification applicant to change their copies. The
Commission is requiring that the design certification applicant
maintain an up-to-date copy under X.A.1 of this appendix because it is
likely that most applicants intending to reference the standard design
will obtain the generic DCD from the design certification applicant.
Plant-specific changes to and departures from the generic DCD will be
maintained by the applicant or licensee that references this appendix
in a plant-specific DCD, under X.A.2 of this appendix.
In addition to requiring compliance with this appendix, paragraph B
clarifies that the conceptual design information and the ``Technical
Support Document for the ABWR'' are not considered to be part of this
appendix. The conceptual design information is for those portions of
the plant that are outside the scope of the standard design and are
intermingled throughout Tier 2. As provided by 10 CFR 52.47(a)(1)(ix),
these conceptual designs are not part of this appendix and, therefore,
are not applicable to an application that references this appendix.
Therefore, the applicant does not need to conform with the conceptual
design information that was provided by the design certification
applicant. The conceptual design information, which consists of site-
specific design features, was required to facilitate the design
certification review. Conceptual design information is neither Tier 1
nor Tier 2. The introduction to Tier 2 identifies the location of the
conceptual design information. The Technical Support Document provides
GE's evaluation of various design alternatives to prevent and mitigate
severe accidents, and does not constitute design requirements. The
Commission's assessment of this information is discussed in section IV
of this SOC on environmental impacts. Paragraph B also states that the
cross references from certain locations in Tier 2 of the DCD to
portions of the probabilistic risk assessment (PRA) in the ABWR
Standard Safety Analysis Report (SSAR) do not incorporate the PRA into
Tier 2. These cross references were included to clarify the format of
the DCD. The detailed methodology and quantitative portions of the
design-specific probabilistic risk assessment (PRA), as required by 10
CFR 52.47(a)(1)(v), were not included in the DCD, as requested by NEI
and the applicant for design certification. The NRC agreed with the
request to delete this information because conformance with the deleted
portions of the PRA is not necessary. Also, the NRC's position is
predicated in part upon NEI's acceptance, in conceptual form, of a
future generic rulemaking that will require a COL applicant or licensee
to have a plant-specific PRA that updates and supersedes the design-
specific PRA supporting this rulemaking and maintain it throughout the
operational life of the facility. Cross references from Tier 2 to the
proprietary and safeguards information in the ABWR SSAR do incorporate
that information into Tier 2 (refer to discussion on secondary
references).
Paragraphs C and D set forth the manner in which potential
conflicts are to be resolved. Paragraph C establishes the Tier 1
description in the DCD as controlling in the event of an inconsistency
between the Tier 1 and
[[Page 25818]]
Tier 2 information in the DCD. Paragraph D establishes the generic DCD
as the controlling document in the event of an inconsistency between
the DCD and either the application for certification of the standard
design, referred to as the Standard Safety Analysis Report, or the
final safety evaluation report for the certified design and its
supplement.
Paragraph E makes it clear that design activities that are wholly
outside the scope of this design certification may be performed using
site-specific design parameters, provided the design activities do not
affect Tier 1 or Tier 2, or conflict with the interface requirements in
the DCD. This provision applies to site-specific portions of the plant,
such as the service water intake structure. NEI requested insertion of
this clarification into the final rule (refer to its comments on the
Tier 1 definition dated July 23, 1996). Because this statement is not a
definition, the Commission decided that the appropriate location is in
Section III of the final rule.
D. Additional Requirements and Restrictions
Section IV of this appendix sets forth additional requirements and
restrictions imposed upon an applicant who references this appendix.
Paragraph IV.A sets forth the information requirements for these
applicants. This appendix distinguishes between information and/or
documents which must actually be included in the application or the
DCD, versus those which may be incorporated by reference (i.e.,
referenced in the application as if the information or documents were
actually included in the application), thereby reducing the physical
bulk of the application. Any incorporation by reference in the
application should be clear and should specify the title, date,
edition, or version of a document, and the page number(s) and table(s)
containing the relevant information to be incorporated by reference.
Paragraph A.1 requires an applicant who references this appendix to
incorporate by reference this appendix in its application. The legal
effect of such incorporation by reference is that this appendix is
legally binding on the applicant or licensee. Paragraph A.2.a is
intended to make clear that the initial application must include a
plant-specific DCD. This assures, among other things, that the
applicant commits to complying with the DCD. This paragraph also
requires the plant-specific DCD to use the same format as the generic
DCD and to reflect the applicant's proposed departures and exemptions
from the generic DCD as of the time of submission of the application.
The Commission expects that the plant-specific DCD will become the
plant's final safety analysis report (FSAR), by including within its
pages, at the appropriate points, information such as site-specific
information for the portions of the plant outside the scope of the
referenced design, including related ITAAC, and other matters required
to be included in an FSAR by 10 CFR 50.34. Integration of the plant-
specific DCD and remaining site-specific information into the plant's
FSAR, will result in an application that is easier to use and should
minimize ``duplicate documentation'' and the attendant possibility for
confusion (refer to sections II.C.3 and III.J of this SOC). Paragraph
A.2.a is also intended to make clear that the initial application must
include the reports on departures and exemptions as of the time of
submission of the application.
Paragraph A.2.b requires that the application include the reports
required by paragraph X.B of this appendix for exemptions and
departures proposed by the applicant as of the date of submission of
its application. Paragraph A.2.c requires submission of plant-specific
technical specifications for the plant that consists of the generic
technical specifications from Chapter 16 of the DCD, with any changes
made under Section VIII.C of this appendix, and the technical
specifications for the site-specific portions of the plant that are
either partially or wholly outside the scope of this design
certification, such as the ultimate heat sink. The applicant must also
provide the plant-specific information designated in the generic
technical specifications, such as bracketed values. Paragraph A.2.d
makes it clear that the applicant must provide information
demonstrating that the proposed site falls within the site parameters
for this appendix and that the plant-specific design complies with the
interface requirements, as required by 10 CFR 52.79(b).
If the proposed site has a characteristic that exceeds one or more
of the site parameters in the DCD, then the proposed site is
unacceptable for this design unless the applicant seeks an exemption
under Section VIII of this appendix and justifies why the certified
design should be found acceptable on the proposed site. Paragraph A.2.e
requires submission of information addressing COL Action Items, which
are identified in the generic DCD as COL License Information, in the
application. The COL Action Items (COL License Information) identify
matters that need to be addressed by an applicant that references this
appendix, as required by Subpart C of 10 CFR Part 52. An applicant may
depart from or omit these items, provided that the departure or
omission is identified and justified in its application (FSAR).
Paragraph A.2.f requires that the application include the information
required by 10 CFR 52.47(a) that is not within the scope of this rule,
such as generic issues that must be addressed by an applicant that
references this rule. Paragraph A.3 requires the applicant to
physically include, not simply reference, the proprietary and
safeguards information referenced in the U.S. ABWR DCD, or its
equivalent, to assure that the applicant has actual notice of these
requirements.
Paragraph IV.B reserves to the Commission the right to determine in
what manner this design certification may be referenced by an applicant
for a construction permit or operating license under 10 CFR Part 50.
This determination may occur in the context of a subsequent rulemaking
modifying 10 CFR Part 52 or this design certification rule, or on a
case-by-case basis in the context of a specific application for a Part
50 construction permit or operating license. This provision was
necessary because the evolutionary design certifications were not
implemented in the manner that was originally envisioned at the time
that Part 52 was created. The Commission's concern is with the manner
in which ITAAC were developed and the lack of experience with design
certifications in license proceedings (refer to section II.B.9 of this
SOC). Therefore, it is appropriate for the final rule to have some
uncertainty regarding the manner in which this appendix could be
referenced in a Part 50 licensing proceeding.
E. Applicable Regulations
The purpose of Section V of this appendix is to specify the
regulations that were applicable and in effect at the time that this
design certification was approved. These regulations consist of the
technically relevant regulations identified in paragraph A, except for
the regulations in paragraph B that are not applicable to this
certified design.
Paragraph A identifies the regulations in 10 CFR Parts 20, 50, 73,
and 100 that are applicable to the U.S. ABWR design. After the NRC
staff completed its FSER for the U.S. ABWR design (July 1994), the
Commission amended several existing regulations and adopted several new
regulations in those Parts of Title 10 of the Code of Federal
Regulations. The Commission has reviewed these regulations to determine
if they are
[[Page 25819]]
applicable to this design and, if so, to determine if the design meets
these regulations. The Commission finds that the U.S. ABWR design
either meets the requirements of these regulations or that these
regulations are not applicable to the design, as discussed below. The
Commission's determination of the applicable regulations was made as of
the date specified in paragraph V.A of this appendix. The specified
date is the date that this appendix was approved by the Commission and
signed by the Secretary of the Commission.
10 CFR Part 73, Protection Against Malevolent Use of Vehicles at
Nuclear Power Plants (59 FR 38889; August 1, 1994)
The objective of this regulation is to modify the design basis
threat for radiological sabotage to include use of a land vehicle by
adversaries for transporting personnel and their hand-carried equipment
to the proximity of vital areas and to include a land vehicle bomb.
This regulation also requires reactor licensees to install vehicle
control measures, including vehicle barrier systems, to protect against
the malevolent use of a land vehicle. The Commission has determined
that this regulation will be addressed in the COL applicant's site-
specific security plan. Therefore, no additional actions are required
for this design.
10 CFR 19 and 20, Radiation Protection Requirements: Amended
Definitions and Criteria (60 FR 36038; July 13, 1995)
The objective of this regulation is to revise the radiation
protection training requirement so that it applies to workers who are
likely to receive, in a year, an occupational dose in excess of 100
mrem (1 mSv); revise the definition of the ``Member of the public'' to
include anyone who is not a worker receiving an occupational dose;
revise the definition of ``Occupational Dose'' to delete reference to
location so that the occupational dose limit applies only to workers
whose assigned duties involve exposure to radiation and not to members
of the public; revise the definition of the ``Public Dose'' to apply to
doses received by members of the public from material released by a
licensee or from any other source of radiation under control of the
licensee; assure that prior dose is determined for anyone subject to
the monitoring requirements in 10 CFR Part 20, or in other words,
anyone likely to receive, in a year, 10 percent of the annual
occupational dose limit; and retain a requirement that known
overexposed individuals receive copies of any reports of the exposure
that are required to be submitted to the NRC. The Commission has
determined that these requirements will be addressed in the COL
applicant's operational radiation protection program. Therefore, no
additional actions are required for this design.
10 CFR 50, Technical Specifications (60 FR 36953; July 19, 1995)
The objective of this revised regulation is to codify criteria for
determining the content of technical specification (TS). The four
criteria were first adopted and discussed in detail in the Final Policy
Statement on Technical Specification Improvements for Nuclear Power
Reactors (58 FR 39132; July 22, 1993). The Commission has determined
that these requirements will be addressed in the COL applicant's
technical specifications. Therefore, no additional actions are required
for this design.
10 CFR 73, Changes to Nuclear Power Plant Security Requirements
Associated With Containment Access Control (60 FR 46497; September 7,
1995)
The objective of this revised regulation is to delete certain
security requirements for controlling the access of personnel and
materials into reactor containment during periods of high traffic such
as refueling and major maintenance. This action relieves nuclear power
plant licensees of requirement to separately control access to reactor
containments during these periods. The Commission has determined that
this regulation will be addressed in the COL applicant's site-specific
security plan. Therefore, no additional actions are required for this
design.
10 CFR Part 50, Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors (60 FR 49495; September 26, 1995)
The objective of this revised regulation is to provide a
performance-based option for leakage-rate testing of containments of
light-water-cooled nuclear power plants. This performance-based option,
option B to Appendix J, is available for voluntary adoption by
licensees in lieu of compliance with the prescriptive requirements
contained in the current regulation. Appendix J includes two options, A
and B, either of which can be chosen for meeting the requirements of
this appendix. The Commission has determined that option B to Appendix
J has no impact on the U.S. ABWR design because GE elected to comply
with option A.
10 CFR Parts 50, 70, and 72, Physical Security Plan Format (60 FR
53507; October 16, 1995)
The objective of this revised regulation is to eliminate the
requirement for applicants for power reactor, Category I fuel cycle,
and spent fuel storage licenses to submit physical security plans in
two parts. This action is necessary to allow for a quicker and more
efficient review of the physical security plans. The Commission has
determined that this revised regulation will be addressed in the COL
applicant's site-specific security plan. Therefore, no additional
action is required for this design.
10 CFR Part 50, Fracture Toughness Requirements for Light Water Reactor
Pressure Vessels (60 FR 65456; December 19, 1995)
The objective of this revised regulation is to clarify several
items related to fracture toughness requirements for reactor pressure
vessels (RPV). This regulation clarifies the pressurized thermal shock
(PTS) requirements, makes changes to the fractures toughness
requirements and the reactor vessel material surveillance program
requirements, and provides new requirements for thermal annealing of a
reactor pressure vessel. The Commission has determined that 10 CFR
50.61 only applies to pressurized water reactors for which an operating
license has been issued. Likewise, 10 CFR 50.66 applies only to those
light-water reactors where neutron radiation has reduced the fracture
toughness of the reactor vessel materials. Because the U.S. ABWR design
is not a pressurized water reactor and has not been licensed, neither
Secs. 50.61 nor 50.66 apply to this design or to applicants referencing
this appendix.
10 CFR Parts 21, 50, 52, 54, and 100, Reactor Site Criteria Including
Seismic and Earthquake Engineering Criteria for Nuclear Power Plants
(61 FR 65157; December 11, 1996)
The objective of this regulation is to update the criteria used in
decisions regarding power reactor siting, including geologic, seismic,
and earthquake engineering considerations for future nuclear power
plants. Two sections of this regulation apply to applications for
design certification. With regard to the revised design basis accident
radiation dose acceptance criteria in 10 CFR 50.34, the Commission has
determined that the ABWR design meets the new dose criteria, based on
the NRC staff's radiological consequence analyses,
[[Page 25820]]
provided that the site parameters are not revised. With regard to the
revised earthquake engineering criteria for nuclear power plants in
Appendix S to 10 CFR Part 50, the Commission has determined that the
ABWR design meets the new single earthquake design requirements based
on the NRC staff's evaluation in NUREG-1503. Therefore, the Commission
has determined that the ABWR design meets the applicable requirements
of this new regulation.
10 CFR Parts 20 and 35, Criteria for the Release of Individuals
Administered Radioactive Material (62 FR 4120; January 29, 1997)
The objective of this revised regulation is to specifically state
that the limitation on dose to individual members of the public in 10
CFR Part 20 does not include doses received by individuals exposed to
patients who were administered radioactive materials and released under
the new criteria in 10 CFR Part 35. This revision to Part 20 is not
applicable to the design or operation of nuclear power plants and,
therefore, does not affect the safety findings for this design.
In paragraph V.B of this appendix, the Commission identified the
regulations that do not apply to the U.S. ABWR design. The Commission
has determined that the U.S. ABWR design should be exempt from portions
of 10 CFR 50.34(f), as described in the FSER (NUREG-1503) and
summarized below:
(1) Paragraph (f)(2)(iv) of 10 CFR 50.34--Separate Plant Safety
Parameter Display Console
10 CFR 50.34(f)(2)(iv) requires that an application provide a plant
safety parameter display console that will display to operators a
minimum set of parameters defining the safety status of the plant, be
capable of displaying a full range of important plant parameters and
data trends on demand, and be capable of indicating when process limits
are being approached or exceeded.
The purpose of the requirement for a safety parameter display
system (SPDS), as stated in NUREG-0737, ``Clarification of TMI Action
Plan Requirements,'' Supplement 1, is to ``* * * provide a concise
display of critical plant variables to the control room operators to
aid them in rapidly and reliably determining the safety status of the
plant. * * * and in assessing whether abnormal conditions warrant
corrective action by operators to avoid a degraded core.''
GE committed to meet the intent of this requirement. However, the
functions of the SPDS will be integrated into the control room design
rather than on a separate ``console.'' GE has made the following
commitments in the generic DCD:
Section 18.2(6) states that the functions of the SPDS will
be integrated into the design, Section 18.4.2.1(14) states that the
SPDS function will be part of the plant summary information which is
continuously displayed on the fixed-position displays on the large
display panel,
Section 18.4.2.8 states that the information presented in
the fixed-position displays includes the critical plant parameter
information, and
Section 18.4.2.11 describes the SPDS for the ABWR and
states that the displays of critical plant variables sufficient to
provide information to plant operators about the following critical
safety functions are continuously displayed on the large display panel
as an integral part of the fixed-position displays:
(a) Reactivity control,
(b) Reactor core cooling and heat removal from the primary system,
(c) Reactor coolant system integrity, d) Radioactivity control, and
(e) Containment conditions.
In view of the above, the Commission has determined that an
exemption from the requirement for an SPDS ``console'' is justified
based upon (1) the description in the generic DCD of the intent to
incorporate the SPDS function as part of the plant status summary
information which is continuously displayed on the fixed-position
displays on the large display panel; and (2) a separate ``console'' is
not necessary to achieve the underlying purpose of the SPDS rule which
is to display to operators a minimum set of parameters defining the
safety status of the plant. Therefore, the Commission concludes that an
exemption from 10 CFR 50.34(f)(2)(iv) is justified by the special
circumstances set forth in 10 CFR 50.12(a)(2)(ii).
(2) Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-Accident Sampling for
Boron, Chloride, and Dissolved Gases
In SECY-93-087, the NRC staff recommended that the Commission
approve its position that for evolutionary and passive ALWRs of boiling
water reactor design there would be no need for the post-accident
sampling system (PASS) to analyze dissolved gases in accordance with
the requirements of 10 CFR 50.34(f)(2)(viii) and Item III.B.3 of NUREG-
0737. In its April 2, 1993, SRM, the Commission approved the
recommendation to exempt the PASS for the evolutionary and passive
ALWRs of boiling water reactor design from analyzing dissolved gases in
accordance with the requirements of 10 CFR 50.34(f)(2)(viii) and Item
III.B.3 of NUREG-0737. In SECY-93-087, the NRC staff also recommended
that the Commission approve the deviation from the requirements of Item
II.B.3 of NUREG-0737 with regard to the requirements for sampling
reactor coolant for boron concentration and activity measurements using
the PASS in evolutionary and passive ALWRs. The modified requirement
would require the capability to take boron concentration samples and
activity measurements 8 hours and 24 hours, respectively, following the
accident. In its April 2, 1993, SRM, the Commission approved the
recommendation to require the capability to take boron concentration
samples and activities measurements 8 hours and 24 hours, respectively,
following the accident.
The U.S. ABWR design will have PASS which meets the requirements of
10 CFR 50.34(f)(2)(viii) and Item II.B.3 of NUREG-0737 with the
modifications described in SECY-93-087. The system will have the
capability to sample and analyze for activity in the reactor coolant
and containment atmosphere 24 hours following the accident. This
information is needed for evaluating the conditions of the core and
will be provided during the accident management phase by the
containment high-range area monitor, the containment hydrogen monitor
and the reactor vessel water level indicator. The need for PASS
activity measurements will arise only during the accident recovery
phase and therefore, 24 hours sampling time is adequate. PASS will also
be able to determine boron concentration in the reactor coolant. It
will be capable of making this determination within 8 hours following
the accident. Knowledge of the concentration of boron is required for
providing insights for accident mitigation measures. Immediately after
the accident this information will be obtained by the neutron flux
monitoring instrumentation which is designed to comply with the
criteria of RG 1.97, and which has fully qualified redundant channels
capable of monitoring flux over the full power range. Boron
concentration measurements therefore will not be required for the first
8 hours after the accident.
For the U.S. ABWR, whenever core uncovering is suspected, the
reactor vessel is depressurized to approximately the pressure within
the wetwell and the drywell which results in partial release of the
dissolved gases. Under these conditions, pressurized samples would not
yield meaningful data. Therefore,
[[Page 25821]]
application of the regulation in this particular circumstance would not
serve the underlying purpose of the rule. During accidents when the
reactor vessel has not been depressurized (such as when a small amount
of cladding damage has occurred), reactor coolant samples can be
obtained by the process sampling system.
With regard to the need for chloride analysis, determination of
chloride concentrations is of a secondary importance because it is
needed only for determining the likelihood of accelerated primary
system corrosion which is a slow-occurring phenomenon. Chloride
analyses can be performed on the samples taken by the process sampling
system. In this case, the intended purpose of the rule can be achieved
without the need for the PASS to have chloride sampling capabilities.
Accordingly, the Commission has determined that special
circumstances required by 10 CFR 50.12(2)(ii) exist for the U.S. ABWR
in that the regulation would not serve the underlying purpose of the
rule in one circumstance and is not necessary in the other circumstance
because the intent of rule could be met with alternate design
requirements proposed by the applicant. On this basis, the Commission
concludes that the exemption from analyzing dissolved gases and
chlorides in the reactor coolant sample is justified.
(3) Paragraph (f)(3)(iv) of 10 CFR 50.34--Dedicated Containment
Penetration
Paragraph (3)(iv) of 10 CFR 50.34(f) requires one or more dedicated
containment penetrations, equivalent in size to a single .91 m (3 ft)
diameter opening, in order not to preclude future installation of
systems to prevent containment failure such as a filtered vented
containment system. This requirement is intended to ensure provision of
a containment vent design feature with sufficient safety margin well
ahead of a need that may be perceived in the future to mitigate the
consequences of a severe accident situation. The NRC staff's evaluation
of ABWR compliance with the requirement is limited to the effective
penetration size for venting provided in the U.S. ABWR primary
containment design.
The NRC staff found that the size of the primary containment
penetration that could be used during a severe accident for venting the
containment was smaller than the specific size identified in the
previous paragraph. However, in the generic DCD (Section 19A.2.44), GE
states that the containment overpressure protection system (COPS)
precludes the need for a dedicated penetration equivalent in size to a
single 0.91-m (3-ft) diameter opening. The COPS is part of the
atmospheric control system and is discussed in DCD Section 6.2.5.6. The
COPS consists of two 200-mm (8-in.) diameter rupture disks mounted in
series in a 250-mm (10-in.) line and is sized to allow 35 kg/sec (15.86
lbm/sec) of steam flow at the opening pressure of 6.3 kg/cm\2\g (90
psig), which corresponds to an energy flow of about 2.4 percent of
rated power. The DCD states that the COPS is capable of keeping
containment pressures below ASME Service Level C limits for an
anticipated transient without scram (ATWS) event with failure of the
standby liquid control system (SLCS) and containment heat removal
systems.
Although the diameter of the COPS pathway is only 200 mm (8 in.),
the NRC staff determined that this exception from the requirement of a
0.91-m (3-ft) diameter opening is acceptable because: (1) The limiting
diameter of the COPS pathway is adequate to permit the needed vent
relief path, and (2) a need for venting capability beyond that provided
by the COPS has not been identified. The Commission has determined that
GE's approach adequately addresses the requirements of this TMI item
for the ABWR design. Therefore, an exemption in accordance with 10 CFR
50.12(a)(2)(ii) is justified because the COPS provides sufficient
venting capability to preclude the need for a 0.91 m (3-ft) diameter
equivalent dedicated containment penetration.
Paragraph (b)(3) of 10 CFR 50.49--Environmental Qualification of Post-
Accident Monitoring Equipment
In the generic DCD, GE stated that the design of the information
systems important to safety will be in conformance with the guidelines
of Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-Cooled
Nuclear Power Plants to Assess Plant and Environs Conditions During and
Following an Accident,'' Revision 3. The footnote for Sec. 50.49(b)(3)
references Revision 2 of RG 1.97 for selection of the types of post-
accident monitoring equipment. As a result, the proposed design
certification rule provided an exemption to this requirement. In
section C.1 of its comments, dated August 4, 1995, ABB-CE stated that
it did not believe that an exemption from paragraph (b)(3) of 10 CFR
50.49 is needed or required. The Commission agrees with ABB-CE's
assertion that Revision 2 of RG 1.97 is identified in footnote 4 of 10
CFR 50.49 and should not be viewed as binding in this instance.
Therefore, the Commission has determined that there is no need for an
exemption from paragraph (b)(3) of 10 CFR 50.49 and has removed it from
V.B of this appendix.
F. Issue Resolution
The purpose of Section VI of this appendix is to identify the scope
of issues that are resolved by the Commission in this rulemaking and;
therefore, are ``matters resolved'' within the meaning and intent of 10
CFR 52.63(a)(4). The section is divided into five parts: (A) The
Commission's safety findings in adopting this appendix, (B) the scope
and nature of issues which are resolved by this rulemaking, (C) issues
which are not resolved by this rulemaking, (D) the backfit restrictions
applicable to the Commission with respect to this appendix, and (E)
availability of secondary references.
Paragraph A describes in general terms the nature of the
Commission's findings, and makes the finding required by 10 CFR 52.54
for the Commission's approval of this final design certification rule.
Furthermore, paragraph A explicitly states the Commission's
determination that this design provides adequate protection to the
public health and safety.
Paragraph B sets forth the scope of issues which may not be
challenged as a matter of right in subsequent proceedings. The
introductory phrase of paragraph B clarifies that issue resolution as
described in the remainder of the paragraph extends to the delineated
NRC proceedings referencing this appendix. The remaining portion of
paragraph B describes the general categories of information for which
there is issue resolution.
Specifically, paragraph B.1 provides that all nuclear safety issues
arising from the Atomic Energy Act of 1954, as amended, that are
associated with the information in the NRC staff's FSER (NUREG-1503)
and Supplement No. 1, the Tier 1 and Tier 2 information, and the
rulemaking record for this appendix are resolved within the meaning of
Sec. 52.63(a)(4). These issues include the information referenced in
the DCD that are requirements (i.e., ``secondary references''), as well
as all issues arising from proprietary and safeguards information which
are intended to be requirements. Paragraph B.2 provides for issue
preclusion of proprietary and safeguards information. As discussed in
section II.A.1 of this SOC, the inclusion of proprietary and safeguards
information within the scope of issues resolved within the meaning of
[[Page 25822]]
Sec. 52.63(a)(4) represents a change from the Commission's intent
during the proposed rule. Paragraphs B.3, B.4, B.5, and B.6 clarify
that approved changes to and departures from the DCD which are
accomplished in compliance with the relevant procedures and criteria in
Section VIII of this appendix continue to be matters resolved in
connection with this rulemaking (refer to the discussion in section
II.A.1 of this SOC). Paragraph B.7 provides that, for those plants
located on sites whose site parameters do not exceed those assumed in
Revision 1 of the Technical Support Document (December 1994), all
issues with respect to severe accident mitigation design alternatives
(SAMDAs) arising under the National Environmental Policy Act of 1969
associated with the information in the Environmental Assessment for
this design and the information regarding SAMDAs in Revision 1 of the
applicant's Technical Support Document (December 1994) are also
resolved within the meaning and intent of Sec. 52.63(a)(4). Refer to
the discussion in section II.A.1 of this SOC regarding finality of
SAMDAs in the event an exemption from a site parameter is granted. The
exemption applicant has the initial burden of demonstrating that the
original SAMDA analysis still applies to the actual site parameters
but, if the exemption is approved, requests for litigation at the COL
stage must meet the requirements of Sec. 2.714 and present sufficient
information to create a genuine controversy in order to obtain a
hearing on the site parameter exemption.
Paragraph C reserves the right of the Commission to impose
operational requirements on applicants that reference this appendix.
This provision reflects the fact that operational requirements,
including technical specifications, were not completely or
comprehensively reviewed at the design certification stage. Therefore,
the special backfit provisions of Sec. 52.63 do not apply to
operational requirements. However, all design changes would be
restricted by the appropriate provision in Section VIII of this
appendix (refer to section III.H of this SOC). Although the information
in the DCD that is related to operational requirements was necessary to
support the NRC staff's safety review of this design, the review of
this information was not sufficient to conclude that the operational
requirements are fully resolved and ready to be assigned finality under
Sec. 52.63. As a result, if the NRC wanted to change a temperature
limit on the ABWR suppression pool and that operational change required
a consequential change to an ABWR design feature, then the temperature
limit backfit would be restricted by Sec. 52.63. However, changes to
other operational issues, such as in-service testing and in-service
inspection programs, post-fuel load verification activities, and
shutdown risk that do not require a design change would not be
restricted by Sec. 52.63.
Paragraph C allows the NRC to impose future operational
requirements (distinct from design matters) on applicants who reference
this design certification. Also, license conditions for portions of the
plant within the scope of this design certification, e.g. start-up and
power ascension testing, are not restricted by Sec. 52.63. The
requirement to perform these testing programs is contained in Tier 1
information. However, ITAAC cannot be specified for these subjects
because the matters to be addressed in these license conditions cannot
be verified prior to fuel load and operation, when the ITAAC are
satisfied. Therefore, another regulatory vehicle is necessary to ensure
that licensees comply with the matters contained in the license
conditions. License conditions for these areas cannot be developed now
because this requires the type of detailed design information that will
be developed after design certification. In the absence of detailed
design information to evaluate the need for and develop specific post-
fuel load verifications for these matters, the Commission is reserving
the right to impose license conditions by rule for post-fuel load
verification activities for portions of the plant within the scope of
this design certification.
Paragraph D reiterates the restrictions (contained in 10 CFR 52.63
and Section VIII of this appendix) placed upon the Commission when
ordering generic or plant-specific modifications, changes or additions
to structures, systems or components, design features, design criteria,
and ITAAC (VI.D.3 addresses ITAAC) within the scope of the certified
design. Although the Commission does not believe that this language is
necessary, the Commission has included this language to provide a
concise statement of the scope and finality of this rule in response to
comments from NEI.
Paragraph E provides the procedure for an interested member of the
public to obtain access to proprietary and safeguards information for
the U.S. ABWR design, in order to request and participate in
proceedings identified in VI.B of this appendix, viz., proceedings
involving licenses and applications which reference this appendix. As
set forth in paragraph E, access must first be sought from the design
certification applicant. If GE Nuclear Energy refuses to provide the
information, the person seeking access shall request access from the
Commission or the presiding officer, as applicable. Access to the
proprietary and safeguards information may be ordered by the
Commission, but must be subject to an appropriate non-disclosure
agreement.
G. Duration of this Appendix
The purpose of Section VII of this appendix is in part to specify
the time period during which this design certification may be
referenced by an applicant for a combined license, pursuant to 10 CFR
52.55. This section also states that the design certification remains
valid for an applicant or licensee that references the design
certification until the application is withdrawn or the license
expires. Therefore, if an application references this design
certification during the 15-year period, then the design certification
continues in effect until the application is withdrawn or the license
issued on that application expires. Also, the design certification
continues in effect for the referencing license if the license is
renewed. The Commission intends for this appendix to remain valid for
the life of the plant that references the design certification to
achieve the benefits of standardization and licensing stability. This
means that changes to or plant-specific departures from information in
the plant-specific DCD must be made pursuant to the change processes in
Section VIII of this appendix for the life of the plant.
In its comments, dated August 3, 1995, GE noted that the proposed
design certification rule for the U.S. ABWR design indicated that the
duration was for a period of 15 years from May 8, 1995, which is
inconsistent with the provisions of 10 CFR Part 52. The date of May 8,
1995, was inserted into the proposed rule as a result of an
administrative error by the Office of the Federal Register. The
duration in the final rule is for a period of 15 years from the date of
effectiveness of the final rule, which is in accordance with 10 CFR
Part 52.
H. Processes for Changes and Departures
The purpose of Section VIII of this appendix is to set forth the
processes for generic changes to or plant-specific departures
(including exemptions) from the DCD. The Commission adopted this
restrictive change process in order to achieve a more stable licensing
process for applicants and licensees that
[[Page 25823]]
reference this design certification rule. Section VIII is divided into
three paragraphs, which correspond to Tier 1, Tier 2, and Operational
requirements. The language of Section VIII distinguishes between
generic changes to the DCD versus plant-specific departures from the
DCD. Generic changes must be accomplished by rulemaking because the
intended subject of the change is the design certification rule itself,
as is contemplated by 10 CFR 52.63(a)(1). Consistent with 10 CFR
52.63(a)(2), any generic rulemaking changes are applicable to all
plants, absent circumstances which render the change (``modification''
in the language of Sec. 52.63(a)(2)) ``technically irrelevant.'' By
contrast, plant-specific departures could be either a Commission-issued
order to one or more applicants or licensees; or an applicant or
licensee-initiated departure applicable only to that applicant's or
licensee's plant(s), i.e., a Sec. 50.59-like departure or an exemption.
Because these plant-specific departures will result in a DCD that
is unique for that plant, Section X of this appendix requires an
applicant or licensee to maintain a plant-specific DCD. For purposes of
brevity, this discussion refers to both generic changes and plant-
specific departures as ``change processes.''
Both Section VIII of this appendix and this SOC refer to an
``exemption'' from one or more requirements of this appendix and the
criteria for granting an exemption. The Commission cautions that where
the exemption involves an underlying substantive requirement
(applicable regulation), then the applicant or licensee requesting the
exemption must also show that an exemption from the underlying
applicable requirement meets the criteria of 10 CFR 50.12.
Tier 1
The change processes for Tier 1 information are covered in
paragraph VIII.A. Generic changes to Tier 1 are accomplished by
rulemaking that amends the generic DCD and are governed by the
standards in 10 CFR 52.63(a)(1). This provision provides that the
Commission may not modify, change, rescind, or impose new requirements
by rulemaking except where necessary either to bring the certification
into compliance with the Commission's regulations applicable and in
effect at the time of approval of the design certification or to ensure
adequate protection of the public health and safety or common defense
and security. The rulemakings must include an opportunity for hearing
with respect to the proposed change, as required by 10 CFR 52.63(a)(1),
and the Commission expects such hearings to be conducted in accordance
with 10 CFR Part 2, Subpart H. Departures from Tier 1 may occur in two
ways: (1) The Commission may order a licensee to depart from Tier 1, as
provided in paragraph A.3; or (2) an applicant or licensee may request
an exemption from Tier 1, as provided in paragraph A.4. If the
Commission seeks to order a licensee to depart from Tier 1, paragraph
A.3 requires that the Commission find both that the departure is
necessary for adequate protection or for compliance, and that special
circumstances are present. Paragraph A.4 provides that exemptions from
Tier 1 requested by an applicant or licensee are governed by the
requirements of 10 CFR 52.63(b)(1) and 52.97(b), which provide an
opportunity for a hearing. In addition, the Commission will not grant
requests for exemptions that may result in a significant decrease in
the level of safety otherwise provided by the design (refer to
discussion in II.A.3 of this SOC).
Tier 2
The change processes for the three different categories of Tier 2
information, viz., Tier 2, Tier 2 *, and Tier 2 * with a time of
expiration are set forth in paragraph VIII.B. The change process for
Tier 2 has the same elements as the Tier 1 change process, but some of
the standards for plant-specific orders and exemptions are different.
The Commission also adopted a ``Sec. 50.59-like'' change process in
accordance with its SRMs on SECY-90-377 and SECY-92-287A.
The process for generic Tier 2 changes (including changes to Tier 2
* and Tier 2 * with a time of expiration) tracks the process for
generic Tier 1 changes. As set forth in paragraph B.1, generic Tier 2
changes are accomplished by rulemaking amending the generic DCD, and
are governed by the standards in 10 CFR 52.63(a)(1). This provision
provides that the Commission may not modify, change, rescind or impose
new requirements by rulemaking except where necessary either to bring
the certification into compliance with the Commission's regulations
applicable and in effect at the time of approval of the design
certification or to assure adequate protection of the public health and
safety or common defense and security. If a generic change is made to
Tier 2 * information, then the category and expiration, if necessary,
of the new information would also be determined in the rulemaking and
the appropriate change process for that new information would apply
(refer to II.A.2 of this SOC).
Departures from Tier 2 may occur in five ways: (1) the Commission
may order a plant-specific departure, as set forth in paragraph B.3;
(2) an applicant or licensee may request an exemption from a Tier 2
requirement as set forth in paragraph B.4; (3) a licensee may make a
departure without prior NRC approval in accordance with paragraph B.5
[the ``Sec. 50.59-like'' process]; (4) the licensee may request NRC
approval for proposed departures which do not meet the requirements in
paragraph B.5 as provided in paragraph B.5.d; and (5) the licensee may
request NRC approval for a departure from Tier 2 * information, in
accordance with paragraph B.6.
Similar to Commission-ordered Tier 1 departures and generic Tier 2
changes, Commission-ordered Tier 2 departures cannot be imposed except
where necessary either to bring the certification into compliance with
the Commission's regulations applicable and in effect at the time of
approval of the design certification or to ensure adequate protection
of the public health and safety or common defense and security, as set
forth in paragraph B.3. However, the special circumstances for the
Commission-ordered Tier 2 departures do not have to outweigh any
decrease in safety that may result from the reduction in
standardization caused by the plant-specific order, as required by 10
CFR 52.63(a)(3). The Commission determined that it was not necessary to
impose an additional limitation similar to that imposed on Tier 1
departures by 10 CFR 52.63(a)(3) and (b)(1). This type of additional
limitation for standardization would unnecessarily restrict the
flexibility of applicants and licensees with respect to Tier 2, which
by its nature is not as safety significant as Tier 1.
An applicant or licensee may request an exemption from Tier 2
information as set forth in paragraph B.4. The applicant or licensee
must demonstrate that the exemption complies with one of the special
circumstances in 10 CFR 50.12(a). In addition, the Commission will not
grant requests for exemptions that may result in a significant decrease
in the level of safety otherwise provided by the design (refer to
discussion in II.A.3 of this SOC). However, the special circumstances
for the exemption do not have to outweigh any decrease in safety that
may result from the reduction in standardization caused by the
exemption. If the exemption is requested by an applicant for a license,
the exemption is subject to litigation in the same manner as other
issues in the license hearing, consistent with 10 CFR
[[Page 25824]]
52.63(b)(1). If the exemption is requested by a licensee, then the
exemption is subject to litigation in the same manner as a license
amendment.
Paragraph B.5 allows an applicant or licensee to depart from Tier 2
information, without prior NRC approval, if the proposed departure does
not involve a change to or departure from Tier 1 or Tier 2 *
information, technical specifications, or involves an unreviewed safety
question (USQ) as defined in B.5.b and B.5.c of this paragraph. The
technical specifications referred to in B.5.a and B.5.b of this
paragraph are the technical specifications in Chapter 16 of the generic
DCD, including bases, for departures made prior to issuance of the COL.
After issuance of the COL, the plant-specific technical specifications
are controlling under paragraph B.5 (refer to discussion in II.A.1 of
this SOC on Finality for Technical Specifications). The bases for the
plant-specific technical specifications will be controlled by the bases
control procedures for the plant-specific technical specifications
(analogous to the bases control provision in the Improved Standard
Technical Specifications). The definition of a USQ in paragraph B.5.b
is similar to the definition in 10 CFR 50.59 and it applies to all
information in Tier 2 except for the information that resolves the
severe accident issues. The process for evaluating proposed tests or
experiments not described in Tier 2 will be incorporated into the
change process for the portion of the design that is outside the scope
of this design certification. Although paragraph B.5 does not
specifically state, the Commission has determined that departures must
also comply with all applicable regulations unless an exemption or
other relief is obtained.
The Commission believes that it is important to preserve and
maintain the resolution of severe accident issues just like all other
safety issues that were resolved during the design certification review
(refer to SRM on SECY-90-377). However, because of the increased
uncertainty in severe accident issue resolutions, the Commission has
adopted separate criteria in B.5.c for determining whether a departure
from information that resolves severe accident issues constitutes a
USQ. For purposes of applying the special criteria in B.5.c, severe
accident resolutions are limited to design features when the intended
function of the design feature is relied upon to resolve postulated
accidents where the reactor core has melted and exited the reactor
vessel and the containment is being challenged (refer to discussion in
II.A.2 of this SOC). These design features are identified in Section
19.11 of the System 80+ DCD and Section 19E of the ABWR DCD, but may be
described in other sections of the DCD. Therefore, the location of
design information in the DCD is not important to the application of
this special procedure for severe accident issues. However, the special
procedure in B.5.c does not apply to design features that resolve so-
called beyond design basis accidents or other low probability events.
The important aspect of this special procedure is that it is limited
solely to severe accident design features, as defined above. Some
design features of the evolutionary designs have intended functions to
meet both ``design basis'' requirements and to resolve ``severe
accidents.'' If these design features are reviewed under paragraph
VIII.B.5, then the appropriate criteria from either B.5.b or B.5.c are
selected depending upon the design function being changed.
An applicant or licensee that plans to depart from Tier 2
information, under VIII.B.5, must prepare a safety evaluation which
provides the bases for the determination that the proposed change does
not involve an unreviewed safety question, a change to Tier 1 or Tier
2* information, or a change to the technical specifications, as
explained above. In order to achieve the Commission's goals for design
certification, the evaluation needs to consider all of the matters that
were resolved in the DCD, such as generic issue resolutions that are
relevant to the proposed departure. The benefits of the early
resolution of safety issues would be lost if departures from the DCD
were made that violated these resolutions without appropriate review.
The evaluation of the relevant matters needs to consider the proposed
departure over the full range of power operation from startup to
shutdown, as it relates to anticipated operational occurrences,
transients, design basis accidents, and severe accidents. The
evaluation must also include a review of all relevant secondary
references from the DCD because Tier 2 information intended to be
treated as requirements is contained in the secondary references. The
evaluation should consider the tables in Sections 14.3 and 19.8 of the
DCD to ensure that the proposed change does not impact Tier 1. These
tables contain various cross-references from the plant safety analyses
in Tier 2 to the important parameters that were included in Tier 1.
Although many issues and analyses could have been cross-referenced, the
listings in these tables were developed only for key plant safety
analyses for the design. GE provided more detailed cross-references to
Tier 1 for these analyses in a letter dated March 31, 1994.
If a proposed departure from Tier 2 involves a change to or
departure from Tier 1 or Tier 2* information, technical specifications,
or otherwise constitutes a USQ, then the applicant or licensee must
obtain NRC approval through the appropriate process set forth in this
appendix before implementing the proposed departure. The NRC does not
endorse NSAC-125, ``Guidelines for 10 CFR 50.59 Safety Evaluations,''
for performing safety evaluations required by VIII.B.5 of this
appendix. However, the NRC will work with industry, if it is desired,
to develop an appropriate guidance document for processing proposed
changes under VIII.B of this appendix.
A party to an adjudicatory proceeding (e.g., for issuance of a
combined license) who believes that an applicant or licensee has not
complied with VIII.B.5 when departing from Tier 2 information, may
petition to admit such a contention into the proceeding. As set forth
in B.5.f, the petition must comply with the requirements of
Sec. 2.714(b)(2) and show that the departure does not comply with
paragraph B.5. Any other party may file a response to the petition. If
on the basis of the petition and any responses, the presiding officer
in the proceeding determines that the required showing has been made,
the matter shall be certified to the Commission for its final
determination. In the absence of a proceeding, petitions alleging non-
conformance with paragraph B.5 requirements applicable to Tier 2
departures will be treated as petitions for enforcement action under 10
CFR 2.206.
Paragraph B.6 provides a process for departing from Tier 2*
information. This provision is bifurcated because of the expiration of
some Tier 2* information. The Commission determined that the Tier 2*
designation should expire for some Tier 2* information in response to
comments from NEI (refer to section II.A.2 of this SOC). Therefore,
certain Tier 2* information listed in B.6.c is no longer designated as
Tier 2* information after full power operation is first achieved
following the Commission finding in 10 CFR 52.103(g). Thereafter, that
information is deemed to be Tier 2 information that is subject to the
departure requirements in paragraph B.5. By contrast, the Tier 2*
information identified in B.6.b retains its Tier 2* designation
throughout the duration of the license, including any period of
[[Page 25825]]
renewal. Any requests for departures from Tier 2* information that
affect Tier 1 must also comply with the requirements in VIII.A of this
appendix.
If Tier 2* information is changed in a generic rulemaking, the
designation of the new information (Tier 1, 2*, or 2) would also be
determined in the rulemaking and the appropriate process for future
changes would apply. If a plant-specific departure is made from Tier 2*
information, then the new designation would apply only to that plant.
If an applicant who references this design certification makes a
departure from Tier 2* information, the new information is subject to
litigation in the same manner as other plant-specific issues in the
licensing hearing (refer to B.6.a). If a licensee makes a departure, it
will be treated as a license amendment under 10 CFR 50.90 and the
finality is in accordance with paragraph VI.B.5 of this appendix.
Operational Requirements
The change process for technical specifications and other
operational requirements is set forth in paragraph VIII.C. This change
process has elements similar to the Tier 1 and Tier 2 change process in
paragraphs VIII.A and VIII.B, but with significantly different change
standards (refer to the explanation in II.A.1 of this SOC). The
Commission did not support NEI's request to extend the special backfit
provisions of 10 CFR 52.63 to technical specifications and other
operational requirements (refer to explanation in III.F of this SOC).
Rather, the Commission decided to designate a special category of
information, consisting of the technical specifications and other
operational requirements, with its own change process in paragraph
VIII.C. The key to using the change processes in Section VIII is to
determine if the proposed change or departure requires a change to a
design feature described in the generic DCD. If a design change is
required, then the appropriate change process in paragraph VIII.A or
VIII.B applies. However, if a proposed change to the technical
specifications or other operational requirements does not require a
change to a design feature in the generic DCD, then paragraph VIII.C
applies. The language in paragraph VIII.C also distinguishes between
generic and plant-specific technical specifications to account for the
different treatment and finality accorded technical specifications
before and after a license is issued.
The process in C.1 for making generic changes to the generic
technical specifications in Chapter 16 of the DCD or other operational
requirements in the generic DCD is accomplished by rulemaking and
governed by the backfit standards in 10 CFR 50.109. The determination
of whether the generic technical specifications and other operational
requirements were completely reviewed and approved in the design
certification rulemaking is based upon the extent to which an NRC
safety conclusion in the FSER or its supplement is being modified or
changed. If it cannot be determined that the technical specification or
operational requirement was comprehensively reviewed and finalized in
the design certification rulemaking, then there is no backfit
restriction under 10 CFR 50.109 because no prior position was taken on
this safety matter. Some generic technical specifications contain
bracketed values, which clearly indicate that the NRC staff's review
was not complete. Generic changes made under VIII.C.1 are applicable to
all applicants or licensees, unless the change is irrelevant because of
a plant-specific departure (refer to VIII.C.2).
Plant-specific departures may occur by either a Commission order
under VIII.C.3 or an applicant's exemption request under VIII.C.4. The
basis for determining if the technical specification or operational
requirement was completely reviewed and approved is the same as for
VIII.C.1 above. If the technical specification or operational
requirement was comprehensively reviewed and finalized in the design
certification rulemaking, then the Commission must demonstrate that
special circumstances are present before ordering a plant-specific
departure. If not, there is no restriction on plant-specific changes to
the technical specifications or operational requirements, prior to
issuance of a license, provided a design change is not required.
Although the generic technical specifications were reviewed by the NRC
staff to facilitate the design certification review, the Commission
intends to consider the lessons learned from subsequent operating
experience during its licensing review of the plant-specific technical
specifications. The process for petitioning to intervene on a technical
specification or operational requirement is similar to other issues in
a licensing hearing, except that the petitioner must also demonstrate
why special circumstances are present (refer to VIII.C.5).
Finally, the generic technical specifications will have no further
effect on the plant-specific technical specifications after the
issuance of a license that references this appendix (refer to sections
II.A.1 and II.B.3 of this SOC). The bases for the generic technical
specifications will be controlled by the change process in Section
VIII.C of this appendix. After a license is issued, the bases will be
controlled by the bases change provision set forth in the
administrative controls section of the plant-specific technical
specifications.
I. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
The purpose of Section IX of this appendix is to set forth how the
ITAAC in Tier 1 of this design certification rule are to be treated in
a license proceeding. Paragraph A restates the responsibilities of an
applicant or licensee for performing and successfully completing ITAAC,
and notifying the NRC of such completion. Paragraph A.1 makes it clear
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 successfully
completed. Paragraph A.2 requires 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.
Paragraphs B.1 and B.2 essentially reiterate the NRC's
responsibilities with respect to ITAAC as set forth in 10 CFR 52.99 and
52.103(g) [refer to explanation in section II.C.1 of this SOC].
Finally, paragraph B.3 states that ITAAC do not, by virtue of their
inclusion in the DCD, 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 DCD unless the applicable requirements in 10 CFR
52.97 and Section VIII of this appendix have been complied with. As
discussed in sections II.B.9 and III.D of this SOC, the Commission will
defer a determination of the applicability of ITAAC and their effect in
terms of issue resolution in 10 CFR Part 50 licensing proceedings to
such time that a Part 50 applicant decides to reference this appendix.
J. Records and Reporting
The purpose of Section X of this appendix is to set forth the
requirements for maintaining records of changes to and departures from
the generic DCD, which are to be reflected in the plant-
[[Page 25826]]
specific DCD. Section X also sets forth the requirements for submitting
reports (including updates to the plant-specific DCD) to the NRC. This
section of the appendix is similar to the requirements for records and
reports in 10 CFR Part 50, except for minor differences in information
collection and reporting requirements, as discussed in section V of
this SOC. Paragraph X.A.1 of this appendix requires that a generic DCD
and the proprietary and safeguards information referenced in the
generic DCD be maintained by the applicant for this rule. The generic
DCD was developed, in part, to meet the requirements for incorporation
by reference, including availability requirements. Therefore, the
proprietary and safeguards information could not be included in the
generic DCD because it is not publicly available. However, the
proprietary and safeguards information was reviewed by the NRC and, as
stated in paragraph VI.B.2 of this appendix, the Commission considers
the information to be resolved within the meaning of 10 CFR
52.63(a)(4). Because this information is not in the generic DCD, the
proprietary and safeguards information, or its equivalent, is required
to be provided by an applicant for a license. Therefore, to ensure that
this information will be available, a requirement for the design
certification applicant to maintain the proprietary and safeguards
information was added to paragraph X.A.1 of this appendix. The
acceptable version of the proprietary and safeguards information is
identified in the version of the DCD that is incorporated into this
rule. The generic DCD and the acceptable version of the proprietary and
safeguards information must be maintained for the period of time that
this appendix may be referenced.
Paragraphs A.2 and A.3 place record-keeping requirements on the
applicant or licensee that references this design certification to
maintain its plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made pursuant
to Section VIII of this appendix. The term ``plant-specific'' was added
to paragraph A.2 and other Sections of this appendix to distinguish
between the generic DCD that is incorporated by reference into this
appendix, and the plant-specific DCD that the applicant is required to
submit under IV.A of this appendix. The requirement to maintain the
generic changes to the generic DCD is explicitly stated to ensure that
these changes are not only reflected in the generic DCD, which will be
maintained by the applicant for design certification, but that the
changes are also reflected in the plant-specific DCD. Therefore,
records of generic changes to the DCD will be required to be maintained
by both entities to ensure that both entities have up-to-date DCDs.
Section X.A of this appendix does not place record-keeping
requirements on site-specific information that is outside the scope of
this rule. As discussed in section III.D of this SOC, the final safety
analysis report required by 10 CFR 52.79 will contain the plant-
specific DCD and the site-specific information for a facility that
references this rule. The phrase ``site-specific portion of the final
safety analysis report'' in paragraph X.B.3.d of this appendix refers
to the information that is contained in the final safety analysis
report for a facility (required by 10 CFR 52.79) but is not part of the
plant-specific DCD (required by IV.A of this appendix). Therefore, this
rule does not require that duplicate documentation be maintained by an
applicant or licensee that references this rule, because the plant-
specific DCD is part of the final safety analysis report for the
facility (refer to section II.C.3 of this SOC).
Paragraphs B.1 and B.2 establish reporting requirements for
applicants or licensees that reference this rule that are similar to
the reporting requirements in 10 CFR Part 50. For currently operating
plants, a licensee is required to maintain records of the basis for any
design changes to the facility made under 10 CFR 50.59. Section
50.59(b)(2) requires a licensee to provide a summary report of these
changes to the NRC annually, or along with updates to the facility
final safety analysis report under 10 CFR 50.71(e). Section 50.71(e)(4)
requires that these updates be submitted annually, or 6 months after
each refueling outage if the interval between successive updates does
not exceed 24 months.
The reporting requirements vary according to four different time
periods during a facilities' lifetime as specified in paragraph B.3.
Paragraph B.3.a requires that if an applicant that references this rule
decides to make departures from the generic DCD, then the departures
and any updates to the plant-specific DCD must be submitted with the
initial application for a license. Under B.3.b, the applicant may
submit any subsequent reports and updates along with its amendments to
the application provided that the submittals are made at least once per
year. Because amendments to an application are typically made more
frequently than once a year, this should not be an excessive burden on
the applicant.
Paragraph B.3.c requires that the reports be submitted quarterly
during the period of facility construction. This increase in frequency
of summary reports of departures from the plant-specific DCD is in
response to the Commission's guidance on reporting frequency in its SRM
on SECY-90-377, dated February 15, 1991. NEI stated in its comments
dated August 4, 1995 (Attachment B, p. 116) that * * * ``the
requirement for quarterly reporting imposes unnecessary additional
burdens on licensees and the NRC.'' NEI recommended that the Commission
adopt a ``less onerous'' requirement (e.g., semi-annual reports). The
Commission disagrees with the NEI request because it does not provide
for sufficiently timely notification of design changes during the
critical period of facility construction. Also, the Commission
disagrees that the reports are an onerous burden because they are only
summary reports, which describe the design changes, rather than
detailed evaluations of the changes and determinations. The detailed
evaluations remain available for audit on site, consistent with the
requirements of 10 CFR Part 50.
Quarterly reporting of design changes during the period of
construction is necessary to closely monitor the status and progress of
the construction of the plant. To make its finding under 10 CFR 52.99,
the NRC must monitor the design changes made in accordance with Section
VIII of this appendix. The ITAAC verify that the as-built facility
conforms with the approved design and emphasizes design reconciliation
and design verification. Quarterly reporting of design changes is
particularly important in times where the number of design changes
could be significant, such as during the procurement of components and
equipment, detailed design of the plant at the start of construction,
and during pre-operational testing. The frequency of updates to the
plant-specific DCD is not increased during facility construction. After
the facility begins operation, the frequency of reporting reverts to
the requirement in paragraph X.B.3.d, which is consistent with the
requirement for plants licensed under 10 CFR Part 50.
IV. Finding of No Significant Environmental Impact: Availability
The Commission has determined under the National Environmental
Policy Act of 1969, as amended (NEPA), and the Commission's regulations
in 10 CFR Part 51, Subpart A, that this design certification rule is
not a major Federal action significantly affecting the quality
[[Page 25827]]
of the human environment and, therefore, an environmental impact
statement (EIS) is not required. The basis for this determination, as
documented in the final environmental assessment, is that this
amendment to 10 CFR Part 52 does not authorize the siting,
construction, or operation of a facility using the U.S. ABWR design; it
only codifies the U.S. ABWR design in a rule. The NRC will evaluate the
environmental impacts and issue an EIS as appropriate in accordance
with NEPA as part of the application(s) for the construction and
operation of a facility.
In addition, as part of the final environmental assessment for the
U.S. ABWR design, the NRC reviewed GE's evaluation of various design
alternatives to prevent and mitigate severe accidents that was
submitted in GE's ``Technical Support Document for the ABWR,'' Rev. 1,
dated December 1994. The Commission finds that GE's evaluation provides
a sufficient basis to conclude that there are no additional severe
accident design alternatives beyond those currently incorporated into
the U.S. ABWR design which are cost-beneficial, whether considered at
the time of the approval of the U.S. ABWR design certification or in
connection with the licensing of a future facility referencing the U.S.
ABWR design certification, where the plant referencing this appendix is
located on a site whose site parameters are within those specified in
the Technical Support Document. These issues are considered resolved
for the U.S. ABWR design.
The final environmental assessment, upon which the Commission's
finding of no significant impact is based, and the Technical Support
Document for the U.S. ABWR design are available for examination and
copying at the NRC Public Document Room, 2120 L Street, NW. (Lower
Level), Washington, DC. Single copies are also available from Mr. Dino
C. Scaletti, Mailstop O-11 H3, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, (301) 415-1104.
V. Paperwork Reduction Act Statement
This final rule amends information collection requirements that are
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These requirements were approved by the Office of Management and
Budget, approval number 3150-0151. Should an application be received,
the additional public reporting burden for this collection of
information, above those contained in Part 52, is estimated to average
8 hours per response, including the time for reviewing instructions,
searching existing data sources, gathering and maintaining the data
needed, and completing and reviewing the collection of information.
Send comments on any aspect of this collection of information,
including suggestions for reducing the burden, to the Information and
Records Management Branch (T-6 F33), U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, or by Internet electronic mail
at [email protected]; and to the Desk Officer, Office of Information and
Regulatory Affairs, NEOB-10202, (3150-0151), Office of Management and
Budget, Washington, DC 20503.
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless it displays a currently
valid OMB control number.
VI. Regulatory Analysis
The NRC has not prepared a regulatory analysis for this final rule.
The NRC prepares regulatory analyses for rulemakings that establish
generic regulatory requirements applicable to all licensees. Design
certifications are not generic rulemakings in the sense that design
certifications do not establish standards or requirements with which
all 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. Preparation
of a regulatory analysis in this circumstance would not be useful
because the design to be certified is proposed by the applicant rather
than the NRC. For these reasons, the Commission concludes that
preparation of a regulatory analysis is neither required nor
appropriate.
VII. Regulatory Flexibility Act Certification
In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C.
605(b), the Commission certifies that this rulemaking will not have a
significant economic impact upon a substantial number of small
entities. The rule provides certification for a nuclear power plant
design. Neither the design certification applicant nor prospective
nuclear power plant licensees who reference this design certification
rule fall within the scope of the definition of ``small entities'' set
forth in the Regulatory Flexibility Act, 15 U.S.C. 632, or the Small
Business Size Standards set out in regulations issued by the Small
Business Administration in 13 CFR Part 121. Thus, this rule does not
fall within the purview of the act.
VIII. Backfit Analysis
The Commission has determined that the backfit rule, 10 CFR 50.109,
does not apply to this final rule because these amendments do not
impose requirements on existing 10 CFR Part 50 licensees. Therefore, a
backfit analysis was not prepared for this rule.
List of Subjects in 10 CFR Part 52
Administrative practice and procedure, Antitrust, Backfitting,
Combined license, Early site permit, Emergency planning, Fees,
Incorporation by reference, Inspection, Limited work authorization,
Nuclear power plants and reactors, Probabilistic risk assessment,
Prototype, Reactor siting criteria, Redress of site, Reporting and
record keeping requirements, Standard design, Standard design
certification.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; and 5 U.S.C. 552 and 553; the NRC is adopting
the following amendments to 10 CFR Part 52.
1. The authority citation for 10 CFR Part 52 continues to read as
follows:
Authority: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat.
936, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244,
as amended (42 U.S.C. 2133, 2201, 2232, 2233, 2236, 2239, 2282);
secs. 201, 202, 206, 88 Stat. 1243, 1244, 1246, 1246, as amended (42
U.S.C. 5841, 5842, 5846).
2. In Sec. 52.8, paragraph (b) is revised to read as follows:
Sec. 52.8 Information collection requirements: OMB approval.
* * * * *
(b) The approved information collection requirements contained in
this part appear in Secs. 52.15, 52.17, 52.29, 52.45, 52.47, 52.57,
52.75, 52.77, 52.78, 52.79, Appendix A, and Appendix B.
3. A new Appendix A to 10 CFR Part 52 is added to read as follows:
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
[[Page 25828]]
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 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 shall 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 VIII.B.6 of this appendix. This designation
expires for some Tier 2* information under VIII.B.6.
G. All other terms in this appendix have the meaning set out in
10 CFR 50.2, 10 CFR 52.3, 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, VA 22161. A copy is available for
examination and copying at the NRC Public Document Room, 2120 L
Street NW. (Lower Level), Washington, DC 20555. Copies are also
available for examination at the NRC Library, 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-specific design parameters, 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 information and
utilizing 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 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
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)(4) in subsequent proceedings
for issuance of a combined license, amendment of a combined license,
or renewal of a combined license, proceedings held pursuant to 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
[[Page 25829]]
requirements in the generic DCD for the U.S. ABWR design;
3. All generic changes to the DCD pursuant to 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 pursuant to and in compliance
with the change processes in Sections VIII.A.4 and VIII.B.4 of this
appendix, but only for that proceeding;
5. All departures from the DCD that are approved by license
amendment, but only for that proceeding;
6. Except as provided in VIII.B.5.f of this appendix, all
departures from Tier 2 pursuant to and in compliance with the change
processes in VIII.B.5 of this appendix that do not require prior NRC
approval;
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)(4). 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 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 GE Nuclear
Energy declines to provide the information sought, GE Nuclear Energy
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 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
July 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)(3).
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 Secs. 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 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 technical specifications,
or involves an unreviewed safety question as defined in paragraphs
B.5.b and 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, involves an unreviewed safety question if--
(1) The probability of occurrence or the consequences of an
accident or malfunction of equipment important to safety previously
evaluated in the plant-specific DCD may be increased;
(2) A possibility for an accident or malfunction of a different
type than any evaluated previously in the plant-specific DCD may be
created; or
(3) The margin of safety as defined in the basis for any
technical specification is reduced.
c. A proposed departure from Tier 2 affecting resolution of a
severe accident issue identified in the plant-specific DCD, involves
an unreviewed safety question 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 involves an unreviewed safety question as
defined in paragraph B.5 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 25830]]
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 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 2.714(b)(2), the
petition must demonstrate that the departure does not comply with
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 fact regarding compliance with
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)(4).
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 N-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 2.758(b) 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 2.714(b)(2) and must
demonstrate why special circumstances as defined in 10 CFR 2.758(b)
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
satisfied.
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 satisfied, 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 52.99 and 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 Section 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
[[Page 25831]]
10 CFR 52.97 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
pursuant to 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 safety 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
departures from the plant-specific DCD, including a summary of the
safety evaluation of each. This report must be filed in accordance
with the filing requirements applicable to reports in 10 CFR 50.4.
2. An applicant or licensee who references this appendix shall
submit updates to its plant-specific DCD, which reflect the generic
changes to the generic DCD and the plant-specific departures made
pursuant to Section VIII of this appendix. These updates shall be
filed in accordance with the filing requirements applicable to final
safety analysis report updates in 10 CFR 50.4 and 50.71(e).
3. The reports and updates required by paragraphs B.1 and B.2 of
this section must be submitted as follows:
a. On the date that an application for a license referencing
this appendix is submitted, the application shall include the report
and any updates to the plant-specific DCD.
b. During the interval from the date of application to the date
of issuance of a license, the report and any updates to the plant-
specific DCD must be submitted annually and may be submitted along
with amendments to the application.
c. During the interval from the date of issuance of a license to
the date the Commission makes its findings under 10 CFR 52.103(g),
the report must be submitted quarterly. Updates to the plant-
specific DCD must be submitted annually.
d. After the Commission has made its finding under 10 CFR
52.103(g), reports and updates to the plant-specific DCD may be
submitted annually or along with updates to the site-specific
portion of the final safety analysis report for the facility at the
intervals required by 10 CFR 50.71(e), or at shorter intervals as
specified in the license.
Dated at Rockville, Maryland, this 2nd day of May, 1997.
For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 97-11968 Filed 5-9-97; 8:45 am]
BILLING CODE 7590-01-P