[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