ML20206F707

Requests Commission Approval to Publish in Fr Attached Proposed Rule That Would Amend 10CFR52 to Certify AP600 Standard Plant Design

ML20206F707
Person / Time
Site:	05200003
Issue date:	03/31/1999
From:	Travers W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
SECY-99-101, SECY-99-101-01, SECY-99-101-1, SECY-99-101-R, NUDOCS 9905060215
Download: ML20206F707 (79)

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RULEMAKING ISSUE v2-3 (Notation Vote)

March 31.1999 SECY-99-101 EQB: The Commissioners s FROM: William D. Travers Executive Director for Operations

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SUBJECT:

PROPOSED RULE - AP600 DESIGN CERTIFICATION E

. PURPOSE:

To obtain the Commission's approval to publish in the Federal Reaister the attached proposed rule that would amend 10 CFR Part 52 to certify the AP600 standard plant design.

BACKGROUND:

Westinghouse Electric Company submitted an application for certification of its AP600 standard plant design on June 26,1992. The NRC staff issued a final design approval to Westinghouse on September 3.1998, that signified completion of the technical review phase and readiness for

- the rulemaking phase of the AP600 application. The Commission approved the rulemaking plan for the AP600 design in its staff requirements memorandum dated December 4,1998.

DISCUSSION:

The NRC staff completed its review of the AP600 standard p! ant design and issued NUREG-1512," Final Safety Evaluation Report related to Certification of the AP600 Standard Design,"in J September 1998 (see COMSECY-98-025). Certification of the AP600 standard plant design /

will be performed under Subpart B of 10 CFR Part 52 and in accordance with SECY-98-267, "Rulemaking Plan for the AP600," dated November 16,1998.

I This proposed design certification rule (DCR) is nearly identical to the two previously issued -

DCRs for the U.S. ABWR and System 80+ designs (Appendices A and B to 10 CFR Part 52, respectively). The staff believes that the AP600 DCR should emulate the existing DCRs for the ABWR and the System 80+, inasmuch as the three designs were reviewed contemporaneously against the same technical requirements. Furthermore, many of the procedural issues and their resolutions for the ABWR and the System 80+ DORS (gfi., the two-tier structure, Tier 2*,

[ CONTACT:

o Jerry N. Wilson, NRR/ DRIP uGO O 4,1 415-3145 9905060215 990331 PDR SECY W 101 R PDR

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The Commissioners  !

and the scope of issue resolution) were developed after extensive discussions with nuclear industry representatives, and Westinghouse participated in those discussions. It was the NRC's intent (and likely Westiaghouse's expectation) that the resolutions for these issues in the ABWR and System 80+ rulemakings would also be applied to the AP600 design. Accordingly, the staff has modeled the AP600 DCR on the existing DCRs for the ABWR and System 80+, I with certain departures. The departures from these DCRs were necessary to account for different applicants; design documentation, including Tier 2* Information; design features; and the environmental assessment. The only significant change was the inclusion of the investment )

l protection short-term availability controls in Sections ll, lil, and VI of the AP600 DCR.

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Westinghouse was notified by letters dated June 9 and August 2,1997, that these availability controls would be binding on applicants and licensees that reference the AP600 DCR.

The attached Federal Reaister notice provides the public with an opportunity to comment en the ,

proposed DCR; the AP600 Design Control Document, which is incorporated into the DCR by l reference; and the environmental assessment. The Federal Reaister notice also provides the 1 public with an opportunity to request an informal hearing under 10 CFR 52.51(b) and Section ll

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of the notice. The time period for submitting comments or requesting an informal hearing was 1 120 days for the previous DCRs, in accordance with SECY-92-381, "Rulemaking Procedures for Design Certification," dated November 10,1992. The time period for commenting on the proposed AP600 DCR is 75 days, under the North American Free Trade Agreement. The staff believes that a 75-day comment period is sufficient for the AP600 DCR because of the multiple comment periods and extensive interactions with stakeholders, including Westinghouse, on the previously issued DCRs (see SECY-96-077, " Certification of Two Evolutionary Designs").

COORDINATION:

The Office of the General Counsel has reviewed this paper and has no legal objections. The Chief Financial Officer has reviewed this paper for resource implications and has no objections.

The Chief Information Officer has reviewed this paper for information technology and information management implications and concurs in it. However, the proposed DCR requires ,

a change in the information collection requirements that will require a submission to the Office i of Management and Budget. A copy of the attached Feaeral Reaister notice was provided to l the ACRS for its consideration. l RECOMMENDATION:

That the Commission:

1. Acorove the proposed amendment to 10 CFR Part 52 for publication in the Federal Register.

2. Certify that this rule, if promulgated, will not have a negative economic impact on a substantial number of small entities in order to satisfy requirements of the Regulatory Flexibility Act,5 U.S.C. 605(b).

The Commissioners 3. Determine that the backfit rule,10 CFR 50.109, does not apply to this proposed rule.

4. HQ1g:

a. The proposed DCR will be published in the Federa/ Registerfor a 75-day comment period and an opportunity to request an informal hearing;

b. An environmental assessment and a finding of no significant impact have been prepared (Attachment 2);

c. This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This rule will be submitted to the Office of Management and Budget for review and approval of the paperwork requirements (Section V of Attachment 1);

d. The Chief Counsel for Advocacy of the Small Business Administration will be informed of the certification regarding the economic impact on small entities and the reasons for it as required by the Regulatory Flexibility Act (Section Vil);

e. The appropriate congressional committees will be informed (Attachment 3); and

f. The Office of Public Affairs willissue a press release (Attachment 4).

m illiam D. ve f ecutive Director i for Operations i Attachments:

1. Federal Reaister notice  ;

2. Environmental Assessment '

3. Congressional Letters

4. Press Release Commissioners' completed vote sheets / comme nts should be provided directly to the_ Office of the Secretary by COB Friday, April 16, 1999.

Commission Staff Office comments, if any, should be submitted to the Commissioners NLT April 9,.1999, with an information copy to the Office of the Secretary. If the paper is of such a nature that it requires additional review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected. .

DISTRIBUTION:

Commissioners CIO OGC CFO OCAA EDO OIG SECY OPA OCA ACRS

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[7590-01-P]

NUCLEAR REGULATORY COMMISSION 10 CFR PART 52 RIN 3150 - AG23 AP600 Design Certification AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

SUMMARY

The Nuclear Regulatory Commission (NRC or Commission) proposes to amend its regulations to certify the AP600 standard plant design under Subpart B of 10 CFR Part 52.

This action is necessary so that applicants or licensees intending to construct and operate an

- AP600 design may do so by referencing the proposed rule. This proposed design certification rule (DCR), set out as Appendix C, is nearly identical to the two previously codified DCRs in

. Appendices A and B of 10 CFR Part 52. The applicant for certification of the AP600 design is Westinghouse Electric Company LLC (hereinafter referred to as Westinghouse).

The public is invited to submit comments on this proposed DCR and the AP600 design control document (DCD) that is incorporated by reference into the DCR. In addition, interested parties may request an informal hearing before an NRC Atomic Safety and Licensing Board, in accordance with 10 CFR 52.51(b), on matters pertaining to this proposed DCR. The NRC also invites the public to submit comments on the environmental assessment for the AP600 design. l DATE: Submit comments by (Insert date 75 days after publication in the Federal Register). Comments received after this date will be considered if it is practical to consider them, but the Commission is only able to ensure consideration for comments received on or before this date. Requests for an informal hearing must be submitted by (Insert date 75 days after publication in the Federal Register).

l ADDRESSES: Mail written comments and requests for an informal hearing to: Secretary,  !

U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, Mall Stop O-16 C1. Comments may also be delivered to: One White Flint North,11555 Rockville Pike, Rockville, Maryland, between 7:30 am and 4:15 pm on Federal workdays. Copies of comments received, the DCD, and the environmental assessment will be available for examination and copying at the NRC Public Document Room at 2120 L j Street NW. (Lower Level), Washington, DC.

Electronic comments may be provided via the NRC's interactive rulemaking website

' through the NRC home page Iwww.nrc.aovl. From the home page, select "Rulemaking" from the tool bar at the bottom of the page. The interactive rulemaking website can then be accessed by selecting "Rulemaking Forum." This site provides the ability to upload comments as files [any format), if your web browser supports that function. Contact Ms. Carol Gallagher by telephone (301) 415-5905 or e-mail:caa O nrc.aov for information about the interactive rulemaking website.

Attachment 1

FOR FURTHER INFORMATION CONTACT: Jerry N. Wilson, Mail Stop O-12 G15, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or telephone (301) 415-3145, or e-mail: inwonrc.aov.

SUPPLEMENTARY INFORMATION:

TABLE OF CONTENTS

1. Background.

II. Comments and Hearings in the Design Certification Rulemaking.

A. Opportunity to submit written and electronic comments.

B. Opportunity to request hearing.

C. Hearing process.

D. Resolution of issues for the final rulemaking.

E. Access to proprietary information in rulemaking.

F. Ex parte and separation of functions restrictions.

Ill. Section-by-section discussion.

A. Introduction (Section 1).

B. Definitions (Section 11).

C. Scope and contents (Section lil).

D. Additional requirements and restrictions (Section IV).

E. Applicable regulations (Section V).

F. Issue resolution (Section VI).

G. Duration of this appendix (Section Vil).

H. Processes for changes and departures (Section Vill).

l. 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.

Vll. Regulatory Flexibility Act certification.

Vill. Backfit analysis.

IX. Consensus standards.

l. BACKGROUND 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. On June 26,1992, Westinghouse tendered its application for certification of the AP600 standard plant design with the NRC. Westinghouse submitted this application in accordance with Subpart B and Appendix O of 10 CFR Part 52. The NRC formally accepted the application as a docketed application for design certification (Docket No.52-003) on December 31,1992.

Information submitted before that date can be found under Project No. 676.

The NRC staff issued a final safety evaluation report (FSER) related to certification of the AP600 standard plant design in September 1998 (NUREG-1512). The FSER documents the results of the NRC staff's safety review of the AP600 design against the requirements of 10 CFR Part 52, Subpart B, and delineates the scope of the technical details considered in evaluating the design. The FSER provides the bases for Commission approval of the AP600 design through design certification. A copy of the FSER may be obtained from the 2

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Superintendent of Documents, U. S. Govemment Printing Office, P.O. Box 37082, Washington, '

' DC 20402-9328 or the National Technical Information Service, Springfield, VA 22161-0002.

The final design approval for the AP600 design was issued on September 3,1998, and published in the Federal Register on September 11,1998 (63 FR 48772).

Rulemaking Procedures Subpart B of 10 CFR Part 52 provides for Commission approval of standard designs for nuclear power facilities (gg., design certification) through rulemaking. In accordance with the Administrative Procedure Act (APA), Part 52 provides the opportunity for the public to submit written comments on the proposed design certification rule. However, Part 52 goes beyond the requirements of the APA by providing the public with an opportunity to request a hearing before the Atomic Safety and Licensing Board Panciin a design certification rulemaking. While Part 52 describes a general framework for conducting a design certification rulemaking, f 52.51(a) states that more detailed procedures for the conduct of each design certification will be specified by the Commission.

To assist the Commission in developing the detailed rulemaking procedures, the NRC's 1 Office of the General Counsel prepared a paper (SECY-92-381, "Rulemaking Procedures for l

Design Certification," dated November 10,1992), that recommended design certification i rulemaking procedures. This paper was prepared after consideration of the panel discussions at a public workshop and the written comments received after the workshop. On April 30,1993, the Commission issued a Memorandum to the General Counsel that provided the Commission's determinations with respect to the procedural issues raised by the General Counsel's paper.

Section ll, " Comments and Hearings in the Design Certification Rulemaking," describes the l

. procedures to be utilized in this design certification rulemaking. 1 1

II. COMMENTS AND HEARINGS IN THE DESIGN CERTIFICATION RULEMAKING A. Opportunity to Submit Written and Electronic Comments Any person may submit written comments on the proposed design certification rule to the Commission for its consideration.' Commenters have 75 days from the publication of this notice to file written comments on the proposed design certification rule. Commenters needing access to proprietary or safeguards information in order to provide written comments must follow the procedures and filing deadlines (including the date for filing written comments) set forth in Section E below.

Commenters are encouraged to submit, in addition to the original paper copy, a copy of the comment letter in electronic format on a 3.5 inch computer diskette. Text files should be provided in Wordperfect 8 format or unformatted ASCll code. The format and version should be identified on the diskette's extemal label.

B. Opportunity to Request Hearing Any person may request an informalhearing on one or more specific matters with respect to the proposed design certification rule. An informal hearing provides the admitted party with an opportunity to provide written and oral presentations on those matters to an 2

An opportunity for public comment is required by Section 553 of the Administrative Procedures Act and 10 CFR 52.51(b).

rAn opportunity for 2. hearing is provided by 10 CFR 52.51(b).

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. i Atomic Safety and Licensing Board, and to request that the licensing board c;uestion the applicant on those matters. The conduct of an informal hearing is discussed in more detail in  !

Section C. below. Under certain circumstances, a party in an informal hearing may request that the Commission hold a formal hearing on specific and substantial factual disputes necessary to resolve the matters for which the party was granted an informal hearing (Section C.11 below).

A person may request an informal hearing even though that person has not submitted separate written comments on the design certification rule (.ia., is not a commenter). Requests for an informal hearing must be received by the Commission no later than 75 days from the publication of this notice, and a copy of the request must be sent via ovemight mail to the design certification applicant at the following address: Mr. Brian A. McIntyre, Manager, Advanced Plant Safety and Licensing, Westinghouse Electric Company, P.O. Box 355, Pittsburgh, PA 15230-0355. The information which a person requesting a hearing must provide in the hearing request, as well as the procedures and standards to be used by the Commission in its determination of the request, are discussed in Sections C.1 through C.4 below.

A person who needs to review proprietary information submitted by the design certification applicant in order to prepare a request for an informal hearing must follow the procedures and filing schedule set forth in Section E. below.

The Commission is also providing an opportunity for interested State, county, and city / municipal and other local Governments, as well as Native American tribal governments, to l participate as " interested governments" in any informal hearings which the Commission I authorizes, similar to their participation as " interested govemments" in Subpart G hearings under 10 CFR 2.715. State, county, city / municipal, local, and tribal Govemments wishing to j

participate as an " interested govemment" in any design certification rulemaking hearings must i file their request to participate no later than 75 days from the publication of this notice.

C. Hearing Process

1. Filings and Computation of Times All notices, papers, or other filings discussed in this section must be filed by express mail.8 The time periods specified in this section have been established based upon such a filing. The express mail filing requirement shall be considered in establishing other filing deadlines.

In computing any period of time, the day of the act, event, or default after which the designated period of time begins to run is not included. The last day of the period so computed is included, unless it is a Saturday, Sunday, or legal holiday at the place where the action or event is to occur, in which case the period runs until the next day which is neither a Saturday, Sunday, nor holiday.

2. Content of Hearing Request 3

Filings discussed in this section may also be served upon the Commission in electronic form in lieu of express mail. However, parties must serve copies of their filings on other parties by express mail, unless the receiving party agrees to filing in electronic form. These filings must be transmitted no later than the last day of the time period specified for filing and must be in accordance with the requirements specified under Date and Addresses in this notice.

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4 The Commission will grant a request for an informal hearing only if the hearing request satisfies each of the following two requirements. First, the hearing request must include the written presentations that the requestor wishes to be included in the record of the hearing. The written presentations must:

(i) Identify the specific portion of the proposed design certification rule or supporting bases which are challengod, l j

(ii) Describe the reasons why the proposed rule or supporting bases are incorrect or I insufficient, and f

(iii) Identify the references or sources upon which the person requesting the hearing relies.

If the requestor has submitted written comments in the public comment period addressing tnese three factors for the specific issue for which the requestor seeks a hearing, it will be sufficient for the requestor to identify the portions of the written comme % that the requestor intends to submit as a written presentation. Also, the hearing requem must 3

demonstrate that the requestor (or other persons identified in the hearing request who will represent, assist, or speak on behalf of the requestor at the hearing) has appropriate knowledge and qualifications to enable the requestor to contribute significantly to the development of the hearing record on the specific matters at issue. The Commission does not intend that the requestor meet a judicial " expert witness" standard in order to meet the second criterion. Nonetheless, given the substantial commitment of time and resources associated with any hearirig, the Commission believes it to be a reasonable prerequisite that the requestor demonstrate that he/she (or his/her assistant) has:

(i) Substantial familiarity with the publicly available docketed information relevant to the issue for which a hearing is requested; (ii) The requisite technical capability to understand the factual matters and develop a record on the issue for which a hearing is requested, and (iii) An understanding of the NRC's hearing procedures in 10 CFR Part 2.'

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3. Request to Hold Hearing Outside of Washington, DC Any hearing (s) which the Commission may authorize ordinarily will be conducted in the Washington, DC, metropolitan area. However, the Commission at its discretion may schedule hearings outside the Washington, DC. metropolitan area in response to requests submitted by  ;

a person requesting a hearing that all or part of the hearing be held elsewhere. These requests must be submitted in conjunction with the request for hearing, and must specifically explain the special circumstances for holding a hearing outside the Washington, DC. metropolitan area.

4. Responses to Hearing Request The applicant may file a response to any hearing request within 15 days of the date of the hearing request. The NRC staff will not provide a response to the hearing request unless requested to do so by the Commission but may assist the Commission in its ruling on the request.

5. Commission Determination of Hearing Request The Commission intends to rule on a hearing request within 20 days of the close of the

' Requesters will satisfy this requirement by stating that they possess and have read a copy of 10 CFR Part 2, Subparts A. G. and L.

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h.

. period for requesting a hearing. The Commission's determination will be based upon the

- materials accompanying the hearing request and the applicant's response (and the NRC staff's response, if requested by the Commission). The hearing request shall be granted if:

(i) The request is accompanied by a written presentation containing the information required by Section C.2. above; and 1 (ii) The requestor has the appropriate knowledge and qualifications to enable the I requestor to contribute significantly to the development of the hearing record on the matters sought to be controverted.

' The Commission may consult with the NRC staff before its determination of a hearing request. A written decision either granting or denying the hearing request will be published by the Commission.

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If a hearing request is granted in whole or in part, the Commission's decision will 4 delineate the controverted matter that will be the subject of the hearing and whether any issues and/or parties are to be consolidated (agg Section C.7. below). The Commission's decision

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granting the hearing will direct the establishment of a licensing board to preside over the informal hearing. Finally, the Commission's decision will specify:

(i) The date by which any requests for discovery must be filed with the licensing board (normally 20 days after the date of the Commission's decision), and (ii) The date by which any objections to discovery must be filed (Eng Section C.9.

below). ..

The Commission's decision will be sent to each admitted party by overnight mail.

Separate hearings may be granted for each controverted matter or set of consolidated matters.

Thus, if there are three different controverted matters, the Commission may establish three separate hearings. In this fashion, closing of the hearing record on a controverted matter and its referral to the Commission for resolution need not await completion of the hearing on the ,

other controverted matters. Finally, the Commission's decision will rule on any requests for  !

hearings outside of the Washington, DC. metropolitan area (agg Section C.3 above).

6. Authority of the Licensing Board If the Commission authorizes an informal hearing on a controverted matter, the licensing  ;

board will function as a " limited magistrate" in that hearing with the authority and responsibility '

for assuring that a sufficient record is developed on those controverted matters which the Commission has determined are appropriate for consideration in that hearing. The licensing board shall have the following specific responsibilities and authority:

- (i) Schedule and expeditiously conduct the informal hearing for each admitted controverted matter, consistent with the rights of all the parties and with the Commission's Statement of Policy on Conduct of Adjudicatory Proceedings5 , CLl-98-12,48 NRC 18 (1998),

(63 FR 41872, August 5,1998),

(ii) Review all discovery requests against the criteria established by the Commission, and refer all appropriate requests to the Commission with a decision explaining the licensing board's action, (iii) Preside over and resolve any issues regarding the scheduling and conduct of any 5

Although the opportunity for an informal hearing provided for in Section 52.51(b) and this rulemaking notice is aqt an adjudicatory hearing p_qr sg. the underlying principals and goal of expeditious and fair conduct of adjudicatory hearings are also applicable to informal hearings.

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l discovery authorized by the Commission, (iv) Order such further consolidation of parties and issues as the licensing board determines is necessary or desirable, (v) Orally examine persons making oral presentations in the informal hearing, based in part upon the licensing board's review of the parties' proposed oral questions to be asked of persons making oral presentations, (vi) Request that the NRC staff:

(A) Answer licensing board questions about the FSER or the proposed rule, (B) Provide additional information or documentation with respect to the design certification, and (C) Pr> vide other assistance as the licensing board may request. Licensing board requests for NRC staff assistance should be framed such that the NRC staff does not assume a role as an adversary party in the informal hearing (agg Section C.8 below),

(vi!) Review all requests for additional hearing procedures and refer all appropriate requests to the Commission with a decision explaining the licensing board's action, (viii) Certify the hearing record to the Commission, based upon the licensing board's determination that the hearing record contains sufficient information for the Commission to make a reasoned determination on the controverted matter; and (ix) Include with its certification any concems identified by the licensing board in the course of the hearing which, although neither raised by the parties nor necessary to resolution of the controverted hearing matters, are significant enough in the licensing board's view to warrant attention by the Commission.

Licensing board determinations with respect to referral of requests to the Commission, as well as licensing board determinations of parties' motions, are not appealable to the Commission as an interlocutory matter. Instead, any disagreements with the licensing board's determinations and a specific discussion of how the hearing record is deficient with respect to the contested issue must be set forth in the parties' proposed findings of fact which are submitted directly to the Commission (agg Section C.13 below).

As suggested by item (ix) above, the licensing board shall not have any "sua sponte" authority analogous to 10 CFR 2.760a. The Commission believes that in the absence of a l request for an informal hearing on a matter, the Commission should resolve issues with respect  !

to the design certification rule in the same manner as other agency-identified rulemaking i issues, yig., through NRC staff consideration of the issue followed by the Commission's review '

and its final resolution of the matter. However, when it certifies the completed hearing record to the Commission (agg Section C.12. below), the licensing board should identify to the Commission any concerns identified during the hearing that are significant enough to warrant Commission consideration but that are unnecessary or irrelevant to the resolution of the controverted hearing matter.

The licensing board shall close the hearing and certify the record to the Commission only after it determines that the record on the controverted matter is sufficiently complete for the Commission to make a reasoned determination with respect to that matter. However, the licensing board shall not have any responsibility or authority to resolve and decide controverted matters in either an informal or a formal hearing. Rather, the Commission retains its traditional authority in rulemaking proceedings to evaluate and resolve all rulemaking issues identified in public comments on a proposed rule. Therefore, the Commission will resolve any controverted matters that are the subject of a hearing in this design certification rulemaking.

7. Consolidation of Parties and issues; Joint Hearings on Related Issues 7

If two or more persons seek an informal hearing on the same or similar matters, the Commission may, in its discretion, grant an informal hearing and consolidate the matters into a single issue (as defined by the Commission). The Commission may also, in its discretion, require that the parties be consolidated analogous to the consolidation permitted under 10 CFR 2.715a. If the Commission consolidates two or more issues into a single consolidated issue but does not consolidate parties, each admitted person will be deemed a separate party with an individual right to:

(i) Submit separate written presentations, (ii) Submit separate sets of proposed oral questions to be asked by the licensing board (agg Section C.10 below),

(iii) Make separate oral presentation, and (iv) Submit and separately respond to motions.

If the Commission also requires that parties be consolidated, the consolidated parties must participate jointly, including deciding upon written and oral presentations, submitting a single set of written questions, submitting motions supported by each of the consolidated parties, and responding to motions filed by other parties.

During the informal hearing, the licensing board may decide that further consolidation of issues or parties would simplify the overall conduct of informal hearings or materially reduce the time or resources devoted to the hearings. In these instances, the licensing board may direct such consolidation. The licensing board shall set forth the issues and/or parties to be consolidated and the reasons for such consolidation in a written order.

8. Statu's of the Design Certification Applicant, the NRC staff, and Requesting Party The design certification applicant shall be a party in the informal hearing, with the right to submit written and oral presentations, propose questions to be asked by the licensing board of oral presenters, and file and submit appropriate motions.

The NRC staff shall not be a party in the informal hearing but shall be available in the informal hearing to answer licensing board questions about the FSER or the proposed rule, provide additional information or documentation with respect to the design certification, and provide other assistance that the licensing board may request without the NRC staff assuming the role of a party in the informal hearing.

A party whose hearing requests have been granted with respect to a particular controverted matter shall not participate with respect to any controverted matter on which the )

party was not granted a hearing. For example, if Person 1 has been authorized as a party on issue A and Person 2 has been authorized as a party on issue B, then Person 1 may participate only in the informal hearing on Issue A, and may not participate in the informal hearing on issue B. Conversely, Person 2 may participate only in the informal hearing on issue B, and may not participate in the informal hearing on issue A.

9. Requests for Discovery Any party may request the opportunity to conduct discovery against another party before the oral phase of the informal hearing. The request for discovery must: .

(i) Identify the type of discovery permitted under 10 CFR ff 2.740,2.740a,2.740a(b),

2.741, and 2.742 which the party seeks to use; (ii) Identify the subject matter or nature of the information sought to be obtained by discovery; and (iii) Explain with particularitythe relevance of the information sought to the controverted matter which is the subject of the hearing and why this information is indispensable to the l 8

presentation of the party's position on the controverted matter.

l The request shall be filed with the licensing board, with copies of the request to be filed with the

l. party against which discovery is sought, and the NRC staff. The requests must be received no l later than the deadline specified by the Commission in its decision granting a party's hearing request (agg Section C.5. above). A party against whom discovery is sought may file a response objecting to part or all of the request. Such a response must explain with particularity

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l why the discovery request should not be granted.

The licensing board shall review all discovery requests and refer to the Commission l those requests that it believes should be granted within 7 days after the date for receiving a party's objections to a discovery request. The licensing board shall issue a written decision explaining its basis for either referring the request to the Commission or declining to refer it.

The written decision shall accompany the discovery requests which are referred by the licensing ]

board to the Commission. .

,l

- The Commission will determine whether to grant any discovery requests forwarded to it based upon the licensing board's decision, together with the request and the design certification applicant's response (and any NRC staff response requested by the licensing board). j Discovery will be at the discretion of the Commission. In this regard, the Commission notes that there are two docket files in which the NRC staff has placed information and documents received from the applicant for the AP600 design certification review. The application was L

' docketed on December 31,1992 and assigned Docket No.52-003. Correspondence relating to the application prior to this date was addressed to Project No. 676. This information includes the AP600 Design Control Document, Revision 2 (3/99) and the AP600 Standard Safety Analysis Report, Revision 25. Furthermore, the docket files contain NRC staff communications and documents, such as written questions and comments provided to the design certification applicant, and summaries of meetings held between the NRC staff and the design certification applicant. The NRC staff's bases for approving the AP600 design are set forth in the FSER (NUREG-1512), dated September 1998. The Commission also notes that each admitted party has already disclosed a substantial amount of information in its hearing request, relating both to bases for the party's position with respect to the controverted matter as well as information on l the qualifications of the party (or its representatives and_ witnesses in the hearing).

! As discussed above, much of the information documenting the NRC staff's review and i approval of the design certification application has been routinely placed in the docket file.

I Furthermore, as discussed above in Section C.8, the NRC staff is not a party in an informal hearing. Therefore, the Commission has decided that in an informal hearing, the parties should not be afforded discovery against the NRC staff.

10. Conduct of informal Hearing l if the Commission authorizes discovery, the licensing board shall establish a schedule  !

l for the conduct and completion of discovery. Normally, the licensing board should not permit more than one round of discovery. The Commission will not entertain any interlocutory appeals

- from licensing board orders resolving any discovery disputes or otherwise complaining of the scheduling of discovery. ,

Following the completion of discovery, the licensing board should issue an order setting l forth the date of commencement of the oral phase of each informal hearing, and the date (no l

l less than 30 days before the commencement of the oral phase of the hearing) by which parties j l must submit:

L (i) The identities and curriculum vitae of those persons presiding oral presentations; (ii) The outlines of the oral presentations; and 9

(iii) Any questions which a party would like the licensing board to ask.

The licensing board may schedule the oral phases of two or more informal hearings to be held during the same session. The licensing board shall publish a notice in the Federal Register announcing the commencement of the oral phase of the informal hearing (s). The notice shall set forth the place and time of the oral hearing session, the subject matter (s) of the informal hearing (s), a brief description of the informal hearing procedures, and a statement j indicating that the public may observe the informal hearing.  !

Based upon the parties' outlines of the oral presentations and proposed questions, the l licensing board should determine whether it has specific questions of the NRC staff with respect to the staff's review of the design certification application. These questions should be submitted in writing to the NRC staff no less than 20 days before the commencement of the oral phase of the hearing and must specify the date by which the NRC staff shall provide its written answers to the licensing board. The licensing board shall send copies of the request by ovemight mail to all parties. The NRC staff shall file its written answers with the licensing board and the parties.

During the oral phase of the hearing, the licensing board shall receive into evidence the written presentations of the parties and permit each party (or the representatives identified in their hearing request) to make oral presentations addressing the controverted matter.

Normally, the party raising the controverted matter should make their presentations, followed by i the presentations of the design certification applicant. The licensing board may question the j persons making oral presentations, using its own questions as well as those submitted to the  !

licensing board by the other parties. Based upon the parties' oral presentations and/or i responses to licensing board questions, the licensing board also may orally question the NRC staff. I

11. Additional Hearing Procedures and Formal Hearings After the parties have made their oral presentations and the licensing board has concluded its questioning of the presenters (and, as applicable, the NRC staff), the licensing board should declare that the oral phase of an informal hearing on a controverted matter (or consolidated set of controverted matters) is complete.

No later than 10 days after the licensing board has declared that the oral phase of the informal hearing has been completed, parties may file with the licensing board (with copies to the applicant and the NRC staff) a request that some or all of the procedures described in 10 CFR Part 2, Subpart G (gg., direct and cross-examination by the parties) be utilized. The request shall:

(i) Identify the specific hearing procedures which the party seeks, or state that a formal hearing is requested; (ii) Identify the specific factualissues for which the additional procedures would be utilized; (iii) Explain why resolution of these factual disputes are necessary to the Commission's decision on the controverted issue; (iv) Explain, with specific citations to the hearing record, why the record is insufficient on the controverted matter; and (v) Identify the nature of the evidence that would be developed utilizing the additional procedures requested.

The design certification applicant may file a response to these requests no later than 7 days after the applicant's receipt of a request for additional procedures. The NRC staff wn! not provide a response unless specifically requested to do so by the licensing board.

10

The licensing board will review all requests for additional hearing procedures or a formal hearing and refer those that it believes should be granted to the Commission for its determination. The licensing board shall issue a written decision explaining its determination whether to forward the request to the Commission no later than 7 days after receipt of any applicant response to the request. The decision will provide the basis for either forwarding the request to the Commission or declining to forward it. In the absence of any requests for hearing procedures or if the licensing board concludes that none of the requests should be referred to the Commission, the licensing board should declare that the hearing record is closed (agg Section C.12 below).

The Commission will determine whether to grant any requests for additional procedures or a formal hearing that are forwarded by the licensing board. The Commission's determination shall be based upon the licensing board's decision along with the request and the design certification applicant's response. If the Commission directs that a formal hearing be held on a controverted factual matter, the NRC staff shall be a party in the formal hearing. Any formal hearing authorized by the Commission shall be conducted in accordance with the Commission's Statement of Policy on Conduct of Adjudicatory Proceedings. As noted in that Policy Statement, the Commission may, in individual cases, establish specific milestone schedules for the conduct of the formal hearing and require the presiding officer to explain and mitigate any significant deviations from that milestone schedule. After either the additional hearing procedures authorized by the Commission are completed or the formal hearing is concluded on the factual dispute, the licensing board should declare the hearing record closed (agg Section C.12 below).

12. Licensing Board's Certification of Hearing Record to the Commission After the oral phase of a hearing is completed and either:

(i) There are no requests for additional hearing procedures or a formal hearing or (ii) The licensing board concludes that none of the requests should be referred w the Commission, then the licensing board should declare that the hearing record is closed.

If the Commission directs that additional hearing procedures should be utilized or a f armal hearing be held on specific factual disputes, the licensing board should declare the hearing record closed after completion of the additional hearing procedures or the formal hearing.

Within 30 days of the closing of the hearing record the licensing board should certify the hearing record to the Commission on each controverted matter (or consolidated set of controverted matters).8 The licensing board's certification for each controverted matter (or consolidated set of controverted matters) shall contain:

(i) The hearing record, including a transcript of the oral phase of the hearing (and any pre-hearing conferences) and copies of all filings by the parties and the licensing board, (ii) A list of all documentary evidence admitted by the licensing board, including the written presentations of the parties, (iii) Copies of the documentary evidence admitted by the licensing board, (iv) A list of all witnesses who provided oral testimony, (v) The NRC staff's written answers to licensing board requests, and (vi) A licensing board statement that the hearing record contains sufficient information 6

An informal hearing is deemed to be completed when the period for requesting additional procedures or a formal hearing expires and no request is received.

11

for the Commission to make a reasoned determination on the controverted matter.

Finally, as discussed in Section C.6 above, the licensing board should identify any issues not raised by the parties or otherwise are not relevant to the controverted matters in the hearing, that the licensing board believes are significant enough to warrant attention by the

' Commission.

13. Parties' Proposed Findings of Fact and Conclusions The applicant must file directly with the Commission proposed findings of fact and conclusions for each controverted hearing matter (or consolidated set of controverted matters) within 30 days following the close of the hearing record c: that matter in the form of a proposed final rule and statement of considerations with respect to the controverted hearing issucs.

Other parties are encouraged, but not required, to file with the Commission proposed findings of fact and conclusions limited to those issues which a party was afforded a hearing by the Commission (ia., i a party may not file proposed findings of fact and conclusions on issues which it was notadmitted). Any findings that a party wishes the Commission to consider must be received by the Commission no later than 30 days after the licensing board closes the hearing record on that issue. Although parties are not required to file proposed findings and conclusions, a party who does not file a finding may not, upon appeal, claim or otherwise argue that the Commission either misunderstood the party's position, or failed to address a specific piece of evidence or issue.

D. Resolution of lasues for the Final Rulemaking

1. Absence of Qualifying Hearing Request if the Commission does not receive any request for hearing within the 75-day period for submitting a request, or does not grant any of the requests (see Section B. above), the Commission will determine whether the proposed design certification rule meets the applicable standards and requirements of the Atomic Energy Act of 1954, as amended (AEA), the National Environmental Policy Act of 1969, as amended (NEPA), and the Commission's rules and regulations. The Commission's determination will be based upon the rulemaking record, which includes: the application for design certification, including the AP600 Standard Safety Analysis Report (SSAR) and DCD; the applicant's responses to the NRC staff's requests for additional information; the NRC staff's FSER and any supplements thereto; the report on the application by the ACRS; the applicant's evaluation of severe accident mitigation design altematives for purposes of NEPA in Appendix 18 of the SSAR the NRC staff's draft EA and FONSl; the proposed rule, and the public comments received on the proposed rule. If the Commission makes an affirmative finding, it will issue a standard design certification in the form of a rule by adding a new appendix to 10 CFR Part 52, and publish the design certification rule and a statement of considerations in the Federal Register.

2. Commission Resolution of issues Where a Hearing is Granted All matters related to the proposed design certification rule, including those matters for which the Commission authorizes a hearing (agg Sections B. and C. abave), will be resolved by the Commission after the licensing board has closed the hearing record and certified it to the Commission. The Commission will determine whether the proposed design certification rule meets the applicable standards and requirements of the AEA, NEPA, and the Commission's rules and regulations. The Commission's determination will be based upon the rulemaking record as described in Section D.1 above, with the addition of the hearing record for controverted matters. If the Commission makes an affirmative finding, the Commission will 12 l

E 1 lasue a final design certification rule as described in Section D.1.

l E. Access to Proprietary Information in Rulemaking l

1. Access to Proprietary information for the Preparation of Written l Comments or Informal Hearing Requests Persons who determine that they need to review proprietary information submitted by the design certification applicant to the NRC in order to submit written comments on the l

proposed certification or to prepare an informal hearing request, may request access to such l information from the applicant.

The request shall state with particularity:-

(i) The nature of the proprietary information sought, (ii) The reason why the nonproprietary information currently available to the public in the NRC's Public Document Room is insufficient either to develop public comments or to prepare for the hearing, (iii) The relevance of the requested information either to the issue which the commenter wishes to comment on, and (iv) A showing that the person requesting the information has the capability to understand and utilize the requested information.

Requests must be filed with the applicant such that they are received by the applicant no later than 45 days after the date that this notice of proposed rulemaking is published in the 4 Federal Register.

Within ten (10) days of receiving the request, the applicant must send a written response to the person seeking access. The response must either provide the documents requested (or state that the document will be provided no later than ten days after the date of the response), or state that access has been denied if access is denied, the response shall state with particularity the reasons for its refusal. The applicant's response must be provided via express mail.

The person seeking access may then request a Commission hearing for the purpose of obtaining a Commission order directing the design certification applicant to disclose the requested information. The person must include copies of the original request (and any subsequent clarifying information provided by the person requesting access to the applicant) and the applicant's response. The Commission will base its decision solely on the person's original request (including any clarifying information provided to the applicant by the person requesting access), and the applicant's response. Accordingly, a person seeking access to proprietary information should ensure that the request sets forth in sufficient detail and particularity the information required to be included in the request. Similarly, the applicant should ensure that its response to any request states with sufficient detail and particularity the reasons for its refusal to provide the requested information.

If the Commission orders access in whole or part, the Commission will specify the date by which the requesting party must file with the Commission written comments and any request for an informal hearing before a licensing board as discussed in Section V.C. above. A request for an informal hearing must meet the requirements set forth above in Section V.C., in particular the requirements goveming the content of the hearing request, and shall be govemed by the procedures and standards goveming such requests set forth in Section V.C.

2. Access to Proprietary information in a Hearing Parties wno are granted a hearing may request access to proprietary information.

Parties must first request access to proprietary information regarding the proposed design 13

. certification from the applicant. The request shall state with particularity; (i) The nature of the proprietary information sought, (ii) The reason why the nonproprietary information currently available to the public in the NRC's Public Document Room is insufficient to prepare for the hearing, (iii) The relevance of the requested information to the hearing issue (s) for which the party has been admitted, and (iv) A showing that the requesting party has the capability to understand and utilize the requested information.

The request must be filed with the applicant no later than the date established by the Commission for filing discovery requests with the licensing board.

If the applicant declines to provide the information sought, within 10 days of receiving the request, the applicant must send a written response to the requesting party setting forth with particularity the reasons for its refusal. .The party may then request the licensing board to order disclosure. The party must 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 licensing board shall base its decision solelyon the party's original request (including any clarifying information provided by the requesting party to the applicant), and the applicant's response.

Accordingly, a party requesting proprietary information from the applicant should ensure that its request sets forth in sufficient detail and particularity the information required to be included in the request. Similarly, the applicant should ensure that its response to any request states with sufficient detail and particularity the reasons for its refusal to provide the requested information. The licensing board may order the applicant to provide access to some or all of the requested information, subject to an appropriate non-disclosure agreement.

F. Ex Parte and Separation of Functions Restrictions Unless the formal procedures of 10 CFR Part 2, Subpart G are approved for a formal hearing in the design certification rulemaking proceeding, the NRC staff will not be a party in the hearing and separation of functions limitations will not apply. The NRC staff may assist in the hearing by answering questions about the FSER put to it by the licensing board, or to provide additional information, documentation, or other assistance as the licensing board may request. Furthermore, other than in a formal hearing, the NRC staff shall not be subject to discovery by any party, whether by way of interrogatory, deposition, or request for production of documents.

Second, the Commission has determined that once a request for an informal or formal hearing is received, certain elements of the exparte restrictions in 10 CFR 2.780(a) will be ,

applicable with respect to the subject matter of that hearing request. Under these restrictions,  !

the Commission will communicate with interested persons / parties, the NRC staff, and the licensing board with respect to the issues covered by the hearing request only through docketed, publicly-available written communications and public meetings. Individual Commissioners may communicate privately with interested persons and the NRC staff; however, the substance of the communication shall be memorialized in a document which will be placed in the PDR and distributed to the licensing board and relevant parties.

Ill. SECTION-BY-SECTION DISCUSSION OF DESIGN CERTIFICATION RULE The proposed design certification rule (DCR) for the AP600 standard plant design is nearly identical to the two design certification rules for the U.S. ABWR and the System 80+

designs, which the NRC previously adopted. These DCRs are set forth in 10 CFR Part 52, 14

Appendix A (U.S. ABWR,62 FR 25800, May 12,1997) and Appendix B (System 80+,62 FR

'_27840, May 21,1997). The AP600 DCR emulates the U.S. ABWR and System 80+ DCRs, inasmuch as the three designs were reviewed contemporaneously against the same technical

~

requirements. Furthermore, many of the proceduralissues and their resolutions for the ABWR and the System 80+ DCRs (Lg., the two-tier structure, Tier 2*, the scope of issue resolution) were developed after extensive discussions with nuclear industry representatives, and Westinghouse participated in those discussions. It was the NRC's intent (and likely Westinghouse's expectation) that the resolutions for these issues in the ABWR and System 80+ rulemakings would also be applied to the AP600 design. Accordingly, the NRC has <

modeled the AP600 DCR on the existing DCRs for the ABWR and Svstem 80+, with certain departures. These departures are necessary to reflect that Westinghouse is the applicant for the AP600 DCR, and to account for differences in the AP600 design documentation, design features (including the investment protection short-term availability controls), and environmental:

assessment (including severe accident mitigation design alternatives).

The following discussion sets forth the purpose and key aspects of each section and paragraph of the proposed AP600 design certification rule. All section and paragraph references are to the provisions in the proposed Appendix C to 10 CFR Part 52.

A. Introduction. j The purpose of Section I of Appendix C 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 cedification applicant to provide this information, I then the COL applicant must meet the 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 representatives of the nuclear industry afte>r issuance of 10 CFR Part 52. During consideration of the comments received on Appendices A and B to Part 52, the Commission determined that it would be useful to distinguish between the Tiant-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 distincQn is necessary in order to clarify the obligations of applicants and licensees that reference this appendix. Also, the technical specifications that are located in Section 16.1 of the generic DCD are designated as " generic technical specifications" in order to facilitate the special treatment of this information under this appendix. Therefore, appropriate definitions for these additional terms are included in this appendix.

The Tier 1 portion of the ' design-related information contained in the DCD is certifiedby this appendix and, therefore, subject to the special backfit provisions in Vill.A of this appendix.

An applicant who references this appendix is required to incorporate by reference and comply with Tier 1, under lil.B and IV.A.1 of this appendix. This information consists of an introduction 15

r.

]

l to Tier 1, the system based and non-system based design descriptions and corresponding inspections, tests, analyses, and acceptance criteria (ITAAC), significant interface i

requirements, and significant site parameters for the design. The design descriptions, interface l 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 is provided in Section 14.3 of the FSER. Changes to or departures from the Tier 1 information i must comply with Section Vill.A of this appendix.

The Tier 1 design descriptions serve as design commitments for the lifetime of a facility l

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-l specific DCD unless changes are made in accordance with the change process in Section Vill

- 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 this appendix. 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 Ill.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 Vill.B of this appendix. Tier 2 includes the information required by 10 CFR 52.47 (with the exception of generic technical specifications, conceptual design information, and the evaluation of severe accident mitigation design attematives) and the supporting information on inspections, tests, and analyses that will be performed to demonstrate that the acceptance criteria in the ITAAC have been met. As with Tier 1, Ill.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 I Tier 2, except for the COL action items, including the investment protection short-term '

availability controls in Section 16.3 of the generic DCD. The definition of Tier 2 makes clear that Tier 2 information has been determined by the Commission, by virtue of its inclusion in this j appendix and its designation as Tier 2 information, to be an approved (" sufficient") method for l 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 specified in Section l Vill.B of this appendix. Departures from Tier 2 do not negate the requirement in Section Ill.B to '

reference Tier 2.

A definition of " combined licenea (COL) action items" (combined license information),

which is part of the Tier 2 information, has been added to clarify that COL applicants, who reference this appendix, 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 combined license, these items are not requirements for the licensee unless such items are restated in its FSAR.

The investment protection short-term availability controls, which are set forth in Section 16.3 of the generic DCD, were added to the list of information that is part of Tier 2. This set of requirements was added to Tier 2 to make it clear that the availability controls are not

(

16 L

1 operational requirements for the purposes of Vill.C of this appendix. Rather, the availability controls are associated with specific design features, and the availability controls may be changed if the associated design feature is changed under Vill.B of this appendix.

. 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 Vill.B, 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 Vill.B.6 of this appendix. Some Tier 2* requirements, conceming -

special preoperational tests, are designated to be performed only for the first plant or first three plants referencing the AP600 DCR. The Tier 2* designation for these selected tests will expire after the first plant or first three plants complete the specified tests. However, a COL action item requires that subsequent plants shall also perform the tests or justify that the results of the first-plant-only or first-three-plants-only tests are applicable to the subsequent plant. The Commission is interested in comments addressing whether the first-plant-only or first-three-plants-only limitations should be part of the Tier 2* information for these specified tests.

During development of Appendices A and B to Part 52, the Commission decided that l

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 appendix. 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 Vill 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 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 Vlli of this appendix. Thus, the plant-specific DCD would function akin to an updated Final Safety Analysis Report, in the sense 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. Therefore, this appendix defines both a I generic DCD and plant-specific DCD. Also, the Commission decided to treat the technical

{

specifications in Section 16.1 of the generic DCD as a special category of information and to '

designate them as generic technical specifications. A COL applicant must submit plant-specific technical specifications that consist of the generic technical specifications, which may be modified under Vill.C of this appendix, and the remaining plant-specific information needed to complete the technical specifications, including bracketed values. The Final Safety Analysis Report (FSAR) that is required by 52.79(b) will consist of the plant-specific DCD, the site-specific portion of the FSAR, and the plant-specific technical specifications.

C. Scope and contents.

L The purpose of Section lli 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 17

l

'l

. 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, 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 generic DCD available upon request after the final rule becomes effective. Therefore, lil.A of this appendix identifies a representative of Westinghouse who can be contacted to obtain a copy of the generic DCD.

Paragraphs A and B also identify the investment protection short-term availability controls in Section 16.3 of the generic DCD as part of the Tier 2 information. During its review of the AP600 design, the NRC determined that residual uncertainties associated with passive safety system performance increased the importance of non-safety-related active systems in providing defense-in-depth functions that back-up the passive systems. As a result, Westinghouse developed some administrative controls to provide a high level of confidence that -

l active systems having a significant safety role are available when challenged. Westinghouse '

named these additional controls " investment protection short-term availability controls," and the Commission included this statement in Section ill to ensure that these availability controls are binding on applicants and licensees that reference this appendix and will be enforceable by the NRC. The NRC's evaluation of the availability controls is provided in Chapter 22 of the FSER.

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 generic DCD will also be available at the NRC's Public Document Room. Questions conceming the accuracy of information in an application that references this appendix will be resolved by checking the master copy of 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 Vill 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 Westinghouse's evaluation of severe accident mitigation design attematives 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. Section 1.8 of Tier 2 identifies the location of the conceptual design information.

18

Westinghouse's evaluation of various design alternatives to prevent and mitigate severe accidents does not constitute design requirements. The Commission's assessment of this information is discussed in Section IV of this SOC on environmentalimpacts. 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 generic DCD, as requested by NEl and the applicant for design certification. The NRC agreed with the request 4

. to delete this information because conformance with the deleted portions of the PRA is not l necessary. Also, the NRC's position is predicated in part upon NEl's acceptance, in conceptual i form, of a future generic rulemaking that will require a COL applicant or licensee to have a l plant-specific PRA that updates and supersedes the design-specific PRA supporting this rulemaking and maintain it throughout the operational life of the facility. j 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 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 AP600 design (AP600 Standard Safety Analysis Report) or the final safety evaluation report for the certified standard design.

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 administration building. Because this statement is not a definition, the Commission decided that the appropriate location is in Section lll of this appendix.

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 includedin 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. i 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 referenta o 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 and 52.79. 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 19

confusion. 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.R.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 Section 16.1 of the DCD, with any changes made under Section Vill.C of this appendix, and the technical specifications for the site-specific portions of the plant that are either partially or wholly outsido the scope of this design certification. 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 Vill 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 Combined License Information, in the application. The Combined License information identifies matters that need to be addressed by an applicant that references this appendix, as required by i 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 l within the scope of this rule, such as generic issues that must be addressed, in whole or in part, 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 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 10 CFR Part 50 construction permit or operating license. This provision is necessary because the previous design certifications were not implemented in the manner that was originally envisioned at the time that 10 CFR Part 52 was created. The Commission's concem is with the manner in which ITAAC were developed and the lack of experience with design certifications in license proceedings. Therefore, it is appropriate to have some uncertainty regarding the manner in which this appendix could be referenced in a 10 CFR Part 50 licensing proceeding.

E. ' Applicable regulations.

The purpose of Section V of this appendix is to specify the regulations that will be applicable and in effect (if and) when this proposed design certification is approved. These regulations will consist of the technically relevant regulations identified in paragraph A, except for the regulations in paragraph B that will not applicable to this certified design.

Paragraph A will identify the regulations in 10 CFR Parts 20,50,73, and 100 that are applicable to the AP600 design. The Commission's determination of the applicable regulations will be made as of the date specified in paragraph V.A of this appendix, which will be the date 20

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that this appendix is approved by the Commission and signed by the Secretary.

In paragraph V.B of this appendix, the Commission identified the regulations that do not apply to the AP600 design. The Commission has determined that the AP600 design should be exempt from portions of 10 CFR 50.34,50.62, and Appendix A to Part 50, as described in the l

FSER (NUREG-1512) and summarized below: 1 (1) Paragraph (a)(1) of 10 CFR 50.34 - whole body dose criterion. '

This regulation sets forth dose criteria to be used in siting determinations. The NRC y staff performed its evaluation of the radiological consequences of postulated design basis accidents for the AP600 design against the dose criterion specified in 10 CFR 50.34(a)(1)(ii)(D) because it was the Commission's intent that the new dose criterion be used for future nuclear power plants. However, when the NRC codified the new reactor site criteria for nuclear power plants (61 FR 65157; December 11,1996), it made an error in the assignment of applicants that could use the new dose criterion [25 rem TEDE), versus those that must use the whole body criterion. The assignment of applicants in 10 CFR 50.34(a)(1), who must use the whole body <

criterion, should not have included applicants for a design certification or combined license who  !

applied prior to January 10,1997 (refer to 61 FR 65158). The Commission adopted 25 rem TEDE as the new dose criterion for future plant evaluation purposes, because this value is essentially the same level of risk as the current criterion (61 FR 65160). Therefore, the Commission has determined that the special circumstances described in 10 CFR 50.12(a)(2)(ii) exist in that application of the 25 rem whole body criterion is not necessary to achieve the underlying purpose of the rule because 25 rem TEDE is essentially the same level of risk. On this basis, the Commission concludes that the AP600 design review can be performed pursuant to the new dose criterion [25 rem TEDE] and an exemption from the requirements of 10 CFR 50.34(a)(1) is authorized by law, will not present an undue risk to public health and

- safety, and is consistent with the common defense and security.

(2) Paragraph (f)(2)(iv) of 10 CFR 50.34 - 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. Westinghouse answered this requirement, in Section 18.8.2 of the DCD, with an integrated design rather than a stand-alone, add-on system, as is used at most current operating plants. Specifically, Westinghouse integrated the SPDS requirements into the design ,

requirements for the alarm and display systems. In NUREG-0800, the NRC staff indicated that, j for applicants who are in the early stages of the control room design, the " function of a separate l SPDS may be integrated into the overall control room design" (p.18.0-1). Therefore, the Commission has determined that the special circumstances described in 10 CFR 50.12(a)(2)(ii) exist in that the requirement for an SPDS console need not be applied in this particular circumstance to achieve the underlying purpose because Westinghouse has provided an acceptable altemative that accompl shes the intent of the regulation. On this basis, the Commission concludes that an exemption from the requirements of 10 CFR 50.34(f)(2)(iv) is authorized by law, will not present an undue risk to public health and safety, and is consistent with the common defense and security.

(3) Paragraphs (f)(2)(vii), (viii), (xxvi), and (xxviii) of 10 CFR 50.34 - Accident Source Termsin TID 14844.

21

. Pursuant to 10 CFR 52.47(a)(ii), an applicant for design certification must demonstrate compliance with any technically relevant TMl requirements in 10 CFR 50.34(f). The TMI

~

requirements in 10 CFR 50.34(f)(2)(vii), (viii), (xxvi), and (xxviii) refer to the accident source term in TlO 14844. Specifically,10 CFR 50.34(f)(2)(xxviii) requires the evaluation of pathways ,

that may lead to control room habitability problems "under accident conditions resulting in a TID 1 14844 source term release." Similar wording appears in requirements (vii), (viii), and (xxvi).

Westinghouse has adopted the new source term technology summarized in NUREG-1465,

" Accident Source Terms for Light-Water Nuclear Power Plants," dated February 1995, not the old TID 14844 source term cited in 10 CFR Part 50.34(f). The Commission has determined that the special circumstances described in 10 CFR 50.12(a)(ii) exist in that these regulations need not be applied in this particular circumstance to achieve the underlying purpose because Westinghouse has adopted acceptable attematives that accomplish the intent of the regulations that specify TID 14844. On this basis, the Commission concludes that a partial exemption from the requirements of paragraphs (f)(2)(vii), (viii), (xxvi), and (xxviii) of 10 CFR 50.34 is authorized by law, will not present an undue risk to public health and safety, and is consistent with the common defense and security.

(4) Paragraph (c)(1) of 10 CFR 50.62 - Auxiliary feedwater system.

The AP600 design relies on the pacsive residual heat removal system (PRHR) in lieu of an auxiliary or emergency feedwater system as its safety-related method of removing decay heat. Westinghouse requested an exemption from a portion of 10 CFR 50.62(c)(1), which requires auxiliary or emergency feedwater as an altemate system for decay heat removal during an ATWS event. The NRC staff concluded that Westinghouse met the intent of the rule by relying on the PRHR system to remove the decay heat and, thereby, met the underlying purpose of the rule. Therefore, the Commission has determined that the special circumstances described in 10 CFR 50.12(a)(2)(ii) exist in that the requirement for an auxiliary or emergency feedwater system is not necessary to achieve the underlying purpose of 10 CFR 50.62(c)(1),

because Westinghouse has adopted acceptable alternatives that accomplish the intent of this regulation, and the exemption is authorized by law, will not present an undue risk to public health and safety, and is consistent with the common defense and security.

(5) Appendix A to 10 CFR Part 50, GDC 17- Offsite Power Sources.

Westinghouse requested a partial exemption from the requirement in GDC 17 for a second offsite power supply circuit. The AP600 plant design supports an exemption to this requirement by providing safety-related " passive" systems. These passive safety-related systems only require electric power for valves and the related instrumentation. The onsite Class 1 E batteries and associated de and ac distribution systems can provide the power for these valves and instrumentation. Ir# addition, if no offsite power is available, it is expected that the non-safety-related onsite diesel generators would be available for important plant functions; however, this non-safety- related ac power is not relied on to maintain core cooling or containment integrity. Therefore, the Commission has determined that the special circumstances described in 10 CFR 50.12(a)(2)(ii) exist in that the requirement need not be applied in this particular circumstance to achieve the underlying purpose of having two offsite power sources because the AP600 design includes an acceptable attemative approach to accomplish safety functions that does not rely on power from the offsite system and, therefore, accomplishes the intent of the regulation. On this basis, the Commission concludes that a partial exemption from the requirements of GDC 17 is authorized by law, will not present an j 22  ;

undue risk to public health and safety, and is consistent with the common defense and security.

(6) Appendix A to 10 CFR Part 50, GDC 19 ;whole body dose criterion.

The NRC staff used a criterion of 5 rem TEDE for evaluating the radiological consequences of design basis accidents in the control room of the AP600 design, under GDC 19 of Appendix A to 10 CFR Part 50. The NRC staff used the 5 rem TEDE criterion to be l

consistent with the new reactor site criteria in 10 CFR 50.34(a)(1) (61 FR 65157], although j GDC 19 specifies . . . "5 rem whole body, or its equivalent to any part of the body". . . The Comniission adopted 25 rem TEDE as the new dose criterion for plant evaluation purposes, because this value is essentially the same level of risk as the current criteria (61 FR 65160).

Therefore, the Commission has determined that the special circumstances described in 10 CFR 50.12(a)(2)(ii) exist in that application of the 5 rem whole body criterion is not necessary to achieve the underlying purpose of the rule because 5 rem TEDE is essentially the same level of risk. On this basis, the Commission concludes that a partial exemption from GDC 19 is authorized by law, will not present an undue risk to public health and safety, and is consistent with the common defense and security.

E. Issue resolution.

The purpose of Section VI of this appendix is to identify the scope of issues that are l resolved by the Commission in this rulemaking and; therefore, are " matters resolved" within the l meaning and intent of 10 CFR 52.63(a)(4). The section is divided into five parts: (A) the l 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) the 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 design certification rule. Furthermore, paragraph A explicitly states the Commission's determination that this design provides adequate protection of 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 remainder of paragraph B describes the 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-1512), the Tier 1 and Tier 2 information (including the availability controls in Section 16.3 of the generic DCD), and the rulemaking record for this appendix are resolved within the meaning of 6 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. 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 Vill of this appendix continue to be matters resolved in connection with this rulemaking. Paragraph B.7 provides that, for those plants located on sites whose site parameters do not exceed those assumed in Westinghouse's evaluation of severe accident mitigation design attematives (SAMDAs), all issues with respect to SAMDAs arising under the National Environmental Policy Act of 1969 associated with the information in the 23

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. Environmental Assessment for this design and the information regarding SAMDAs in Appendix 1B of the generic DCD are also resolved within the meaning and intent of 52.63(a)(4). In the event an exemption from a site parameter is granted, the exemption applicant has the initial 1

_ 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 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 generic technical specifications in Section 16.1 of the DCD, were not completely or comprehensively reviewed at the design certification etage. Therefore, the special backfit provisions of 6 52.63 do not apply to operational requirements. However, all design changes will be controlled by the appropriate provision in Section Vill of this appendix.

Although the information in the DCD that is related to operational requirements was necessary j 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 9 52.63. As a result, if the NRC wanted to change a temperature limit and that operational change required a consequential change to a design feature, then the temperature limit backfit would be controlled by Section Vill (paragraph A or B) of this appendix. 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 wot.ld not be restricted by 9 52.63 (see Vill.C of this appendix).

Paragraph C does allow 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 6 52.63. The requirement to perform these testing i 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 l 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  ;

l scope of this design certification.

Paragraph D reiterates the restrictions (contained in Section Vill of this appendix) {

place <1 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. j Paragraph E provides the procedure for an interested member of the public to obtain i access to proprietary or safeguards information for the AP600 desiga, in order to request and participate in proceedings identified in VI.B of this appendix, viz., proceedings involving licenses i and applications which reference this appendix. As set forth in paragraph E, access must first I be sought from the design certification applicant. If Westinghouse 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 or safeguards information may be ordered by the Commission, but must be subject to an appropriate non-disclosure agreement.

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G. Duration of this appendix.

The purpose of Section Vil 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, under  ;

,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 I plant-specific departures from information in the plant-specific DCD must be made pursuant to the change processes in Section Vill of this appendix for the life of the plant.

H. Processes for changes and departures.

The purpose of Section Vill 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 reference this design certification rule. Section Vill is divided into three paragraphs, which correspond to Tier 1, Tier 2, and Operational requirements. The language of Section Vill 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 9 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), similar to a 9 50.59 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 departurec as " change processes."

Both Section Vill 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 1 (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 information i The change processes for Tier 1 information are covered in paragraph Vill.A. Generic l changes to Tier 1 are accomplished by rulemaking that amends the generic DCD and are '

govemed 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 i must include an opportunity for hearing with respect to the proposed change, as required by 10 25

r 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 ordera 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 govemed 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.

Tier 2 Information 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 Vill.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 adopted a "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 govemed by tne 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 necessarv either to bring the certification into compliance with the Commission's regulations applicab9 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 irformation would apply.

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 o' 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 "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 under 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 d1 sign 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 l plant-specific order, as required by 10 CFR 52.63(a)(3). The Commission determined that it l 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.

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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. 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 1 license, the exemption is subject to litigation in the same manner as other issues in the license hearing, consistent with 10 CFR 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 s (USO) 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 Section 16.1 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 i B.S. The bases for the plant-specific 'echnical specifications will be controlled by the bases control procedures for the plant-specnic technical specifications (analogous to the bases control provision in the improved Standard Technical Specifications). The definition of a USQ in paragraph B.S.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 exemotion 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 i uncertainty in severe accident issue resolutions, the Commission has adopted separate criteria in B.5.c for detennining 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 i 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 6xited the reactor vessel and the containment is being challenged (severe accidents). These design features are identified in Section 1.9.5 of the DCD, with other issues, and are 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.S.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 may have intended functions to meet " design basis" requirements and to resolve " severe accidents." If these design features are reviewed under paragraph Vill.B.5, then the appropriate criteria from either B.5.b or B.5.c are selected depending upon the function being changed.

An applicant or licensee that plans to depart from Tier 2 information, under Vill.B.5, must prepare a safety evaluation which provides the bases for the determination that the 27

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 Tables 14.3-1 through 14.3-8 and 19.59-29 of the generic DCD to ensure that the proposed change does not impact Tier 1. These tables contain various cross-references from the safety analyses and probabilistic risk assessment 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 analyses for the AP600 design. Westinghouse provided more detailed cross-references for important analysis assumptions that are included in Tier 1 in its revised response to RAI 640.60 (DCP/NRC 1440 - September 15,1998).

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 Vill.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 Vill.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 Vill.B.5 when departing from Tier 2 information, may petition to admit such a contention into the proceeding under B.5.f. This provision was included because 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. Therefore, it follows that properly-founded contentions alleging such incorrectly-implemented departures cannot be considered " resolved' by this rulemaking.

As set forth in B.S.f, the petition must comply with the requirements of 2.714(b)(2) and show that the departure does not comply with paragraph B.S. Any other party may file a response to the petition. l' 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. The creation of and restrictions on changing Tier 2* information resulted from the development of the Tier 1 information for the ABWR design. During this development process, the applicants for design 1 certification requested that the amount of information in Tier 1 be minimized to provide i additional flexibility for an applicant or licensee who references this appendix. Also, many {

codes, standards, and design processes, which were not specified in Tier 1, that are acceptable l for meeting ITAAC were specified in Tier 2. The result of these actions is that certain l

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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 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 facility.

The determining factors were the Tier 1 information that would govem these areas after first full power and the NRC's judgement on whether prior approval was required before implementation of the change due to the significance of the information. Therefore, certain Tier 2* Information listed in paragraph B.6.c ceases to retain its Tier 2* designation 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 paragraph B.6.b retains its Tier 2* designation throughout the duration of the license, including any period of renewal.

Certain preoperational tests in paragraph B.6.c are designated to be performed only for the first plant or first three plants that reference this appendix. Westinghouse's basis for performing these "first-plant-only" and "first-three-plants-only" preoperational tests is provided in Section 14.2.5 of the DCD. The NRC staff found Westinghouse's basis for performing these tests and its justification for only performing the tests on the first-plant or first-three-plants acceptable. The NRC staff's decision was based on the need to verify that plant-specific manufacturing and/or construction variations do not adversely impact the predicted performance of certain passive safety systems, while recognizing that these special tests will result in significant thermal transients being applied to critical plant components. The NRC staff believes that the range of manufacturing or construction variations that could adversely affect the relevant passive safety systems will be adequately disclosed after performing the designated tests on the first plant, or the first three plants, as applicable. The COL action item in Section 14.4.6 of the DCD states that subsequent plants shall either perform these preoperational tests' or justify that the r,esults of the first-plant-only or first-three-plant-only tests are applicable to the subsequent plant. The Tier 2* designation for these tests will expire after the first plant or first three plants complete these tests, as indicated in paragraph B.6.c.

If Tier 2* Information is changed in a generic rulemaking, the designation of the new information (Tier 1,2*, or 2) would also be deten 9ned in the rulemaking and the appropriate process for future changes would apply. If a plant +pecific 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. 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.

Any requests for departures from Tier 2* information that affect Tier 1 must also comply with the requirements in Vill.A of this appendix.

Operational Requirements l The change process for technical specifications and other operational requirements in i the DCD is set forth in paragraph Vill.C. This change process has elements similar to the Tier

'1 and Tier 2 change process in paragraphs Vill.A and Vill.B, but with significantly different change standards. Because of the different finality status for tachnical specifications and other operational requirements (refer to Ill.F of this SOC), the Commission decided to designate a 29

e special category of information, consisting of the technical specifications and other operational requirements, with its own change process in paragraph Vill.C. The key to using the change processes in Section Vill 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 Vill.A or Vill.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 Vill.C applies. The language in paragraph Vill.C also distinguishes between generic (Section 16.1 of DCD) 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 '

Section 16.1 of the DCD or other operational requirements in the generic DCD is accomplished by rulemaking and govemed 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 j extent to which an NRC safety conclusion in the FSER is being modified or changed. If it l cannot be determined that the technical specification or operational requirement was i comprehensively reviewed and finalized in the design certification rulemaking, then there is no I 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 Vill.C.1 are applicable to ail applicants or licensees (refer to Vill.C.2), unless the change is irrelevant because of a plant-specific departure.

Plant-specific departures may occur by either a Commission order under Vill.C.3 or an applicant's exemption request under Vill.C.4. The basis for determining if the technical ,

specification or operational requirement was completely reviewed and approved for these I processes is the same as for Vill.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 leamed 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 (Vill.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.

The bases for the generic technical specifications will be controlled by the change process in Section Vill.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.

' l. 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, 30

4 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 I 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 j 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). 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, I subsequent modifications must comply with the design descriptions in the DCD unless the applicable requirements in 10 CFR 52.97 and Section Vill of this appendix have been complied with. As discussed in Ill.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.

l 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-specific DCD.Section X a'so 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 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 j 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 considus the information to be resolved j within the meaning of 10 CFR 52.63(a)(4). Because this infor nation 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 (referenced) 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 Vill 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, 31

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 Ill.D of th'; 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.

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 b 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 in paragraph B.3 vary according to four different time periods during a facilities' lifetime. 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 summary 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.  :

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 Vill 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 preoperational 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 X.B.3.d, which is consistent with the requirement for plants licensed under 10 CFR Part 50.

32

i IV. FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY l

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 proposed design certification rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and, therefore, an environmental impact statement (EIS) is not required. The basis for this determination, as documented in the environmental assessment, is that this amendment to 10 CFR Part 52 would not authorize the siting, construction, or operation of a facility using the AP600 design; it would only codify the AP600 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 environmental assessment for the AP600 design, the NRC 1

reviewed Westinghouse's evaluation of various design altematives to prevent and mitigate severe accidents in Appendix 1B of the AP600 Standard Safety Analysis Report (SSAR). The Commission finds that Westinghouse's evaluation provides a reasonable assurance that certifying the AP600 design will not exclude severe accident mitigation design alternatives for a future facility that would prove cost beneficial had they been considered as part of the original

' design certification application. These issues are considered resolved for the AP600 design.

The environmental assessment (EA), upon which the Commission's finding of no significant impact is based, and AP600 SSAR are available for examination and copying at the NRC Public Document Room,2120 L Street, NW. (Lower Level), Washington, DC. Single copies of the EA are also available from Jerry N. Wilson, Mailstop O-12 G15, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

V. PAPERWORK REDUCTION ACT STATEMENT This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This rule has been submitted to the Office of Management and Budget for review and approval of the information collection requirements.

The public reporting burden for this information collection is estimated to average 8 person-hours per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the information collection. The NRC is seeking public comment on the potential impact of the information collections contained in the proposed rule and'on the following issues:

1. Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?

2. Is the estimate of burden accurate?

3. Is there a way to enhance the quality, utility, and claTHE AP600 STANDARD PLAbrD l DOCKET NO. Sie3

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. 1 TABLE OF CONTENTS 1.0 I NTROD U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.0 THE NEED FOR THE PROPOSED ACTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3.0 ALTERNATIVES TO THE PROPOSED ACTION . . . . . . . . . . . . . .......... 4 3.1 Severe Accident Mitigation Design Altematives AMDA 3.2 Potential SAMDAs identified by Westinghou

.M. .......... 4

.... .......... 6 3.3 Staff Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . - .......

3.4 Risk Reduction Potential of SAMDAs . . . . ....... . 11 ... 0 2 3.4.1 Westinghouse Evaluation . . . . . . .... ....... 10 l 3.4.2 Staff Evaluation . . . . . . . . . . . . . ............ ... 12 3.5 Cost impacts of Candidate SAMDAs . . . . ..................... 12 3.5.1 Westinghouse Evaluation . . . . . . . . ................... 12 ,

3.5.2 Staff Evaluation . . . . . . . . . . . . . .. ................ 12 3.6 Cost-Benefit Comparison . . . . . . . . . . . . . . . . . . ............. 14  !

3.6.1 Westinghouse Evaluation . . . ....... ............ 14 3.6.2 Staff Evaluation . . . . . . . 'A~ ....... ............... 15 3.7 Further Considerations . . . . . . .. ..... ................ 17 3.7.1 Uncertainties in Cor ' mage '

ccident-Related Exposures . . . . ...... .... .................... 17 l 3.7.2 Reassessment esign , mative enefit Relationships i in Light of U intie .......g.........................19 3.7.3 Further Ev n of n Altern8tives With Potentially Favorable Cost-Ben acto ... ............................21 3.8 .Gonclusions . . ...... ..............................24 4.0 TH5 ME T OF TE ROPOSED ACTION . . . . . . . . . . . . . . . . 25 5.0 AGENC .

RSO . - LTED, AND SOURCES USED . . . . . . . . . . . . . . 26 Table 1 ted efits from Averted Offsite Exposure . . . . . . . . . . . . . 11 Tabi ey Differen the Westinghouse Approach and NUREG/BR-0058 ... 16 T 3 Key Paramete by FORECAST in Evaluating Maximum SAMDA Benefits . . 19 SAMDA ts Accounting for Uncertainties and Extemal Events Effects (Benefits, 1 996$) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 Maximum Benefit from Individual SAMDAs . . . . . . . . . . . . . . . . . . . . . . 22 2

1.0 INTRODUCTION

The U.S. Nuclear Regulatory Commission (NRC or the Commission) has issued a design certification for the Advanced Passive 600 (AP600) design in response to an application submitted on June 16,1992, by the Westinghouse Electric Corporation (hereinafter referred to as Westinghouse). A design certification is a rulemaking that ame;1ds Title 10, Part 52 of the Code of FederalRegulations (10 CFR Part 52).

This report presents the environmental assessment (EA) for this rule , h the NRC has prepared in accordance with 10 CFR Part 51 and the r uireme (P, National Environmental Policy Act of 1%9 (NEPA), as amended. a environmer. cal impacts of issuing a design certification. In addition, thi "rt add ccided mitigation design alternatives (SAMDAs), which the N as deci rt s final EA for the AP600 design. This report does not a the ronmer constructing and operating a facility which references sign certifi particular site; such impacts will be evaluated as part tion (s) for sitin ,

constructing, and operating such a facility.

As detailed in Section 4.0 of this report, the NRC det ined his design certification does not constitute a major Federal nifica g the quality of the human environment. This finding of no signif based act that the design certification would not independently autho ' ructi r operation of an AP600 reactor design. Rather, the cert n e AP600 design in a rule that could be referenced in a constructi rmit ), it (ESP), combined operating license (COL), or operatin '

nse ( pplica rther, because the certification is a rule,it does not i any r rces th Id have attemative uses.

Therefore, the NRC has ipre an envi mentalimpact statement (EIS)in connection with this action.

In addit ant to NRCa Westinghouse's evaluation of SAMDAs that ge t sign. hat basis, the NRC found that the evaluation provides ass rtifying the AP600 design will not exclude SAMDAs for a future facil ficial had they been considered as part of the original design certifi ion. ues are considered resolved for the AP600 design certificatio 2.0 EED FOR SED ACTION RC has long sou e safety benefits of ccmmercial nuclear power plant rdization, as s the early resolution of design issues and the finality of these

s. The N ans to achieve these benefits by certifying standard plant designs.

B to 10 art 52 allows for certification in the form of rulemaking of an essentially ant design.

Th action would amend 10 CFR Part 52 to certify the AP600 design. The amendment would allow prospective licensee's to reference the certified AP600 design as part of an ESP or a COL application under 10 CFR Part 52 or for a CP application under 10 CFR Part 50. Those portions of the AP600 design included in the scope of the certification 3

.i rulemaking would not be subject to further regulatory review or approval. In addition, the amendment would eliminate the need to consider SAMDAs for any future facilities that l reference the certified AP600 design. '

3.0 ALTERNATIVES TO THE PROPOSED ACTION The NRC had two attematives to certifying the AP600 design in an ame'ndment t CFR Part 52.' Specifically, the NRC could (1) take no action to approve the design, issue a '

final design approval (FDA) without certifying the design. In and of ese  !

attematives would not have a significant impact on the qual' of the vironment  !

because they would not authorize the siting, construction, ra .

)

1 in the first case, the NRC would not approve the desig erefor it us  !

the AP600 design would require licensing under 10 C art 50 CFR C, l as a custom plant application. Moreover, all design i lave to be part of each application to construct and operate an AP600 particular site. result, this altemative would not achieye the benefits of standa ovide early resolution of design issues, or permit finality of design issue resolut' .

- In the second case, the NRC would issue an FDA Appe FR Part 52, but would not certify the design in a rulemaking. T ough t would have

. approved the design, the design could be .

uldr are reevaluation as part of each application to construct and oper NAP rticular site. This alternative would permit early resol

• issues leve the benefits of standardization or establish of desig e re on.

The NRC sees no advantage fr, alte e com to the design certification rulemaking proposed for the A desig .

u ither the attematives nor the proposed design certif' tion rulema ld sig the quality of the human environment in and of th es, the achie rdization. permits early resolution of design issues, of r design s (including SAMDAs) that are within the scope of rtif' fore, the NRC concludes that the alternatives to rulemaking .leve es the Commission intended by certifying the AP600 design pursuan Part rt B.

3.1 n in iv ent with its obj standardization and early resolution of design issues, the ission decided to luate SAMDAs as part of the design certification for the AP600

n. In a 1985 pol' statement, the Commission defined the term " severe accident" as an that is "be e substantial coverage of design-basis events," including events in re is s tial damage to the reactor core (whether or not there are serious offsite sign-basis events are considered to be those analyzed in accordance with rd Review Plan (NUREG-0800) and documented in Chapter 15 of the AP600 Des trol Document (DCD).

As part of its design certification application, Westinghouse performed a probabilistic risk assessment (PRA) for the AP600 design to achieve the following objectives:

4

o

]

l I

I identify the dominant severe accident sequences and associated source terms for the j design.

Modify the design, on the bases of PRA insights, to prevent or mitigate severe accidents and reduce the risk of severe accidents.

Provide a basis for concluding that all reasonable steps have been taken to reduce the chances of occurrence, and to mitigate the consequences, of severe accidents.

~

Westinghouse's PRA analysis is presented in Chapter 19 of the AP6 Safety Analysis Report (SSAR).

In addition to considering attematives to the rulemakin ocess a 3.

applicants for reactor design certification or cps must "a consi itema .

es for severe accidents consistent with the requirements 18 art 50, as w ruling related to NEPA. These requirements can be s .

as follows:

10 CFR 50.34(f)(1)(i) requires the applicant to pM o e-specific probabilistic risk assessment, the aim of which is to seek sUch imp the reliability of core and containment heat removal systems a nificant9 I and do not impact excessively on the plant.

. The U.S. Court of Appeals decisi Ume-k

~m A . v. NRC,869 F.2d 719 (3rd Cir.1989), effectively requ' the NRdy.to int $ deration of certain SAMDAs in the environmentalimpact r perfcded undh@)iMiMion 102(2)(c) of NE of the OL application.

Although these two requiremen e not rel they share a common purpose to consider alt atives to th ed des e potential altematives improvements in I the plant ,which I fety pe uring severe accidents, and to prevent  !

viable I from ed. Id be noted that the Commission is not I required tg ! halte sign in this EA on the rulemaking; however, as a matter of di n etermined that considering SAMDAs is consistent with the intent '

art 5 resolution of issues, finality of design issue resolution, bene standardization.

In its in Lime Action v. NRC, the Court of Appeals for the Third Circuit e ed its opinion th Id likely be difficult to evaluate SAMDAs for NEPA purposes on a ric basis. Howe , e NRC has determined that generic evaluation of SAMDAs for the standard desi warranted for two significant reasons. First, the design and on of all referencing the certified AP600 design will be govemed by the rule a si ign. Second, the site parameters specified in the rule and the AP600 consequences for a reasonable set of SAMDAs for the AP600 design. The of the AP600 and limited potential for further risk reductions provides high at additional cost beneficial SAMDAs would not be found. Should the actual parameters for a particular site exceed those assumed in the rule and the SSAR, SAMDAs would have to be reevaluated in the site-specific environrnental report and the EIS.

5 L._

o 3.2 Potential SAMDAs identified by Westinohouse To identify candidate design altematives, Westinghouse reviewed the design alternatives for other plants including Limerick, Comanche Peak, and the Combustion Engineering (CE)

System 80+ design. Westinghouse also reviewed the results of the AP600 PRA and design attematives suggested by AP600 design personnel.

Appendix 18 of the SSAR does not explicitly state whether Westinghouse's evadlon included plant improvements considered as part of the NRC's Containment Pe provement (CPI) program (NUREG/CR-5562, -5567, -5575, and -5630 Howe house stated that the types of design changes identified in the CPI pr ave n incorporated into the AP600 design or have been considered as desig emativ ements identified in the CPI program were also evaluated in o m Westinghouse, including the CE System 80+ design a , native ations.

Westinghouse eliminated certain SAMDAs from furthe ion on the basi they are already incorporated in the AP600 design. Such featur following:

. hydrogen ignition system reactor cavity flooding system reactor coolant pump seal cooling (AP mot s)

. reactor coolant system depressuriza I extemal reactor vessel cooling .

. nonsafety-grade containment On the basis of the screening, W . ouse ined 14, tential SAMDAs for further consideration. These SAMDAs, edi ction 1 f the SSAR, are summarized below:

(1) U "

the Chem Volum tem (CVCS) for Small Loss-of-Coolant LOC VCSis pable of maintaining the reactor cooling

)i OCAs ffective break sizes up to 0.97 cm (% in.) in l di s de would extend the capability of the CVCS so that it cou e RC uring small and intermediate LOCAs up to an {

effe f15 1.)in diameter. Implementation of this design alte l

're in tion of in-containment refueling water storage tank j T) / ' ulation connections to the CVCS, as well as the addition of '

nd line S pumps to the RCS. Westinghouse estimated that l mplomenting thi attemative would reduce plar't risk by at most 5.5E-04 I person-rem /yr.

Filtered Con ent Vent: This design altemative would involve the installation of a Itered ment vent, including all associated piping and penetrations. This would provide a means to vent the containment to prevent catastrophic ure failures, as well as a filtering capability for source term release. The vent would reduce the risk associated with late containment failures that might occur after failure of the passive containment cooling system (PCS). However, even if the PCS fails, air cooling would be expected to limit the containment pressure to less 6

than the ultimate pressure. Westinghouse estimated that implementing this design alternative would reduce plant risk by at most 1.0E-03 person-rem / reactor-year.

(3) Self-Actuating Containment isolation Valves: Self-actuating containment isolation valves could increase the likelihood of successful containment isolation during a severe accident. This design altemative would involve adding a self-actuating valve or enhancing the existing containment isolation valves on containment penetrptions that are normally-open. (Specifically, penetrations that provide normally hways to the environment during power and normal shutdown conditions i Id permit automatic self-actuation in the event that containmen('conditi e a severe accident. Closed systems inside and outside con nt,s ual heat removal system and component cooling, would luded f n attemati Westinghouse estimated that implementing this ign att p risk by at most 7.4E-04 person-rem /yr.

(4) Passive Containment Sprays: Installing a pass lated containm spray system could result in the following risk benefits:

(a) Scrub fission products, primarily for 7 htion failure.

(b) Provide an attemative means to fl ' the re (in-vessel retention).

(c) Control containment pressure s in wh . has failed.

Westinghouse estimated that implem 'gn alte would reduce plant risk by at most 6.9E-03 person-re m rese minating all release categories except containment b .

(5) Active High-Pressure Safety on S m: A fety-related, active high-pressure safety injection s as wou 'able th actor to prevent a core melt in all events except excessive Yand ipated sient without scram. Note, however, that this design attemat not ent e AP600 design objectives, in that it wou,(change 1 the rom a p passive systems to a plant with both pa&nd a . Wes stimated that implementing this design ul t riskEEst 6.1E-03 person-rem /yr.

'(6) Stea She emoval System: This design attemative would

. involve assiv lated heat removal system to the secondary side of the Thi ancement would provide closed loop secondary system i gvia ral circulation and stored water cooling, thereby preventing i of the prim k given loss of startup feedwater and the passive residual 1 eat removal he nger. Westinghouse estimated that implementing this design attemative wou uce plant risk by at most 5.3E-04 person-rem /yr.

Direct St nerator Safety and Relief Valve Flow to the IRWST: To prevent or uce f roduct release from bypassing containment during an steam generator (SGTR) event, flow from the steam generator safety and relief valves could to the IRWST. An attemative, lower cost approach to this design attemative be to redirect the flow only from the first stage safety valve to the IRWST.

Westinghouse estimated that implementing this design attemative would reduce plant risk by at most 4.2E-04 person-rem /yr.

7

(8) Increased Steam Generator Pressure Capability: In lieu of design altemative (7) above, fission product release bypassing containment could be prevented or reduced by increasing the steam generator secondary side and safety valve set point to a level high

- enough so that an SGTR will not cause the secondary system safety valve to open.

Although detailed analyses have not been performed, it is estimated that the secondary side design pressure would have to be increased by several hundred psi to make this attemative effective. Westinghouse estimated that implementing this design attemative would reduce plant risk by at most 4.2E-04 person-rem /yr.

(9) Secondary Containment Filtered Ventilation: This design alte ld involve installing a passive charcoal and high-efficiency p te a m for the middle and lower annulus region of the seconda crete co low Elevation 135'-3"). Drawing a partial vacuum o middle cto motive power from compressed gas tanks wou lperat ilter s n attemative would reduce particulate fission p l ,from any fa , ment penetrations. Westinghouse estimated that im j this design alte e would reduce plant risk by at most 7.4E-04 person-re 1 ,

(10) Diverse IRWST Injection Valves: In the curre sign Ive in series with a check valve isolates each of the four IRW on pa de diversity, the design could be modified so that a diffe rovide es in two of the lines. This enhancement would r "

of com cause failures of the four IRWST injection paths. We ouse lementing this design attemative would reduce plant ri y at . n-rem / reactor-year, which would represent eliminating re dar '

e sequ" resulting from a failure of IRWST injection (3BE seg s).

(11) Diverse Containment R lation s: current design, a squib valve isolates eac .of the four nt recir . In two of the four paths, each of the sq es is i a ch in the remaining two paths, each squib wiId rie r lve (MOV). To provide diversity, the design c5 lified d trent vendor provides the squib valves in two lines. This enha uldr lihood of common cause failures of the four contain latio estinghouse estimated that implementing this design alte pi isk by at most 1.5E-04 person-rem / reactor-year, which represe all coro damage sequences resulting from a failure of tainment r 3BL sequences).

(1 Ex-Vessel Core t er: This design alternative would inhibit core-concrete interaction (CCl), even in es where the debris bed dries out. The enhancement would involve designing ructure in the containment cavity or using a special concrete or coating.

he curr 600 design incorporates a wet cavity design in which ex-vessel cooling is tain core debris within the vessel. In cases where reactor vessel flooding

, the PRA assumes that containment failure occurs from an ex-vecsel steam '

or CCI. Westinghouse estimated that implementing this design attemative would reduce plant risk by at most 6.1E-03 person rem / reactor-year.

8

o (13) High Pressure Containment Design: The proposed high-pressure containment design j would have a design pressure of approximately 300 psi, and would include a passive  !

cooling feature similar to the existing containment design. This design would reduce the likelihood containment failures from severe accident phenomena such as steam explosions and hydrogen detonation. However, this alternative would not reduce the frequency or magnitude of releases from an unisolated containment. Westinghouse estimated that implementing this design attemative would reduce plant riskpy at most 6.1E-03 person-rem / reactor-year.

(14) Increased Reliability of the Diverse Actuation Systerg DAS): n altemative would involve improving the reliability of the DAS. AS WMiaty system that can automatically trip the reactor and turbine and ate certF1'2;ed safety y features (ESF) equipment if the protection and s%ty monitorpNysguYda'hoable t  :

perform these functions. In addition, the DAS piscuides dinrse monitdMMd l plant parameters to guide manual operation and . ctor trip andhM '

actuations. Westinghouse estimated that imple" is design altemitEdould reduce plant risk by at most 2.2E-04 person-re r.

3.3 Staff Evaluation The staff reviewed the set of potential SAMD West kmse and found it to be

' {

reasonably complete. The activity was acc wing d 'n alternatives '

associated with the following plants: LimerkW(NUR "che Peak (NUREG-0775),

CE System 80+ (NUREG-1462), Wattsgd (NUR -04 Is ABWR (NUREG 1503). ,

Also surveyed were accident mana rapnt stratages (NU, -5474), and alternatives 1 identified through the Containmen rmance improvenment (CPI) Program I (NUREG/CR-5567, -5575, -5630 56 I The results the staff's a ^ ent are summerbippin Appendix A to the " Review of Severe Accident ^ n Des tives (SAGD 6lp)ifor the Westinghouse AP600 Design" (SEA 97-270 # ep , ce anciElig neering Associates Inc., and dated August 29, 1997. . brief s each of the design attematives identified in the

^

foregoing re . ^ ^? Iso i e Westinghouse AP600 design attematives, which R. In all, the staff reviewed more than 120 possible are designdiscussed alter # f? 'MX ~I ost ~ 1'ovements B of' identified as part of the NRC's CPI program.

Specific #'oveme5 i applicable to the AP600 included a filtered containment vent ed rubble b .

ntion device, two improvements specifically mentioned in NU -0660 for evalu part of Three Mile Island (TMI) Item II.B.8. The list of 120 also i potential SAM oriented toward reducing the risk from major contributors to risk for

, including SG vents.

(anda in 1 the use analysis did not consider several design alternatives, in most attematives are either (1) already included in the AP600 design, or (2)

"s of risk reduction by one or more of the design attematives that were included ghouse analysis, in other cases, the design alternatives were pertinent only to boiling water reactors. The staff's preliminary review did not reveal any additional design attematives that obviously should have been considered by Westinghouse. Also, 9

Westinghouse considered some potential design alternatives to be considerations for accident management strategies rather than design altematives. {

The staff noted that the set of SAMDAs reviewed by Westinghouse is not all inclusive, in that additional (perhaps less-expensive) SAMDAs could be postulated. However, the benefits offered by any additional modifications would not likely exceed those for the modifications evaluated, and the costs of attemative improvements are not expected to be less n those of the least expensive improvements evaluated, when the subaidiary costs associa th maintenance, procedures, and training are considered.

The discussions in Appendix 1B of the SSAR do not spe ' bas ess that Westinghouse used to screen the many possible desig atives t finallist 14 selected for further evaluation. Similarly, Westing 's re equ for additional information (RAls) provided few addition ight he p ss, as noted above, the staff's review of the more than 1 esigns did any new attematives likely to be more cost-beneficial than ded in Westing 's evaluation of AP600 design altematives. On this basis, cludes that the set of 4 potential SAMDAs identified by Westinghouse is acce ,

3.4 Risk Reduction Potential of SAMDAs 3.4.1 Westinghouse Evaluation in its evaluation, Westinghouse assume lat ea e would work perfectly to completely eliminate the respective nt nces. umption is conservative, as it maximizes the benefit of each des mat which i easured on the basis of risk reduction. (For example, the ris ion ai igned t *ss ive containment sprays assumes that all release categodes exce Westingh used ana tain ]

is an eliminated.) In each case, ined in the AP600 PRA to estimate the risk red each mative. I ouse then expressed the risk reduction in terms pe .

yearrj d by the total population within a radius of 80.5 km apt . Each of the 14 design attematives was evaluated separately.

Table 1 of s We house's risk reduction estimates by comparing the benefits erting ure using each potential design attemative. The bases for the . mates are p ction 1.B.7 of the SSAR.

10

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p 3.4.2 Staff Evaluation The staff reviewed Westinghouse's bases for estimating the risk reduction associated with the various SAMDAs, and concluded that Westinghouse used a reasonable, and generally conservative, ratic nale and assumptions as the bases for the risk reduction estimates regarding each design attemative.

The level of risk reduction estimated for the various SAMDAs is driven two rlying assumptions in the methodology. Specifically, Westinghouse's risk r fdlitimates reflect only the contribution from intemal events initiated at power ' d theq  : estimate (mean) values without consideration of uncertainties in core da uenc% offsite consequences. Although this is consistent with the app h taken * 'gn attemative evaluations, further consideration of these fdcTors coul d to ~

risk reduction values, given the extremely small CDF sk e ates in th for intemal events.

In assessing the risk reduction potential of SAMDAs for esign, the staff considered Westinghouse's risk reduction estimates for the variou ^ ,

conjunction with supplementary parametric analyses to evaluate the ' tial emal events and uncertaintles. These analyses are further discusaad%ction .

3.5 Cost imoacts of Candidate SAMDAs-3.5.1 Westinghouse Evaluation Sections 1 ,1 B.4.3, of the the capital cost estimates for the AP600 design a s eval estin Table 1B.8-1 of the SSAR presents the results eval gificall each design altemative, Table 18.8-1 lists the potential , th it (assuming the design attemative was highly effective in reducing a ),th , and the net capital benefit. Notably, Westinghouse' ations _ _. ccount for factors such as design engineering, testing, an iat th each design attemative. if included, these factors would i .

e the and decrease the capital benefits of each alternative. Thus, thisa ch is conse Staff Evaluation nableness of the cost estimates that Westinghouse presented in the SSAR, the capital costs for the AP600 design altematives with those evaluated for the nd CE System 80+ designs. However, there is not an exact match in the design -

attematives among the reactor designs, so only broad comparisons are possible.

12

For example, the AP600 active high-pressure safety injection system, which is estimated to cost $20 million, adds an active high-pressure safety injection pump and associated piping, valves, and supports, thus adding an entire new safety-related system to the AP600 design.

This attemative can be compared to the attemative high-pressure safety injection for the CE System 80+ design, which is estimated to cost $2.2 million. However, the design alternative for the CE System 80+ design simply adds parallel piping and valves to an existing system, which would be expected to cost only a fraction of the total system price.

Similarly, the filtered containment vent for the AP600 design,can be k[ to systems with similar functions for the ABWR and the CE System 80+ . Tiik%P900) design included a filtered containment vent and all associated piping and tions. gjesign adde &

an ex-containment filter system to an existing venting s m. T h g" "n incluMd a filtered containment vent similar to the multi-venturi European plants. The estimated costs for the three v '

• • bbin ems ims

'C "ms- $5 milli 8l%on, e and $10 million, respectively- reasonably agree with given the differences in the designs. m The costs for the non safety-grade containment 1

. r AP600 ich was evaluated in an earlier version of SSAR Section 1B befor . . rporat AP600 design, can be compared to the reactor building sprays f T;- . 7 4 n an attemative containment spray for CE System 80+ de . ' . Thi

~

emative involves adding piping and spray headers inside containment, an nn . isting fire water system.

Similarly, for the ABWR, the existing i " ntain [ t fire "

tem would be modified to provide sprays in areas vulnerable ~ ion p uct release. The ABWR modification would thus be limited 80+ design, to providing this attemative involues adding spray (2]illfy necttoto s[' ingedin-containment the existing areyj6f containment. Fo spraysystegtogetherwit ' " umps t f . ater. Estimated costs for these three spray ere $41 he AP , $100,000 for the ABWR design, and

$1.5 mi CE esign. ht of the scope differences among these design a je est AP600 spray system appear to be reasonable.

These co .

I at t st estimates for several of the AP600 design attemat. easona the costs for roughly similar design attematives evaluated for the ABWR and the CE + designs.

I her assess the sonableness of the AP600 design attemative cost estimates, the staff indepenc cost estimates for one particular design attemative, the active, non ted ment spray system. (This analysis was performed before the ray system was incorporated into the AP600 design and deleted from SSAR e assessment assumed the addition of fire protection system grade spray he supply piping inside containment (carbon steel), and the addition of control valves and piping outside containment and connected to the existing fire water supply system. The resulting costs for the containment spray system ranged from about $300,000 to $350,000 (1996 dollars), depending on the assumptions made regarding the required pipe size. These 13

Independent estimates did not include design engineering; first-of-a-kind costs; or allowances for associated personnel training, procedure development, or recurring operations and me!ntenance costs. This approach is similar to that used in Westinghouse's cost estimation.

Thus, the Westinghouse estimate of $415,000 for this design attemative reasonably agrees with the independent estimate. In addition, the staff developed an independent cost estimate for a containment spray system similar to that described above, but with increased pumping capacity. (The increased pumping capacity is needed because Westinghouse's lett,er of March 13,1997, indicated that the currently designed fire water supply s delivering less than 1.89 kUmin (500 gpm) to the proposed containme r ,ystem ispapable o stem.) The

, 'ity so that each system pump wouldevaluated for this11.36 be capable of delivering altemative kUmin (3000would increase guer to the,- the firep}ater nt sprays p e piping u "

against a containment the containment pressure in the current design would beofincreasdd 310.3 kPa (30 psig). @in size todtjIuc lM, fir sistardi#

This modification to the AP600 design was estimated ttii60st aboudS370, .#As with the foregoing estimate, no allowance was made f development, or recurring operations and maintenance

~

• training, p

• $7 On the basis of this audit, the staff viewed Westingho 's ap st estimates as adequate, given the uncertainties surrounding the undeglying co , and the level of precision necessary given the greater uncertai . ' [i ~~ ,on the M"ide, with which these costs were compared. .' ' 7^

Y 3.6 Cost-Benefit Comoarison .

3.6.1 Westinghouse Evaluation .; j

g. k

After c e ris "1 potentia cost impact of the various SAMDAs, Westing ed cl : mmparison to determine whether any of the potential severe-acci on de would be Justified. To do so Westinghouse assessed the ch d mative in terms of potential risk reduction, which was defined as . ole- . person-rem per year received by the total population within a 80.5 the AP600 plant site. Westinghouse then assigned a value of $1 to each pers verted offsite exposure, which was assumed to account for bo alth effects and roperty damage. This value was treated as the annual levelized b t for averted risk.

ine th imum expenditure justified by a given reduction in risk (" maximum capital use divided the annual levelized benefit by the annual levelized fixed e annual levelized fixed charge rate was determined to be 15.7 percent in curr dollars on the basis of factors and methods provided in documents developed by the Electric Power Research Institute (EPRI P-6587-L) and the U.S. Department of Energy (DOE /NE-0095). Westinghouse calculated the %d charge rate using a component " book life" of 30 years. The use of a high charge rate tends e minimize the capital benefit associated with 14

each design altemative. Nevertheless, the 30-year life used in the calculations makes little difference in the economic benefit compared to the more typical 60-year life, particularly when the high levelized annual fixed charge rate of 15.7 percent is used.

The Westinghouse approach for calculating the benefits or reduced risk from each individual design attemative also does not give credit for averted onsite property damage ang replacement energy costs which are realized through a reduction in accident fr cy. The onsite property damage and replacement energy costs may have bee ' I use the estimated CDF is very low. However, as indicated below, t e,se ons' ations can substantially add to the benefits that may be achieved usi " , ign ' .

Table 1 of this EA reports Westinghouse's cost-benefit timate using a screening criterion of $1,000/ person-rem-averts each r _ w"$f 1

whether ar SAMDAs could be cost effective. As shown in Table 1 t capital W. ben,efifeElated

% by Westinghouse for any design attemative is about $50, ital cost for the least expensive design altemative is $33,000.' On this basis concluded that no additional modifications to the AP600 design are war ted.

3.6.2 Staff Evaluation

- The NRC recently updated its recom a ch fo onetary conversion of radiation exposures. . Previous guidance s that erson-r of exposure should be valued at

$1,000. This conversion factor f te was in ed to account for both health effects and offsite property damage, ari xposu ' urred ture years were not to be discounted.

The recent idance give bl INRC's r ysis guidelines (NUREG/BR-0058, Revision me l em of exposure as the monetary conve in a i @l2,000 ssin es and impacts, future exposures are to be discount ,t theirM l , . rth. Offsite property damage from nuclear accidents is to be separa and is t the $2,000 per person-rem value.

Evaluati ecently Brookhaven National 1.aboratory for the NRC assessed total costs iated with ses, including both health effects and property damage / loss eff UREG/CR- ts were assessed for each of the five NUREG-1150 plants

( Gulf, Peach , equoyah, Surry and Zion). The results indicated that overall associated with ite releases of radioactive materials, presented on a cost per

-rem of e to the public, ranged from about $2,000 to more than $5,000 per

, g on factors such as the assumed interdiction criteria. A criterion of rem averted was added to account for offsite property damage and other r severe accidents. Thus, the Westinghouse cost-benefit evaluation approach use 600 design altematives is not consistent with the approach recommended in NUREG/BR-0058 Revision 2. The key differences are summarized in Table 2 of this EA, and the staff's independent evaluation is found below.

15

i i

l Table 2 Key Differences Between the Westinghouse Approach and NUREG/BR-0058  ;

Westinghouse's SAMDA Approach NUREG/BR-0058 Recommended Approach

$2,000 per person-rem averted to account for

$1,000 per person-rem averted (for valuing health effects, plus $3,000 perpj'gson-rem risk reduction) averted to account forather offeRe effects and related costs dk AF 15.7% discount rate 7% dis ate s u-a No accounting for benefits of averted onsito

[

Cons: , --qiven for benefits dWverted 1 cleanup and decontamination costs onsit contamination costs '

No accounting for benefits of averted . tion g nefits of averted replacement energy costs ener costs To arrive at a baseline potential be recommended approach in NUR 00 omt /eductior@P offsite risk, the staff applied the evision 290 the design altematives identified for the AP600 design. This EA us iscou of ent and assumed a reactor life of 60 years. The_ averted risk. desig . s taken from Table 18.8-1 of the SSAR. In , , the s mon rsion factors for radiation exposures. The first is t rson end UREG/BR-0058, Revision 2. The second,

$5, ' intenci nt for offsite property damage and health effects. The results for ea _ temat ', in columns 5 and 6 of Table 1. For comparison purposes, West tima e capital cost, averted risk, and capital benefit for each design alte asent lumns 2,3, and 4 of Table 1). A 100% effective design attema Id red and/or offsite releases to zero. Estimated benefits from a 100 % ive design a re also shown for each of the attemative cost bases (last row of T 1).

T uits shown in e 1 indicate that the benefits calculated using a 7-percent discount r year pl , and a $2,000/ person-rem conversion factor is about a factor of four lated by Westinghouse. The benefits calculated using are about a factor of 10 higher than those estimated by Westinghouse. The hl benefit shown in Table 1 amounts to less than $500, while the capital cost for the least expensive design alternative is $33,000. Thus, even with the highest benefit basis

($5,000/ person-rem, 7-percent discount rate, 60-year life), the calculated benefits are almost two orders of magnitude too small to justify the addition of any of the design altematives listed. It should be noted, however, that this assessment neglected the benefits from averted onsite 16

U

. l i

costs, which are relevant for design attematives that reduce core damage frequency. Dollar savings derived from averted onsite costs are treated as an offset or reduction in the capital cost of the design alternative in the staff's analysis. Averted onsite costs are significant for certain design attematives and are further considered below.

3.7 Further Considerations A

The estimates of potential design alternative benefits listed in Table 1 of this EA r48e'ct Westinghouse's estimates of averted risk and neglect the benefits fromfspertednisite costs. As mentioned in Section 3.2 of this EA, Westinghouse's risk estimates dor ^IJntfor uncertainties either in the CDF or in the offsite radiation e es r " a core dama e event. The uncertainties in both of these key elements a flylarge h " safety features of the AP600 design are unique, and their reli has been$a' gh j analysis and testing programs rather than operating ' nce.gddition, CDF and offsite exposures do not account for the add ' xternal eve e

earthquakes.

To further explore inese areas, the staff screened the r= _ __ -- mas to determine whether i any of the design attematives could be cost-beneficialfifien tWeWoiifedegefit analysis incorporates uncertainties, added risk from extema its, and Egte costs. The staff then performed a more detailed assessment fog]g Q attemaggspflaving potentially  ;

favorable cost-benefit factors under these m 1 eratio ese analyses are discussed in Sections 3.7.1 - 3.7.3 below [En,--

3.7.1 Uncertainties in Core Damag uency and Accid ated Exposures Revision 8 to the PRA discussed < icerta' in the e ated CDF for the AP600 design.

Specifically, the CDF uncertaintyj . ributio, ' '. rized by an error factor (EF) of about 5.7. Assuming a , distrib the ratio of the 95th percentile to the median, he rati edian t rcentile. Thus, the CDF for internal events could be six r than'sssumed in the analysis discussed above.

Additional fa uld s crease the estimated CDF for the AP600 plant include the cont eve ccident sequences that have not yet been identified, as well as th puence t have not yet been analyzed in the PRA. Examples of the latter _ e exte uch as fires and earthquakes. Notably, the CDF base estim f 1.7E-07/rea s not include the contribution of extemal events. In the PRA stinghouse ind at external events, in particular internal fires, are estimated to inc e the CDF by a a ' actor of four. However, the PRA available for this study did not d the potential co utions from seismic events, which could readily increase the CDF by r of magnit more. These extemal events can also degrade the containment nce,so e releases from containment may also be higher than for accidents al events.

The increases in CDF attributed to accident sequences that have not yet been identified is very difficult to estimate. Presumably, the contributions from such sequences should be small If Westinghouse performs the PRA in a thorough and systematic manner. For the purposes of 17 l

I the present analysis, the effects of these sequences are assumed to be captured by the potential increase in CDF attributed extemal events.

Section 18.6 of the SSAR presented Westinghouse's estimates of offsite exposures for the ma}or release categories (RCs) defined for the AP600 design. On the basis of the CDF reference value of 1.7E-07/ reactor-year and the total risk of 7.3E-03 person-rem / reactor-year, Westinghouse estimated that the " average" offsite Exposure is of the order of 50,000 person-rem per core damage event. . However, Westinghouse's documentation did not indicasiNhe uncertainty in the estimated releases.

l The average offsite exposure of 50,000 person-rem pcr A re .

nt estimated by Westinghouse is a factor of 2.7 lower than the average pj exposur for the fiveg current-generation nuclear plants addressed in NUREGj (after EG- p0

- plant releases to that of a 600-MWe plant). The better y orman f the A y be attributed, in part, to methods and assumptions for defi(

erms, as h likelihood of successful RCS depressurization and in-ves n of damag the l AP600 design.

Uncertainties in the offsite exposure estimates for the 600 ' nificant. As described in Section 19.1.3.3.3 of NUREG-1512, Fina , luation Report (FSER), the AP600 risk profile is shaped by th or ass s regarding j containment failure modes and release cha j conservative assumptions rega early from ex-vessel phenomena i optimistic assumptions that e I rea vessel will always prevent reactor I pressure vessel (RPV) brea '

substantial credit for add er movali GTR events if early conta' ment failure (as 2--

^

, deterministic calculations performed subseg a PRA) I breach instead results in a more benign r the pre @ intermediate time frame), overall risk for i release inm l would be .

. By contract, if credit for extemal reactor vessel cooling (ER or tainment failure frequency would increase proportionally, brea assumed to lead to early containment failure in the baseline P limit ssumption that ERVC always fails and leads to early contain ilure, nt failure frequency would approach the core melt frequency and ri uld increase of 20 (to about 0.16 person-rem /yr). Similarly, offsite fish can be s icantly impacted esign fails to realize the decontamination factor (DF) of 100 a to aerosol relea ions for SGTR events predicted by the materials access a zation program P) to account for fission product removal by impaction on steam tubes. W is credit for aerosol removal, the risk contribution from a containment minim rcent of the total). Without this credit, overall risk for intemal events factor of seven and would be dominated by containment bypass releases, id not credit the impact of the non safety-related containment spray system on fi releases. Containment sprays could significantly reduce the estimated risk in the baseline PRA (by perhaps a factor of 2), since the ' sprays would be effective in reducing the source terms in the risk-dominant RCs such as early containment failure (CFE) and containment isolation failure (Cl). However, sprays would not impact releases attributed to SGTR events.

18

p 4

in summary, the actual offsite exposure could range from a factor of two lower to an order of magnitude higher than the Westinghouse estimate, given the uncertainties in the underlying analyses of containment performance. This uncertainty range was factored into the staff's reassessment discussed below.

3.7.2 Reassessment of Design Alternative Cost-Benefit Relationships in Ught of Uncertainties A

The staff-performed analyses reassess the benefits of potential AP600 design altaptatives taking into account the uncertainties in estimated CDF, offsite releases of rad' terials given a

-severe accident, and effects of extemal events. For these analyses, mated the maximum benefits that can be achieved with AP600 desigrdsernat g that a attemative can either completely eliminate all core damag%nts or c9- yminate offsid releases of radioactive materials if a severe arcident doegoccur. T benefits, the staff used the FORECAST code (NUREG/ -5595 hastimald!'

sion 1, Regulatory Effects Cost Analysis Software Manual, Ve ience and Associates, Inc., July 1996). FORECAST allows the ush ,ainty ranges for" ey parameters and provides a means to combine uncertaint rameters. It also provides a distribution for the bottom line costs or ben 'esents a picture of the uncertainty in the " bottom line" figures. Table 3 of thisTA pre parameters used in evaluating the maximum potential benefit.

Table 3 Key Parameters Used by FOR5ji m ,jng m SAMDA Benefits Parameter [ [ MS Value Reference AP600 core damagehency[ 1.7EfNetor-year (EF=5.7)

Average public radiation expoelpe"per 4M200 person-rem (rounded to 50,000)

M accident: . ^ i- d$sumed error factor: 5)

Plant lifak Disc M

3 W ~

"60 years 7%

Conversl%k y $5000/ person-rem Replace,Mgk V $277,000/ day of downtime Ave [ cleanup ankkination costsa $1,690,000,000 / major accident Mrted replacementhd5y costs * $20,200,000,000/ major accident on NUREG/CR ounts for both offsite health effects and offsite property damage effects 8

guidance in NUREG/BR-0184 (not adjusted for AP600-specific features) 8 aver ment energy costs for pressurized water reactors in the 500 - 1000 MWe range F of estimating the maximum potential benefit from AP600 design attematives, the med that extemal events and accident sequences not yet accounted for in the PRA would increase the reference CDF by two orders of magnitude, (i.e., a factor of 100), with an EF of six used for this higher CDF. The staff then evaluated cases assuming the reference value of 50,000 person-rem per accident. Table 4 of this EA presents the results of this analysis.

19

. Table 4 SAMDA Benefits Accounting for Uncertainties and External Events Effects (Benefits,1996$)

% 5% 95 %

Description Confidence Mean Confidence h* ,

Level Level  ;

y (

Base CDF (1.7E-07/yr) and reference offsite release (50,000 person rem); design 1

attematives which reduce the accident '100s $26,600 frequency to zero .

g Base CDF increased by factor of 100 to account for external events and other 2 accident sequences not yet accounted for; other factors same as Case 1; design i $647,000 $2,257,000

. . + 'g altematives which reduce the accident J j frequency to zero P) '

+ .

Base CDF increased by factor to .

account for extemal events; o actors 3 same as Cam 1; design alt ves 1,700 $49,000 $223,000 l reduce the offsite releaset ro, by a not i change the accident freq f f I

& { bK The ent je 4 I hrlesign I s which prevent accidents (reduce the accident izero a re cost-effective than design attematives which reduce or eliminate f' ses b N ect on accident frequency. This is because of the fairly large ben ted $ onsite cleanup and decontamination costs, and avoided rep I ts, er of which are assumed to be impacted by design attemati h cci nt frequency.

Ca the reference zing the base CDF and Westinghouse-estimated offsite e res. In this case e timated benefits are considerably higher than those cited in .,

T 1 of this EA, pri y because they include averted onsite cleanup and decontamination i s well as ave replacement energy costs. l the effects of the higher CDF associated with extemal events, but do not i i of possible higher releases from containment attributed to such events. (in

, ese cases retain the base offsite exposure of 50,000 person-rem / event.) These cases may be used as the basic benefits including extemal events and assuming that extemal events would not impact containment performance. Case 2 shows the potential benefit range for a des!gn attemative which could reduce the accident frequency to zero. Case 3 applies to a l

20

,o.

l design attemative which would eliminate all offsite releases, but which would not impact the j CDF- ,

Table 5 of this EA combines the information in Tables 1 and 4 to estimate the total benefit i possible from specific design alternatives. The design attematives are divided between those l that impact the CDF and those that impact containment performance but not the CDF. Benefits '

have been estimated by taking the fractional reduction in risk for each design alternative (compared to the AP600 baseline risk as defined by Westinghouse) and. plyin t fraction to the mean benefits displayed in Table 4. Design alternatives that reduc " ere applied to the Case 2 mean benefit, while those that only effect contain ent pe were applied to the Case 3 mean benefit.

The values shown in Columns 4 through 7 of Table 5 re benefi o rr .

values. By contrast values shown in Columns 8 throug were lated. )

95*-percentile values. In other words, there is only a 5 ince th'at th fits will be greater than the values shown in Columns 8-11.

0ij io by a factor of The use offive, approximately thebutmaximum benefits does not alter any of the typicallyimproves overalt th[tonclusi sign attematives that have acceptable cost-benefit ratios.

3.7.3 Further Evaluation of Design Alternat Fa _ le Cost-Benefit Factors Design attematives that are within a d of m g efit criteria of

$5,000/ person-rem were subjected t er prop)l6ilistic erministic considerations, including a qualitative assessment foi .

a the impact of additional its tha Id or the design alternative if it would be off in reducin ce "ents, as well as intemal events e im Iready 1 at the plant

. any' isted with the potential design altemative None of th hav st-benefit ratio of less than $5,000/ person-rem.

Howeve lyd ives that come within a decade of the $5,000/ person-rem -

sta re the dive Ives at $19,800/ person-rem and the self-actuating con ent isolation va ,700/ person-rem, as described in the following sections.

21

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1 234

3,7.3.1 Diverse IRWST Injection Valves j in the current AP600 design, a squib valve in series with a check valve isolates each of four IRWST injection paths. This design alternative would reduce the likelihood of common cause failures of IRWST injection to the reactor by utilizing diverse valves in two of the four paths. If it l functioned perfectly, this design attemative could potentially reduce the CDF by about 72-percent.

When taking into account extemal events, other accident sequences not yet included in the AP600 ,

PRA, and other uncertainties, this design attemative is estimated to be highly costiiffective.' In the I absence of a comprehensive extemal events PRA for the AP600 plant, difficistito estimate the i effectiveness of this design attemative in reducing the risk from such owever,it appears likely that failure to inject coolant to the reactor would remaiuM> rom' or to the CDF from extemal events, in which case, diversity in the IRW 75ction eip to red "

the risk from both extemal and internal events. sf Alternative vendors are available for the check valves, questionab eck valves from different vendors would be sufficiently differ nsidered dive ss the i type of check valve was changed from the current swing galve to another. type. The swing disk type is preferred for this application, and oth 1 idered less reliable. i lll I l Adding diversity to the injection line squib valves uire a l res at the plant, and l some additional training for plant operations a significantly to the operational aspects of the, I rea

('

pce staff,l id not appear to add issue concems the l

h a g#r. Squib valves are availability and costs of acquiring diverse s frog '

specialized valve designs for which ther few veWdor3l .

ouse claims that a vendor i may not be willing to design, qualify, uild aj%sonable valve for this AP600 application considering that they would only o va per pl / The cost estimate for this design altemative assumes that a secon valv ndor e and that the vendor would provide  !

only the two diverse IRWST ives. st i does not include the additional first- 1 time engin ' g and qual festing ncurred by the second vendor. l (Westing imat se costs . , ore than a million dollars.) As a result, Westing d dissign a ive would not be practicable because of the uncertain of amig gaguib valve design / vendor and because of the uncertainty in reliability of a che5 e staff considers the rationale set forth by Westinghouse pote ions in reliability and high costs associated with obtaining d rea . e. On the bases of these arguments, the staff concludes that this n altem t be further pursued.

3.7. If-Actuating ent isolation Valves sign attemativ uld reduce the likelihood of containment isolation failure by adding uating valves nhancing the existing containment isolation valves for automatic closure inmen itions indicate that a severe accident has occurred. Conceptually, the er an independent valve or an appendage to an existing' fail-closed valve that post-accident containment conditions within containment. For example, a fusible link in response to elevated ambient temperatures, thereby providing the self-actuating

. function to vent the air operator of a fail-closed valve. This design attemative is estimated to impact releases from containment by only 10-percent. It has a cost-benefit ratio of

$33,000/ person-rem, and achieves this ratio primarily because of its low capital costs.

23

This improvement to the containment isolation capability would appear to be effective in reducing offsite releases for accidents involving either external or internal events. Also, the effectiveness of this design altemative would not be affected by the design changes made as a result of the AP600 PRA. a The addition of this design attemative would impose minor operational disadvantages to the plant in that the operations and maintenance staff would require some additional training 3 1n addition, these automatic features would require periodic testing to ensure that they are functioning properly.

A A Perhaps the biggest question regarding this design attemativetis wh _ can be implemented for a cost of only $33,000. The cost estimate does not ap%e7to includisthifirihime engineerBg and qualification testing that would be required to demonstiate that theNaide" ^ rform,its#

'testir and intended maintenancefunction also do notin a timely appear to haveand beenreliable included ^manner.

' The d65ts assoclifed with"ta#

this design altemative would be substantially higher tha ;Westi6ghouse's estimatS'(iiEiliaps by a factor of 10) when all related costs are realistically consid5riid30kthe basis of the unfavorable cost-benefit ratio, and the expectation that actual costs.w6Sff befde'n. higher than estimated by Westinghouse, the staff concludes that this design alteWiativeliitot' licht-beneficial and need not be further evaluated. 3.8 Conclusions T As discussed in Section 19.1 of AP600 GSER, We h ively used the PRA results to arrive at a final AP600 design. As a result, the esdinated CDF5n'd risk calculated for the AP600 plant are very low both relativ's id operatiig plants nd in absolute terms. Moreover, the low CDF and risk for the AP600 defsiffreflectMestingho0de's efforts to systematically minimize the effect of initiators / sequences lNist have eb' iin importidt contributors to CDF in previous PWR PRAs. West.inghouse has achieniId this obj$ctiverlai6Eiy by incorporating a number of hardware improvementsikin,the AP606'ddslin. Theseliihd Aihir AP600 design features which contribute to j low CDF4 nil Asiffor the'A " "dssign are 2fliciissed in Section 19.1 of the AP600 FSER.  ! Because the

                            'gn aire h

contaigs numerous plant features oriented toward reducing i CDF and risk, t 'tiesiefitsand riskied6cti'on potential of additional plant improvements are significantly ra4MMhiG true fo'EM6th intemally and externally initiated events. Moreover, with the featuqai6fready Iri .in the AP600 design, the ability to estimate CDF and risk approaciies the limitati@onsof { bilistic techniques. Specifically, when CDFs of 1 in 100,000 or 1,00 ' years are estidiiI E a PRA, it is the area of the PRA where modeling is least . co 'te, or supporting diita is sparse or even nonexistent, that could actually be the more i imdichant contributors t(risk. Areas not modeled or incompletely modeled include human l regability, sabotage, rare initiating events, construction or design errors, and system interactions. Aglailtiih improvemairits in the modeling of these areas may introduce additional contributors to . CEN$end ff does not expect that additional contributions would change anything in , atiticlul'oisimik I w I l 1 24 l l

 .                                                                                                                                    i The staff concurs with Westinghouse's conclusion that none of the potential design modifications evaluated are justified on the basis of cost-benefit considerations. The staff further concludes that it is unlikely that any other design changes would be justified on the basis of person-rem exposure considerations, because the estimated CDFs would remain very low on an absolute scale,                                            j 4.0 THE ENVIRONMENTAL IMPACT OF THE PROPOSED ACTION issuing an amendment to 10 CFR Part 52 certifying the AP600 design would not constitute a                                         i significant environmental impact. The amendment would merely codifygresdliof the NRC's review and approval of the AP600 design p: defined in the FSER, datan.gsber 1998 (NUREG-1512). Further, because the amendment is a ru                                , e arepteecerces involved that would have alternative uses.                                                                          S$4 * .                 ?

As described in Section 3 of this EA, the NRC reviewed rnativesto the de ' fE' "1 L j rulemaking and attemativre design features related to p mitigating 9 h nts. j Consideration of alternatives under NEPA was necessa ow that the des 7 certification rule is the appropriate cc,urse of action, and (2) to ensure' ign codified in the certification rule would not exclude any cost beneficial desig ' elated to the prevention and mitigation of severe accidents. The NRC conclud (that t s to design certification did not provde for resolution of issues he prcC '~-n certification a rulemaking. j This design certificatioa rulemaking is in g wi . _' 's intent in the

   " Standardization and Severe Accident               '

State# nts ' R Part 52, to make future , plants safer than the current generat nts,t ieve ' solution of licensing issues, and i to enhance the safety benefits of s izat Throug s own independent analysis, the NRC l also concludes that Westinghous "uate nsider n appropriate set of SAMDAs, and none were found to be cost-bene. l. Alth no c SAMDAs,'Wes,tinghouse basis of t esults U p.2 of th%.[. y%*leJ(;n chang lirii$dy incor; ,

                                                                      . ents in features in the AP600 design on the examples of these features. These design f                  e to      . g

• nt prevehtion and mitigation, but were not considered in the SAMDA e .cause? (m __Iready part of the design. See FSER Section 19.1.6, "Use of PRA in th ess.' " " - - - '

Finally, the tself . Id not authorize the siting, construction, or operation of an AP600 d ~~ nuclea t. The issuance of a CP, ESP, COL, or OL for the AP600 design will rec, a prospectiv  % to address the environmental impacts of constructiori and oper at a specific sitl at time, the NRC will evaluate the environmental impacts and iss ' in EIS in accorda ' p th NEPA. The SAMDA analysis for the AP600, however, has b9en ted as part of th LEA and will not need to be reevaluated as part of an EIS related to siting, ion, or ope . 25

m 5.0 AGENCIES AND PERSONS CONSULTED,' AND SOURCES USED The sources for this EA include Westinghouse's "AP600 Standard Safety Analysis Report," as amended, August 19,1998; and the NRC's " Final Safety Evaluation Report Related to the Certification of the AP600 Standard Design" (NUREG-1512, Volumes 1,2 and 3), September 1998. The Director, Office of Nuclear Reactor Regulation (NRR), has determined underdNational Environmental Policy Act of 1969, as amended, and the NRC's regulati[ ' 'FR Part 51, Subpart A, that this rule is not a major Federal action signifi tly aff ualityof the human environment, and therefore, an EIS not required. sisf mination, as documented in this final EA, is that the amendment to 10 'Part 52 ~ thorize th siting, construction, or operation of a facility using the A 00 desig t AP600 design in a rule. Therefore, the NRC staff did n psue t for co, al, State, and local agencies. However, the NRC's finding cant environ et was published in the Federal Register on XXX XX.1999, t the proposed A esign

     - certification rule and there were no comments received r                     EA. The NRC will evaluate the environmental impacts and issue an EIS as appr                             ce with NEPA as part of the application (s) for the siting, construction, or operat    of a The Director of NRR finds that Westinghouse's that there is reasonable assurance that an arg=,. __ _

hovides ent basis to conclude

                                                                       '2FR    P       2 certifying the AP600
                                                                         ~ J4 referencing the certified design design   that will wouldnot  haveexclude        a severe been cost-bonef,ig   had it accident en as!I        de'elsri       k[iiF-s part of the  original' design certification application. The evaluatio ' thes          ues u          PA is considered resolved for the AP600 design.                        -
                                   ,k w

l "M ,l_. k we j 3 26

F: l l l l l l 1 l l J CONGRESSIONAL LETTERS 1 i l I Attachment 3

l [ g6 i The Honorable Joe L. Barton, Chairman Subcommittee on Energy and Power j

    .' Committee on Commerce                                                                               '

United States House of Representatives Washington, DC 20515 i l Dear Mr. Chairman. The Nuclear Regulatory Commission (NRC) has sent to the Office of the Federal Register the  ; enclosed Federal Register Notice in which it proposes to amend the regulations for licensing

commercial nuclear power plants (10 CFR Part 52). If the Commission decides to issue this s rule in final form, it will certify the AP600 standard plant design, which was submitted to the NRC l for its review by the Westinghouse Electric Company.

This proposed design certification rule is necessary to partially fultill the objectives of 10 CFR I Part 52, which are to provide licensing stability, early resolution oflicensing issues, and to enhance the safety and reliability of nuclear power plants through standardization. Those wishing to obtain a license to build or operate the AP600 design will be able to do so by 3 referencing the AP600 design ceitification rule. 4 Sincerely, j Dennis K. Rathbun, Director Office of Congressional Affairs

Enclosure:

Federal Register Notice cc: Representative Ralph Hall l l 2

w x The Honorable James M. Inhofe, Chairman Subcommittee on Clean Air, Wetlands, Private Property and Nuclear Safety Committee on Environment and Public Works United States Senate Washington, DC 20510

Dear Mr. Chairman:

The Nuclear Regulatory Commission (NRC) has sent to the Office of the Federal Register the enclosed Federal Register Notice in which it proposes to amend the regulations for licensing commercial nuclear power plants (10 CFR Part 52). If the Commission decides to issue this rule in final form, it will certify the AP600 standard plant design, which was submitted to the NRC for its review by the Westinghouse Electric Company. This proposed design certification rule is necessary to partially fulfill the objectives of 10 CFR Part 52, which are to provide licensing stability, early resolution of licensing issues, and to enhance the safety and reliability of nuclear power plants through standardization. Those wishing to obtain a license to build or operate the AP600 design will be able to do so by referencing the AP600 design certification rule. Sincerely, Dennis K. Rathbun, Director Office of Congressional Affairs

Enclosure:

Federal Register Notice cc: Senator Bob Graham

h.. NRC PROPOSES TO CERTIFY WESTINGHOUSE'S AP600 DESIGN The Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to certify the AP600 standard plant design developed by the Westinghouse Electric Company. Interested persons are invited to submit comments or to request an informal hearing before an NRC Atomic Safety and Licensing Board. No application for a license using the AP600 design has been filed with the NRC,.and issuance of this regulation would not authorize construction of any specific new nuclear power plant. However, if the Commission decides to issue this rule in final form and, thereby, certify the AP600 design, a utility that wishes to build and operate a new nuclear power plant could choose to use the AP600 design and reference it in an application for a license. Safety issues within the scope of the certified design would then not be subject to litigation, although site-specific environmental impacts associated with building and operating the plant at a particular location would be litigable. The NRC staff issued a Final Safety Evaluation Report (FSER), NUREG-1512, " Final Safety Evaluation Report Related to Certification of the AP600 Standard Design," and a final design approval for the AP600 design on September 3,1998 (63 FR 48772). The NRC staff reviewed Westinghouse's application for compliance 'with the applicable portions of the Commission's current regulations and determined that the AP600 design should be exempt from six regulations (refer to Section V of the proposed rule). If the Commission decides to issue a final rule certifying the AP600 design, it will be valid for 15 years. Further details on the proposed design certification rule are provided in a Federal Register Notice that was published on (( "i . The public is invited to submit comments on the proposed design certification rule, the AP600 design control document (DCD) Attachment 4

.a 9. submitted by Westinghouse and incorporated by reference into the rule, and the environmental assessment for the AP600 design. In addition, interested parties may also request an informal l hearing before an NRC Atomic Safety and Licensing Board on matters related to this proposed design certification rule. The comments and requests for an informal hearing must be submitted, withing 75 days of the Federal Register Notice, to the Secretary, U.S. Nuclear l Regulatory Commission, Washington, DC 20555-0001, Attention: Rulemakings and Adjudications Staff, Mail Stop O-16 C1. Copies of comments received, the DCD, and the environmental assessment will be available for examination and copying at the NRC Public Document Room at 2120 L Street, NW, (Lower Level), Washington, DC. A copy of the FSER may be obtained from the Superintendent of Documents, U. S. Govemment Printing Office,  ; P.O. Box 37082, Washington, DC 20402-9328 or the National Technical information Service, Springfield, VA 22161-0002. I 2 Attachment 4

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