nrc review of electric power research institute's advanced light

WREG--1242-V01.1

TI92 040789

NRC Review of Electric Power Research Institute’s Advanced Light Water Reactor Utility Requirements Document Program Summary

Projeci; Number 669 - Manuscript Completed: August 1992 Date Put lished: August 1992

Associate Directorate for Advanced Reactors and License Renewal Office alf Nuclear Reactor Regulation US. Nuclear Regulatory Commission Washington, DC 20555

% msmiwnm OF THIS DOCUMENT IS UNLIMITED

DISCLAIMER

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency Thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

DISCLAIMER Portions of this document may be illegible in electronic image products. Images are produced from the best available original document.

ABSTRACT

The E l e c t r i c Power Research I n s t i t u t e ( E P R I ) i s p repar ing a compendium o f t e c h n i c a l requi rements, r e f e r r e d t o as the "Advanced L i g h t Water Reactor [ALgR] U t i l i t y Requirements Document," t h a t i s a p p l i c a b l e t o t h e des ign o f an ALWR power p l a n t . When completed, t h i s document i s in tended t o be a comprehensive statement o f u t i l i t y requi rements f o r t h e des ign, c o n s t r u c t i o n , and perFormance o f an ALWR power p l a n t f o r t h e 1990s and beyond.

The Requirements Document c o n s i s t s o f t h r e e volumes. Volume I, "ALWR P o l i c y and Summary o f Top-Tier Requirements," i s a management-level synopsis o f t h e Requirements Document, i n c l u d i n g t h e des ign o b j e c t i v e s and phi losophy, t h e

\ o v e r a l l phys i ca l c o n f i g u r a t i o n and fea tu res o f a f u t u r e nuc lea r p l a n t des ign, and t h e s teps necessary t o take t h e proposed ALWR des ign c r i t e r i a beyond t h e conceptual des ign s t a t e t o a completed, f u n c t i o n i n g power p l a n t . Volume I1 con: ; is ts o f 13 chapters and con ta ins u t i l i t y des ign requi rements f o r an evolu- t i o r i a r y nuc lea r power p l a n t [approx imate ly 1350 megawat ts -e lec t r i c (MWe)]. Volume 111 con ta ins u t i l i t y des ign requi rements f o r nuc lea r p l a n t s f o r which pas:,ive f e a t u r e s w i l l be used i n t h e i r des igns (approx imate ly 600 MWe).

The s t a f f o f t h e O f f i c e o f Nuclear Reactor Regulat ion, U . S . Nuclear Regulatory Commission, has prepared Volumes 1 and 2 (Par ts 1 and 2) o f i t s s a f e t y e v a l u a t i o n r e p o r t (SER) t o document t h e r e s u l t s o f i t s rev iew o f Volume I and I 1 c f t h e Requirements Document. Volume 1, "NRC Review o f E l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor U t i l i t y Requirements Document - Program Summary," p rov ides a d i scuss ion o f t h e o v e r a l l purpose and scope o f t h e Requirements Document, t he background o f t h e s t a f f ' s rev iew, t h e rev iew approach used by t h e s t a f f , and a summary o f t h e p o l i c y and t e c h n i c a l i ssues r a i s e d by t h e s t a f f d u r i n g i t s rev iew. E l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor U t i l i t y Requirements Document - Evo lu t i ona ry P l a n t Designs," g i v e s t h e r e s u l t s s t a f f ' s rev iew o f t h e 13 chapters o f t h e Requirements Document f o r e v o l u t i o n - a ry p l a n t designs. Volume 3 , "NRC Review o f E l e c t r i c Power Research I n s t i - t u t e ' s Advanced L i g h t Water Reactor Requirements Document - Passive P1 an t Designs," scheduled t o be issued i n September 1993, w i l l g i v e t h e r e s u l t s o f t h e s t a f f ' s rev iew o f t h e 13 chapters o f t h e Requirements Document f o r pass ive p l a n t designs.

Volume 2, "NRC Review o f

o f t h e

P r e l i m i n a r y d r a f t s o f Volumes 1 and 2 were forwarded t o t h e Commission and t h e A d v i j o r y Committee on Reactor Safeguards (ACRS) on May 12, 1992. 1ettl.r dated A p r i l 24, 1992, t h e s t a f f i ssued a d r a f t o f Volume 3 on a l l o f t h e chapters o f t h e Requirements Document f o r pass ive p l a n t des igns. A f t e r t h e ; t a f f has completed i t s rev iew o f EPRI's responses t o t h e d r a f t SER (DSER) on p a s s i v e p l a n t des igns i n t h e form o f r e v i s i o n s t o t h e Requirements Docu- ment , i t w i l l i s sue a f i n a l SER t o d iscuss i t s conc lus ions rega rd ing i t s rev iew o f t h e f i n a l ve rs ion o f t h e document.

I n i t s

I n s t a f f requi rements memoranda (SRM), t h e Commission i n s t r u c t e d t h e s t a f f t o p r o v - d e an a n a l y s i s d e t a i l i n g where t h e s t a f f proposes depar tu re f rom c u r r e n t r e g u l a t i o n s o r where t h e s t a f f i s s u b s t a n t i a l l y supplement ing o r r e v i s i n g

Program Summary iii

interpretive guidance applied to currently licensed LWRs. these to be policy issues. gives the staff's regulatory analysis of those issues identified for the evolutionary plant designs. Requirements Document for passive plant designs gives the regulatory analysis of those issues identified for the passive plant designs. been addressed in Commission papers SECY-90-016, "Evolutionary Light Water Reactor Certification Issues and Their Relationship to Current Regulatory Requirements"; SECY-91-078, "Chapter 11 of the Electric Power Research Insti- tute's Requirements Document and Additional Evolutionary Light Water Reactor Certification Issues"; and in draft Commission papers, "Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirements, and "Design Certification and Licensing Pol icy Issues Pertaining to Passive and Evolutionary Advanced Light Water Reactor Designs," that were issued on February 27 and July 6, 1992, respectively.

The staff considers Appendix B to Chapter 1 of Volume 2 of this report

Appendix B to the DSER on Chapter 1 of the

These issues have

In SRM dated June 26, 1990, and April 1, 1991, the Commission provided its decisions on SECY-90-016 and SECY-91-078 as they apply to evolutionary designs. The Commission will be reviewing the basis for the approach that the staff is proposing for those issues discussed in the draft Commission papers of February 27 and July 6, 1992, and, accordingly, may at some future point in the review determine that such issues involve policy questions that the Commission may wish t o consider. These issues are considered fundamental to agency decisions on the acceptability of the ALWR designs. ensure satisfactory implementation of Commission guidance regarding these matters during its review o f individual applications for final design approval and design certification.

There are no open issues pertain ng to the Requirements Document for evolutionary plant designs other than policy issues on which the staff has taken a position, but for which the Comm ssion has not had the opportunity to provide guidance. In addition, the staff concludes that there are issues that must be satisfactorily resolved before it can complete its review of the Requirements Document for passive plant designs. These issues are summarized in Section 4 of Volume 1 and discussed in detail in this report.

The staff will

Program Summary iv

.

TABLE OF CONTENTS

Paqe

ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

1 [NTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

. !.l Background and Review Sta tus . . . . . . . . . . . . . . . . . 1-1 :!.2 Purpose and Regulatory S ta tus o f EPRI’s ALWR U t i l i t y

Requirements Document . . . . . . . . . . . . . . . . . . . . . 1-2 ... 3 EPRI’s P o l i c y Statements . . . . . . . . . . . . . . . . . . . 1-3 1.4 ALWR Design Bases . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.5 Regulatory S t a b i l i z a t i o n . . . . . . . . . . . . . . . . . . . 1-4 1.6 NRC Review C r i t e r i a . . . . . . . . . . . . . . . . . . . . . . 1-4 1 .7 Format and A v a i l a b i l i t y o f Documentation . . . . . . . . . . . 1-7

2 I’OLICY ISSUES . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

3 F:EVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS . . . . . . . 3-1

4 CUTSTANDING ISSUES . . . . . . . . . . . . . . . . . . . . . . . . . 4-1

4 . 1 Outstanding Issues P e r t a i n i n g t o t h e Requirements Document f o r

4.2 Outstanding Issues P e r t a i n i n g t o t h e Requirements Document f o r E v o l u t i o n a r y P l a n t Designs . . . . . . . . . . . . . . . . . . 4-1

Passive P l a n t Designs . . . . . . . . . . . . . . . . . . . . . 4-2

5 VENDOR- OR UTILITY-SPECIFIC ITEMS . . . . . . . . . . . . . . . . . 5-1

5.1 Vendor- o r U t i l i t y - S p e c i f i c I tems P e r t a i n i n g t o t h e Requirements Document f o r Evo lu t i ona ry P l a n t Designs . . . . . . . . . . . . 5-1

5.2 Vendor- o r U t i 1 i t y - S p e c i f i c I tems P e r t a i n i n g t o t h e Requirements Document . for Passive P l a n t Designs . . . . . . . . . . . . . . 5-21

6 CONCLUSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1

APPEllDIX A CHRONOLOGY OF CORRESPONDENCE . . . . . . . . . . . . . . . A-1

APPElilDIX B REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . B-1

APPEllDIX C LIST OF ABBREVIATIONS . . . . . . . . . . . . . . . . . . . C - 1

APPEEIDIX D PRINCIPAL CONTRIBUTORS . . . . . . . . . . . . . . . . . . D-1

APPENDIX E COMMISSION PAPERS APPLICABLE TO ADVANCED LIGHT WATER REACTORS . . . . . . . . . . . . . . . . . . . . . . . . . E-1

APPEPIDIX F REPORT BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS . . F-1

Program Summary V

1 1 NTRODUCTION

The E l e c t r i c Power Research I n s t i t u t e (EPRI) i s p repar ing a compendium o f t e c h i c a l requirements, r e f e r r e d t o as t h e "Advanced L i g h t Water Reactor [ALWR] U t i l i t y Requirements Document," t h a t i s a p p l i c a b l e t o t h e des ign o f an ALWF: power p l a n t . When completed, t h i s document i s in tended t o be a compre- henzive statement o f u t i l i t y requirements f o r t h e design, c o n s t r u c t i o n , and perf'ormance o f an ALWR power p l a n t f o r t h e 1990s and beyond. Those p a r t i c i - p a t i n g i n t h e program i n c l u d e u t i l i t i e s w i t h nuc lea r p l a n t exper ience, nuc lea r steam supply system vendors, a rch i tec t -eng inee r ing f i r m s , and consu l tan ts i n re1 a t e d f i e l d s .

The Requirements Document c o n s i s t s o f t h ree volumes. Volume I, "ALWR P o l i c y and Summary o f Top-Tier Requirements," i s a management-level synopsis o f t h e Requirements Document, i n c l u d i n g t h e des ign o b j e c t i v e s and phi losophy, t h e o v e r a l l phys i ca l c o n f i g u r a t i o n and fea tu res o f a f u t u r e nuc lea r p l a n t design, and t h e s teps necessary t o take t h e proposed ALWR design c r i t e r i a beyond t h e conceptual des ign s t a t e t o a completed, f u n c t i o n i n g power p l a n t . c o n s i s t s o f 13 chapters and conta ins u t i l i t y des ign requi rements f o r an e v o l u t i o n a r y nuc lea r power p l a n t [approx imate ly 1350 megawat ts -e lec t r i c (MWe)]. (approx imate ly 600 MWe) i n which pass ive fea tu res w i l l be used f o r t h e u l t i m a t e s a f e t y p r o t e c t i o n o f t h e p l a n t .

Volume I 1

Volume I 1 1 con ta ins u t i l i t y des ign requi rements f o r nuc lea r p l a n t s

The s t a f f o f t h e O f f i c e o f Nuclear Reactor Regulat ion, U.S. Nuclear Regulatory Commission (NRC), has prepared Volumes 1 and 2 (Pa r t s 1 and 2 ) o f i t s s a f e t y e v a l u a t i o n r e p o r t (SER) t o document the r e s u l t s o f i t s rev iew o f Volume I and I 1 o f t h e Requirements Document. Volume 1, "NRC Review o f E l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor U t i l i t y Requirements Docuinent - Program Summary," p rov ides a d i scuss ion o f t h e o v e r a l l purpose and scope o f t h e Requirements Document, t he background o f t h e s t a f f ' s rev iew, t h e rev iew approach used by t h e s t a f f , and a summary o f t h e p o l i c y and t e c h n i c a l i ssues r a i s e d by t h e s t a f f d u r i n g i t s rev iew. E l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor U t i l i t y Requirements Document - Evo lu t i ona ry P l a n t Designs," g i ves t h e r e s u l t s o f t h e s t a f f ' s rev iew o f t h e 13 chapters o f t h e Requirements Document f o r e v o l u t i o n - a r y p l a n t designs. Volume 3, "NRC Review o f E l e c t r i c Power Research I n s t i - t u t e ' s Advanced L i g h t Water Reactor U t i l i t y Requirements Document - Passive P l a n t Designs," scheduled t o be issued i n September 1993, w i l l g i v e t h e r e s u l t s o f t h e s t a f f ' s rev iew o f t h e 13 chapters o f t h e Requirements Document f o r i a s s i v e p l a n t designs.

Volume 2 , "NRC Review o f

1.1 Backqround and Review Sta tus

I n 1983, E P R I began i t s program by work ing w i t h t h e NRC s t a f f t o i d e n t i f y and r e s o l v e key s a f e t y and l i c e n s i n g issues. process whereby t h e unresolved and gener i c s a f e t y issues a p p l i c a b l e t o ALWRs

Th is j o i n t e f f o r t r e s u l t e d i n a

Program Summary 1-1

as of July 1, 1986* were identified. procedures described in NUREG-0933, "A Prioritization of Generic Safety Issues." Additional information about this effort and its results is provided i n NUREG-1197, Advanced L i g h t Water Reac tor Program - Program ,Management and Staff Review Methodology," dated December 1986. In 1985, two new phases were added to the EPRI program: ments Document for evolutionary plants and the assessment of small-plant options. Document for passive plant designs.

This process was consistent with the

the development of EPRI's ALWR Utility Require-

This assessment resulted in the development o f the Requirements

Chronoloqy of Review of Reauirements Document for Evolutionary Plant Desiqns

From June 30, 1986, through October 26, 1989, EPRI submitted Revision 0 of the Requirements Document for evolutionary plant designs. 1987, through November 4, 1991, the staff developed and issued its draft SERs (DSERs) on these submittals. were issued.

From September 24,

Table 1.1 gives the dates when these documents

On September 7, 1990, EPRI submitted Revision 1 of the Requirements Document for evolutionary plant designs, modifying the document in its entirety. April 26 and November 25, 1991, and April 17, 1992, EPRI submitted Revi- sions 2, 3 , and 4 , respectively. Volume 2 o f this report addresses the Evolutionary Document for evolutionary plant designs through Revision 3. Where possible, the staff's review included consideration of Revision 4 of the Requirements document.

On

A preliminary draft of Volume 2 was forwarded to the Commission and the Advisory Committee on Reactor Safeguards (ACRS) on May 12, 1992. The staff discussed the contents of the SER with the Committee and has included the views o f the ACRS in Section 3 of this report.

Chronoloqy of Review of Reauirements Document for Passive Plant Desiqns

In its letter dated September 7, 1990, EPRI submitted Revision 0 of the Requirements Document for passive plant designs. On April 26, 1991, and January 2, 1992, EPRI submitted Revisions 1 and 2, respectively. In its letter dated April 24, 1992, the staff issued the DSER on all of the chapters of the Requirements Document for passive plant designs. completed its review of EPRI's responses to this DSER in the form of revisions to the Requirements Document, it will issue a final SER as Volume 3 of this report to provide its conclusions regarding its review of the final version of the Requirements Document.

After the staff has

1.2 PurDose and Requlatorv Status of EPRI's ALWR Utility Reauirements Document

EPRI's ALWR Utility Requirements Document is designed to serve as a vehicle to obtain consistent resolution of common operating problems, issues generically applicable to designs, severe-accident issues, and certain unresolved and generic safety issues. such as utility procurement specifications, that cover the remaining technical

The document is to be used with companion documents,

*This date has since been changed to January 1990.

Program Summary 1-2

requirements applicable to new plant projects. as a vehicle to identify early in the design process major concerns about de!,ign concepts for LWRs in which passive safety systems will be used.

It is also designed to serve

EPFlI ’ s ALWR Uti 1 i ty Requirements Document , because it is an agreement between th(1 vendors and the nuclear power utilities, identifies what utilities desire in future designs. The Requirements Document has no legal or regulatory stztus. Conimission’s regulations, regulatory guidance, or policies, nor is it intended to be used as a basis for supporting final design approval and design certification (FDA/DC) for a specific design.

It is not intended to demonstrate complete compliance with the

- Conimi ssion Guidance

In its staff requirements memorandum (SRM) dated December 15, 1989, the Conmission assigned the review of the Requirements Document for evolutionary plant designs priority equal to those of General Electric Company’s Advanced Boiling Water Reactor and Combustion Engineering, Inc.’s System 80t. In adcition, the Commission directed the staff to compare future designs against thc Requirements Document for evolutionary plant designs.

In the same SRM, the Commission directed the staff to complete its review of the Requirements Document for passive plant designs before it submitted the results o f its review of the licensing review basis (LRB) for passive designs to the ACRS. Although a design certification applicant is no longer required to submit an LRB because the promulgation of Part 52 of Title 10 of the Q& - of Federal Resulations (10 CFR Part $ 2 ) negated the need for such a document, the staff interprets this guidance as directing it to complete its review of the Requirements Document for passive plant designs before significant interaction with the ACRS begins on these designs. that, with the issuance of the DSER on Volume 3 in April 1992, this has been accompl ished.

It is the staff’s position

In its SRM o f June 22, 1990, the Commission directed the staff to formally resolve major technical and policy issues in the context of the review of the Requirements Document for passive plant designs. The staff has identified such issues during its review o f Volume I11 o f the ALWR Utility Requirements Document and conceptual design information on the passive ALWRs. The staff developed the draft Commission papers dated February 27 and July 6, 1992, to address resolution of these issues. policy issues, the staff will continue to evaluate resolutions proposed by EPAI and the ALWR vendors, and will address them in future Commission papers.

To ensure timely resolution of these

1.3 EPRI’s Policy Statements

The ALWR Uti1 ity Steering Committee established policies to provide guidance for the overall development of the EPRI’s ALWR Utility Requirements Document and to provide the plant designer with guidance in applying the design criteria. Although not design criteria themselves, these policies cover fundamental principles that have a broad influence on the design criteria of the Requirements Document. These policies include consideration of simplification, design margin, human factors, safety, regulatory stabilization,

Program Summary 1-3

standardization, use of proven technology, maintainability, constructibility, quality assurance, economics, protection against sabotage, and environmental effects.

1.4 ALWR Desiqn Bases

The term "ALWR design bases," as defined by EPRI, refers to the three sets of requirements that form the foundation for the ALWR design criteria. The first set of requirements forms the "licensing design basis," which includes the requirements necessary to satisfy regulatory criteria. These requirements and associated analytical methods are based on conservative, NRC-approved methods, and equipment is designed to safety-grade standards. "risk-evaluation-basis," which extends the licensing design basis t o meet public safety objectives. risk evaluation basis methods. The third set is the "performance design basis," which is based on economic and investment protection considerations for a utility and for which realistic, designer-selected, best-estimate methodology is used. to provide an adequate level of safety, whereas the risk evaluation and performance design bases provide additional or enhanced protection.

The second set is the

Probabilistic risk assessments are used for the

EPRI states that the licensing design basis is intended

1.5 Requlatorv Stabilization

Consistent with the overall ALWR program approach, as described in NUREG-1197, regulatory stabilization for an ALWR design can be achieved through the identification and resolution of plant optimization subjects and generic safety and licensing issues.

Plant optimization subjects are proposals, initiated by EPRI, to deviate from regul atory requirements. EPRI proposes to resolve these issues by providing technically supportable alternatives to current regulatory requirements. Table 1.2 contains a list of EPRI's proposed plant optimization subjects and their applicability to the evolutionary and passive plant designs. issues are identified for both the evolutionary and passive plant design criteria in Section 2 of Appendix B to Chapter 1 of Volumes I1 and 111 of the Requirements Document. The staff's evaluation is provided in the corresponding sections of Volume 2 of this report and the April 1992 draft of Volume 3 of this report.

These

In addition, EPRI specifically addressed those generic safety issues that were classified as "high" or "medium" priority as of January 1990 in Section 3 of Appendix B to Chapter 1 of Volumes I1 and I11 of the Requirements Document. The staff's evaluation of the generic safety issues specifically addressed by EPRI and those unresolved safety issues and generic safety issues that are considered applicable to ALWRs is provided in the corresponding sections of Volume 2 of this report and the April 1992 draft of Volume 3 of this report.

1.6 NRC Review Criteria

Criteria Governins the Review of EPRI's ALWR Utility Requirements Document

The staff's review of the Requirements Document is being conducted as described in NUREG-3197. As noted therein, the staff is using NUREG-0800, "Standard Review Plan [SRP] for the Review of Safety Analysis Reports for

Program Summary 1-4

Nuclear Power P lan ts , " f o r rev iew guidance. I n a d d i t i o n t o t h e c r t e r i a o f NUREG-0800, t h e s t a f f ' s rev iew r e f l e c t s the requirements o f 10 CFR Par t 52, "Earmly S i t e Permits; Standard Design C e r t i f i c a t i o n s ; and Combined icenses f o r Nuclear Power Reactors," and t h e Commission's " P o l i c y Statement on Severe Reactor Acc idents Regarding Future Designs and E x i s t i n g P lan ts " (50 FR 32138, August 8, 1985) and i t s p o l i c y statement on "Safe ty Goals f o r t h e Operat ions o f huc lea r Power P lan ts " (51 FR 28044, August 4, 1986).

The Requirements Document p laces pr imary emphasis on p reven t ing s i g n i f i c a n t problems t h a t have been exper ienced i n e x i s t i n g p lan ts ; however, many d e t a i l s t h a t w i l l be p rov ided i n s p e c i f i c des ign a p p l i c a t i o n s are miss ing. t h e s t a f f i s rev iew ing t h e proposed requirements a t t h e l e v e l o f d e t a i l presented by E P R I b u t i s n o t de termin ing t h e i r adequacy t o meet a l l NRC r e q u i r e - ment s .

Therefore,

A l though t h e SRP i s be ing used as guidance, t he l e v e l o f d e t a i l does n o t p e r m i t a rev iew f o r completeness. The SRP was w r i t t e n t o suppor t t h e rev iew o f s a f e t y a n a l y s i s r e p o r t s on s p e c i f i c p l a n t designs f o r which a s i g n i f i c a n t amount o f des ign and c o n s t r u c t i o n i n fo rma t ion was a v a i l a b l e . Therefore, t h e s t a f f ' s e v a l u a t i o n became one o f " rev iew by except ion," and t h e s t a f f i s con- d u c t i n g i t s rev iew w i t h t h e understanding t h a t E P R I i n tends t h a t t he Require- ments Document c o n t a i n des ign c r i t e r i a t h a t meet a l l c u r r e n t r e g u l a t i o n s , except where d e v i a t i o n s are i d e n t i f i e d i n t h e document. The s t a f f ' s rev iew o f t h e Requirements Document i s focused p r i m a r i l y on de termin ing whether t h e E P R I c r i t e r i a do o r do n o t c o n f l i c t w i t h c u r r e n t r e g u l a t o r y requi rements. requi rement proposed by E P R I i s found n o t t o c o n f l i c t w i t h NRC requirements, t h e s t a f f judges i t t o be acceptable.

I f a

However, i n c e r t a i n t e c h n i c a l areas (e.g., advanced ins t rumen ta t i on and c o n t r o l design, human f a c t o r s cons ide ra t i ons ) , l i t t l e o r no requi rements o r guidance e x i s t s . assoc ia ted w i th these areas are u n l i k e l y t o c o n f l i c t w i t h r e g u l a t o r y r e q u i r e - ments. f i n d i n g o f a c c e p t a b i l i t y , t h e s t a f f deems i t approp r ia te t o app ly a d d i t i o n a l s t a n l a r d s t o judge t h e a c c e p t a b i l i t y o f a requi rement . I n such cases, t h e s t a f f i s us ing a combinat ion o f cons ide ra t i on o f those r e g u l a t i o n s t h a t do a p p l j , good eng ineer ing judgment, pas t p r a c t i c e and exper ience a t ope ra t i ng power p l a n t s , and a p p l i c a t i o n s i n o t h e r i n d u s t r i e s t o eva lua te t h e r e q u i r e - ment; proposed by E P R I i n t h e Requirements Document.

Therefore, des ign c r i t e r i a i n the Requirements Document

I n those cases, a l though us ing t h e c o n f l i c t s tandard would r e s u l t i n a

Dur ing i t s rev iew, t h e s t a f f a l s o i d e n t i f i e d p o t e n t i a l i n c o m p a t i b i l i t i e s between EPRI-proposed des ign requi rements and c u r r e n t regu l a t o r y requirements, and where p o s s i b l e m i s i n t e r p r e t a t i o n s o f r e g u l a t o r y requi rements e x i s t i n t h e Requirements Document.

I n t h e February 15, 1991, SRM p e r t a i n i n g t o SECY-90-377, "Requirements f o r Design C e r t i f i c a t i o n Under 10 CFR Par t 52," t h e Commission d i r e c t e d t h e s t a f f t o rev iew t h e Requirements Document t o ensure t h a t i t i s s u f f i c i e n t t o a l l o w t h e s t a f f t o eva lua te t h e severe-accident i ssues and t h e i n c o r p o r a t i o n o f exper ience f rom opera t i ng events i n c u r r e n t des igns. The s t a f f has used bo th d e t e r m i n i s t i c and p r o b a b i l i s t i c methods o f eva lua t i on , cons ide r ing how t h e Requ.irements Document addresses these issues through s p e c i f i c des ign c r i t e r i a and th rough t h e g u i d e l i n e s i t prov ides f o r per fo rming a p r o b a b i l i s t i c r i s k asse:;sment.

Program Summary 1-5

In addition to addressing such matters relative to safety, the staff also has provided constructive comments on the document that, while not specifically regulatory in nature, would offer improvements in its requirements. will review an actual ALWR plant in accordance with the most current version of the SRP and will follow the SRP criteria, except for those instances where the staff has specifically accepted other positions in EPRI’s ALWR Utility Requirements Document and those positions have been endorsed in the final SER for the ALWR program.

The staff

Finally, as discussed in Section 1.2, the staff’s review of the Requirements Document is not intended to substitute for any portion of the staff’s review of future applications for final design approval and design certification.

Additional Criteria Governinq the Review of the Reauirements Document for Passive Plant Desiqns

The licensing design-basis analysis of the Requirements Document for passive plant designs relies solely on the passive safety systems to demonstrate compliance with the acceptance criteria for various design-basis transients and accidents. Consequently, uncertainties remain concerning the performance of the unique passive features and overall performance of core and containment heat removal because of lack o f a proven operational performance history. For example, uncertainties exist about the performance of check valves in the passive safety systems, which operate at low differential pressures provided by natural circulation or gravity injection. These low pressures may not provide sufficient force to fully open sticking check valves (i.e., pumped emergency core cooling systems are more likely to overcome stuck valves). a result, these uncertainties enhance the importance of the active non-safety- related systems in providing the defense-in-depth protection to prevent and mitigate accidents and core damage. Therefore, the review of the passive designs requires a review of not only the passive safety systems, but also the functional capability and availability of the active non-safety-related systems to provide significant defense-in-depth protection and the capability to prevent accidents and core damage.

As

8 For those active systems that perform defense-in-depth functions, the Require- ments Document for passive plant designs specifies systems and equipment design and performance requirements. These include radiation shielding to permit access following an accident, the availability and redundancy of non- safety-related electric power, and protection against internal hazards, as well as safety analysis and testing required to demonstrate system capability for defense-in-depth considerations. However, the Requirements Document does not provide specific requirements pertaining to the reliability of these systems. EPRI has indicated that it is evaluating the specific reliability targets and other measures to ensure that the passive plants will meet performance requirements and that it will address these safety concerns for both passive safety and active non-safety-related systems.

In addition, technical specification development is a subset of the overall regulatory treatment of the passive designs. to establish re1 iabil ity-based technical specifications (TS) for the passive plant designs to determine which systems and components (including certain non-safety-related systems) will require the imposition of TS, and the

The staff is evaluating the need

Program Summary 1-6 I

parameters o f t h e TS ( l i m i t i n g c o n d i t i o n s f o r opera t ion , s u r v e i l l a n c e , e t c . ) . The r e l i a b i l i t y assurance program i s expected t o s t r o n g l y i n f l u e n c e TS.

Since t h e pass i ve ALWR des ign ph i losophy depar ts f rom c u r r e n t l i c e n s i n g p r x c t i c e s , t h e s t a f f has r a i s e d t h i s i ssue t o t h e Commission as a p o l i c y i ssue . The s t a f f has n o t completed i t s rev iew o f t h i s i ssue and, t h e r e f o r e , ha:; n o t p rov ided a recommendation t o t h e Commission. f u r t h e r i n t h e s t a f f ’ s r e g u l a t o r y depar tu re ana lys i s , which i s g i v e n i n Appendix B t o t h e DSER on Chapter 1 o f t h e Requirements Document f o r pass i ve p l an t designs.

T h i s i ssue i s discussed

1.7 Format and A v a i l a b i l i t y o f Documentation

Vo-ume 1 o f t h i s r e p o r t i s a program summary o f t h e NRC rev iew o f EPRI’s ALWR U t - l i t y Requirements Document. I t i s n o t in tended t o p r o v i d e a d i scuss ion o f Vo-hme I of t h e Requirements Document ( E P R I submitted Volume I f o r i n f o r m a t i o n on-iy and d i d n o t reques t t h e s t a f f t o rev iew i t ) . p r o v i d e a d i s c u s s i o n o f t h e o v e r a l l purpose and scope o f t h e Requirements Document and t h e r e s u l t s o f t h e s t a f f ’ s rev iew o f t h a t document. Sec t i on 6 g i \ Ies t h e s t a f f ’ s conc lus ions rega rd ing t h e rev iew o f t h e Requirements Docu- merit f o r t h e e v o l u t i o n a r y p l a n t designs and t h e s t a t u s o f t h e rev iew o f t h e Requirements Documents f o r t h e pass ive p l a n t designs. Appendix A i s a chro- no logy o f t h e correspondence r e l a t e d t o t h e rev iew o f t h e document. Appen- d i ) B c o n t a i n s t h e re fe rences f o r Volumes 1-3 o f t h i s r e p o r t . Appendix C i s a l i s t o f a b b r e v i a t i o n s used i n Volumes 1-3 o f t h i s r e p o r t . Appendix D i s a 1 i : t o f p r i n c i p a l c o n t r i b u t o r s . Appendix E i s a l i s t o f Commission papers t h z t a r e a p p l i c a b l e t o t h e s t a f f ’ s rev iew o f t h e ALWR a p p l i c a t i o n s f o r FDA/DC. Appendix F i s a copy o f t h e r e p o r t o f t h e Adv isory Committee Reactor Safe- guzrds on t h e ALWR U t i l i t y Requirements Document.

Sec t ions 1 th rough 5

Thc fo rmat o f Volume 2 and t h e A p r i l 1992 d r a f t o f Volume 3 f o l l o w s t h a t o f Volumes I 1 and I11 o f t h e Requirements Document as c l o s e l y as p o s s i b l e .

Copies o f Volumes 1 and 2 o f t h i s r e p o r t and t h e A p r i l 1992 d r a f t o f Volume 3 arc a v a i l a b l e f o r i n s p e c t i o n a t t h e NRC P u b l i c Document Room, 2120 L S t r e e t , N . k ’ . , Washington, DC 20555.

The NRC p r o j e c t managers f o r t h e s t a f f ’ s rev iew o f EPR1’s.ALWR U t i l i t y Rec,uirements Document a re J . H. Wi lson and T. J . Kenyon. They may be con tac t - ed by c a l l i n g (301) 504-1118 o r by w r i t i n g t o : Assoc ia te D i r e c t o r a t e f o r Advanced Reactors and License Renewal, U . S . Nuclear Regu la to ry Commission, Washington, DC 20555.

*NRl: documents (e.g., NUREG r e p o r t s and r e g u l a t o r y guides) a re n o t i n c l u d e d i n Appendix B because t h e y may be r e t r i e v e d as i n d i c a t e d i n t h e “ A v a i l a b i l i t y No t i ce “ on t h e i n s i d e f r o n t cover o f t h i s r e p o r t .

Program Summary 1-7

Table 1.1 Chronology of the Review o f EPRI’s Requirements Document for Evolutionary P l a n t Designs

~ ~~ ~

Chap- Date of Submittal Date DSER ter Title of Revision 0 Was issued

1

1A

1B

2

3

4

5

6

7

8

9

10

11

12

13

Overall Requirements

PRA Key Assumptions and Groundrul es

Licensing and Regul atory Requirements and Guidance

Power Generation Systems

Reactor Coolant System and Reactor Non-Safety Auxiliary Systems

Reactor Systems

Engineered Safety Systems

Bui 1 di ng Design and Arrange- men t

Fuel i ng and Refuel i ng Systems

P1 ant Cool ing Water Systems

Site Support Systems

Man-Machine Interface Systems

Electric Power Systems

Radioactive Waste Processing Systems

Main Turbine-Generator Systems

June 30, 1986 July 8, 1986

June 30, 1989 February 22, 1990

None issued

October 15, 1986

June 18, 1987 December 11, 1987

June 18, 1987

December 8, 1987

November 18, 1988

February 28, 1989

December 30, 1988

January 11, 1989

October 26, 1989

April 10, 1989

December 23, 1988

February 6, 1989

September 24, 1987 February 18, 1988

November 4, 1991

None issued

February 18, 1988

May 13, 1988

June 10, 1988

February 28, 1990

January 15, 1991

January 15, 1991

January 15, 1991

January 15, 1991

October 8, 1991

April 3, 1991

January 15, 1991

January 15, 1991

Program Summary 1-8

Table 1.2 EPRI-Proposed P lan t Op t im iza t i on Sub jec ts

Appl i cab i 1 i t y

P l i l n t Op t im iza t i on Subject ~ ~~

Evol u t i onary Passive

Opwat ing-Bas is Earthquake and Dynamic Ana lys i s Methods

X X

Tornado Design X X

B o i l i n g Water Reactor (BWR) Main Steamline X 1 so l a t i o n Valves and Leakage Cont ro l

Sinipl i f i c a t i o n o f Postacc ident Sampling !;ystem

Tyile C Containment Leakage Rate Tes t i ng 1 n t e r v a l

X

X X

X X

Source Term X X

Dec!icated Containment Vent Pene t ra t i on X X

M i t i g a t i o n o f A n t i c i p a t e d Trans ien ts k ' i t h o u t Scram f o r t h e Advanced BWR

Sinipl i f i c a t i o n o f O f f s i t e Emergency F ' l anni ng

X X

N /A X

Safe Shutdown o f Pass i ve ALWRs N/A X - N/A = not appl i c a b l e

Program Summary 1-9

1

2 POLICY ISSUES

In the staff requirements memorandum (SRM) dated August 24, 1989, the Commis- sion instructed the staff to provide an analysis detailing where the staff proposes departure from current regulations or where the staff is substantially supplementing or revising interpretive guidance applied to currently licensed light water reactors. Appendix 6 to Chapter 1 of Volume 2 of this report gives the staff's regulatory analysis of those issues identified for the evolutionary plant designs. Appendix 6 to the DSER on Chapter 1 of Volume 3 of the Requirements Document gives the regulatory analysis o f those issues identified for the passive plant designs. These issues have been addressed in Commi ssion papers SECY-90-016, "Evolutionary Light Water Reactor Certification Issues and Their Relationship to Current Regulatory Requirements," and SECY-91-078, "Chapter 1 1 of the Electric Power Research Institute's Requirements Document and Additional Evolutionary Light Water Reactor Certification Issues," and in draft Commis- sion papers, "Issues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirements," and "Design Certification and Licensing Policy Issues Pertaining to Passive and Evol utionary Advance Light Water Reactor Designs," that were issued on February 27 and July 6, 1992, respectively.

The staff considers these to be policy issues.

In its SRMs dated June 26, 1990, and August 15, 1991, the Commission provided its decisions on SECY-90-016 and SECY-91-078 as they apply to evolutionary desilps.

The -ebruary 27 and July 6, 1992, draft Commission papers have been forwarded to tlie Advisory Committee on Reactor Safeguards. views in the final paper and document its final positions before seeking Commission approval. reviewed by the Commission, and, therefore, do not represent agency positions.

These policy issues are considered fundamental to agency decisions on the acceptability of the ALWR designs. To aid in identifying its positions, the stafF underlined those for which it requested Commission approval in the Commission papers discussed above. Table 2.1 contains a list of these policy i s s u m and their applicability to the evolutionary and passive plant designs. Table 46.1 o f Appendix 6 to Chapter 1 of Volume 2 of this report lists the i s s w s that are applicable to the Requirements Document for evolutionary plant designs with the cross-reference to the chapters and sections in which they are discussed. Table 56.1 of Appendix 6 to Chapter 1 of the April 1992 draft of Volume 3 of this report lists those issues that are applicable to the Requirements Document for passive plant designs at the time of issuance along with the appropriate cross-references.

In addition, Appendix E lists the papers that the staff has forwarded to the Comm.tssion regarding policy issues that the staff has identified to date for both evolutionary and passive ALWRs. resuit of its review of the EPRI's ALWR Utility Requirements Document, the finai design approval and design certification applications for the evolutionary plants, and the conceptual design information on the passive plants.

The staff will include its

The approaches to resolving these issues have not been

The staff developed these papers as a

Program Summary 2- 1

Table 2.1 Policy Issues Pertaining to the Evolutionary and Passive P1 ant Designs

Appl i cabi 1 i ty Policy Issue Evolutionary Pass i ve

Use of physically based source term

Anticipated transients without scram

Mid-loop operation

Station blackout

Fire protection

Intersystem loss-of-coolant accident

Hydrogen control

Core-concrete interaction - Capability t o

High-pressure core melt ejection

Containment performance

Dedicated containment vent penetration

Equipment survivability

Elimination o f operating-basis earthquake

Inservice testing o f pumps and valves

Industry codes and standards

Electrical distribution

Sei smi c hazard curves

Leak before break

C1 assi f i cati on of mai n steam1 i ne of

cool core debris

boiling-water reactor

Containment bypass

X

X

x X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

N/A

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

N/A = not appl icable

Program Summary 2-2

1

Table 2.1 ; (Continued)


P o l - c y Issue Evo lu t i ona ry Passive

Tornado des ign bas i s X X

Containment bypass X X

Containment l e a k r a t e t e s t i n g X X

Posl.accident sampl ing system

Level o f d e t a i l

X

X

X

X

P r o t o t y p i ng X X

Inspec t ions , t e s t s , analyses, and acceptance c r i t e r i a

X

P r o t l a b i l i s t i c r i s k assessment beyond X dcmsign c e r t i f i c a t i o n

X

X

Re1 i abi 1 i t y assurance program X X

Sevcre-accident m i t i g a t i o n des ign a1 t e r n a t i v e s

X

Gencr ic ru lemaking r e l a t e d t o des ign X cc r t i f i c a t i o n

Regu la to ry t rea tment o f non-safety- N/A

D e f i n i t i o n o f pass ive f a i l u r e N/A

rcl ated systems

Thermal -hydraul i c s t a b i 1 i t y o f t h e N/A simp1 i f i e d b o i l i n g water r e a c t o r

X

X

X

Safe shutdown requi rements N/A X

Cont ro l room habi t a b i 1 i t y N/A X

Radionucl ide a t t e n u a t i o n N/A X

S i m p l i f i c a t i o n o f o f f s i t e emergency N/A p lann ing

N/A = n o t appl i c a b l e

Program Summary 2-3

Table 2.1 (Continued)


P o l i c y Issue Evol u t i onary Pass i ve

Defense aga ins t common-mode f a i l u r e s i n X d i g i t a l i ns t rumen ta t i on and c o n t r o l

Ana lys i s o f e x t e r n a l events beyond t h e X des ign bas i s

X

M u l t i p l e steam genera tor tubes rup tu res N/A X

Role o f t h e pass ive p l a n t c o n t r o l room N/A ope ra to r

X

Cont ro l room annunciator re1 i ab i 1 i t y X X

N/A = n o t appl i c a b l e

Program Summary 2-4

3 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

I n September 1984, t h e NRC s t a f f , EPRI , and the Adv isory Committee on Reactor Safeguards (ACRS) met t o d iscuss e a r l y e f f o r t s l e a d i n g t o t h e development o f EPRI’s ALWR Requirements Document. p e r i o d i c a l l y t o d i scuss t h e programmatic aspects and progress o f EPRI’s ALWR program.

Since t h a t t ime, these p a r t i e s have met

The ACRS met w i t h t h e s t a f f and EPRI numerous t imes d u r i n g t h e p e r i o d Octo- ber 1987 th rough January 1992 t o d iscuss t h e conten ts o f each chapter o f Volume I1 o f t h e ALWR Requirements Document and corresponding DSER a f t e r issuance o f t h e documents. Subcommittee on Improved L i g h t Water Reactors one o r more t imes.

I n a d d i t i o n , many o f t h e p o l i c y issues i d e n t i f i e d i n Appendix B t o Chapter 1 o f Volume 2 were d iscussed w i t h t h e Committee. t h e s t a f f ’ s responses t o t h e i r comments on these p o l i c y ma t te rs a re i n Annexes A-C o f Appendix B t o Chapter 1 o f Volume 2 o f t h i s r e p o r t .

Each chapter and DSER was d iscussed w i t h t h e ACRS

The r e p o r t s o f t h e ACRS and

The s t a f f met w i t h t h e ACRS Subcommittee on Improved L i g h t Water Reactors t o d i scuss t h e May 1992 d r a f t o f t h e SER on Volume I 1 o f t h e ALWR Requirements Docuinent on June 17 and 18, and J u l y 27, 1992. The s t a f f met w i t h t h e ACRS F u l l Committee t o d iscuss t h e r e s u l t s o f i t s rev iew o f Volume I 1 d u r i n g t h e 387tq and 388th meet ings o f t h e ACRS on J u l y 10 and August 7, 1992, respec- t i v e l y . A copy o f t h e ACRS r e p o r t on Volume I 1 o f EPRI’s ALWR U t i l i t y Requirements Document, da ted August 18, 1992, i s a t tached as Appendix F.

The s t a f f w i l l con t inue t o i n t e r a c t w i t h t h e ACRS d u r i n g i t s rev iew o f p o l i c y issul?s and any a d d i t i o n a l i ssues on e v o l u t i o n a r y ALWRs t h a t may be reso lved i n t h e c o n t e x t o f Volume I 1 o f t h e ALWR U t i l i t y Requirements Document. I n a d d i t i o n , t h e s t a f f w i l l con t inue t o i n t e r a c t w i t h t h e ACRS d u r i n g i t s rev iew o f Volume 111 o f t h e ALWR U t i l i t y Requirements Document and w i l l r e p o r t t h e r e s u l t s o f t h e Committee’s rev iew o f t h a t document and t h e s t a f f ’ s SER when i t i s completed.


4 OUTSTANDING ISSUES

During its review of EPRI's ALWR Utility Requirements Document, the staff ideritified several items for which additional information is required before the staff can reach a final conclusion. The staff considers these issues to be cutstanding. These outstanding issues fall into one of three categories: (1) open policy issues on which the staff has taken a position, but for which the Commission has not provided guidance; (2) open issues that must be satisfactorily resolved before the staff can complete its review of the Require- ments Document; or (3) confirmatory issues for which the staff will ensure that EPRI meets it commitments to revise the Requirements Documerlt. outstanding issues for the entire Requirements Document, provided by chapter or appendix with references to appropriate sections of the SER chapter or appendix given in parentheses, are listed below.

The

The designators in front of each issue provide a unique identifier for each issue. or passive plant designs, respectively. chapter in which it is identified. appendix, if applicable. open or a confirmatory issue, respectively. The final number is the sequential number assigned to it in Section 1.4 o f each chapter or appendix.

The letter "E" or "Pl' indicates that the issue applies to evolutionary The first number designates the

The letter that follows designates the The letter "0" or "C" designates whether it is an

4 . 1 Outstandinq Issues Pertainins to the Reauirements Document for Evolutionary Plant Desiqns

- OPEN ISSUES

The following is a list o f open policy issues pertaining to the Requirements Docuinent for evolutionary plant designs (Volume 11 ) on which the staff has take,? a position, but for which the Commission has not had the opportunity to provide guidance. There are no other types o f open issues resulting from the staf F's review of the Requirements Document for evolutionary plant designs.

Chawter 1 - Overall Reauirements E. 1 .O-1 tornado wind speeds (4 .5 .2 )

E . l .0 -2 leak before break (4 .5 .5 )

ADDeridix B to ChaDter 1 - Licensinq and Requlatorv Reauirements and Guidance

E.1B.O-1 impact of the elimination of the operating-basis earthquake from the design process (2 .1.1, Item 1V.A of Annex A, and Item 1.M o f Annex C, Item C of Annex D)

E.1B40-2 applicability of industry codes and standards (2.1.1 and Item 1I.A of Annex C)

E.1B.O-3 tornado design basis ( 2 . 1 . 2 and Item 1I.F of Annex C)


E.lB.O-4

E. 18.0-5

E. 1B. 0-6

E.lB.O-7

E.lB.O-8

E. lB.O-9

E.lB.O-10

E.lB.O-11

E. lB.O-12

E. lB.O-13

E. 16.0-14

E. 18.0-15

main steamline classification (2.3.1.1 and Item 1I.E of Annex C)

simplification of postaccident sampling system (2.3.2 and Item 11.1 o f Annex C)

containment leak rate testing (2.5.1 and Item 1I.H of Annex C)

source term (2.5.2.1, 2.5.2.2, Item 1.B of Annex A, and Item 1.A o f Annex C)

seismic hazard curves (Item 1I.C of Annex C)

leak before break (Item 1I.D o f Annex C)

containment bypass (Item 1I.G o f Annex C)

prototyping (Item 1I.K o f Annex C)

reliability assurance program (Item 1I.M o f Annex C) defense against common-mode failures in digital instrumentation and control systems (Item A o f Annex D)

analysis of external events beyond the design basis (Item B of Annex D)

control room annunciator reliability (Item G of Annex D)

Chapter 5 - Enqineered Safety Systems

E.5.0-1 core debris coolability (6.6.2)

CONFIRMATORY ISSUES

There are no confirmatory issues pertaining to the Requirements Document for evolutionary plant designs.

4.2 Outstandinq Issues Pertaininq to the Reauirements Document for Passive Plant Desiqns

OPEN ISSUES

The following i s a list of open issues that must be resolved before the staff can complete its review of the Requirements Document for passive plant designs (Volume 111).

ChaDter 1 - Overall Reauirements

P. 1 .o-1 scope of mitigation requirements (2.1 and 2.4)

P. 1 .o-2 regulatory treatment of non-safety-related systems (2.3.1, 4.3.1, 7, 10, 12.2.1, 12.2.3, and Appendix B)

Program Summary 4-2

P. 1 .I)-3

P. 1 .o-4

P. 1 .O-5

P. 1 .O-6

P. 1 .o-7

P. 1 .O-8

P. 1 .o-9

P. 1 .o-10

P. 1 .o-11

P. 1 .o-12

P. 1 .O-13

P. 1 .O-14

P.l .0-15

P. 1 .O-16

P. 1 A1-17

P. 1 .(1-18

automat ic standby l i q u i d c o n t r o l system f o r pass ive bo i l i ng -wa te r - r e a c t o r (BWR) des ign (2.3.2 and Appendix B)

check va l ve c a t e g o r i z a t i o n (2.3.2)

tornado wind speeds (4.5.2.5)

l e a k be fo re break (4.5.5)

seismic e v a l u a t i o n and des ign o f smal l -bore p i p i n g (4.7.3)

use o f I n s t i t u t e o f E l e c t r i c a l and E l e c t r o n i c s Engineers ( IEEE) Standard 323 (4.8.2)

method o f environmental q u a l i f i c a t i o n o f mechanical and e l e c t r i c a l equipment (4.8.2)

l i m i t s on n i t r i t e s , n i t r a t e s , and t o t a l halogens as c h l o r i n e (5.2.8)

pressurized-water-reactor (PWR) water chemist ry (5.5.2)

r e l i a b i l i t y assurance program framework (6.2)

q u a n t i t a t i v e r e l i a b i l i t y and a v a i l a b i l i t y goa ls (6.2)

i n t e g r a t i o n o f r e l i a b i l i t y eng ineer ing techniques (6.2, 6.3, and 6.4)

r e l a t i o n s h i p o f system requirements t o o v e r a l l p l a n t s a f e t y r e 1 i- a b i l i t y and a v a i l a b i l i t y goa ls (6.2)

d i f f e r e n c e between re1 i abi 11 t y assurance program f o r s a f e t y - and non-sa fe ty - re la ted systems (6.3)

human f a c t o r s cons idera t ions f o r ope ra t i on and maintenance p r o v i - sions (8.2)

computer s e c u r i t y re fe rence (11.12)

Amer ld ix A t o Chapter 1 - PRA Key Assumptions and Groundrules

P.1A.O-1 r e p o r t i n g o f core-damage-frequency r e s u l t s as mean va lues (1.7)

P.1A.O-2 po in t -es t ima te q u a n t i f i c a t i o n (1.8)

P.1A.O-3 q u a n t i t a t i v e t rea tment o f u n c e r t a i n t i e s (1.9)

P.1A.O-4

P.1A.O-5 guidance on model ing d e t a i l r e q u i r e d t o represent pass ive system

guidance on p resen t ing r e s u l t s o f p r o b a b i l i s t i c r i s k assessment (PRA) (1.10)

behav io r (2.1)

Program Summary 4-3

P. 1A.O-6

P.1A.O-7

P. 1A.O-8

P. 1A.O-9

P. 1A.O-10

guidance on modeling interactions between passive and active systems (2.1)

guidance for developing the success criteria for passive systems (2.3)

determination of an appropriate mission time

requirements to address the important passive design-specific areas of uncertainty (6.1)

failure rate for the main step-up transformers (8.2)

(2.10)

AoDendix B to ChaDter 1 - Licensinq and Requlatorv Requirements and Guidance

P. 1B.O-1

P. 1B.O-2

P. 1B.O-3

P. 1B.O-4

P. 18.0-5

P. 1B.O-6

P.1B.O-7

P.1B.O-8

P. 1B.O-9

P. 1B.O-10

P, 1B.O-11

P.1B.O-12

P. 18.0-13

P. 18.0-14

P.1B.O-15

impact of the elimination of the OBE from the design process (2.1.1, 3 . 3 . 1 , and Item 1.M o f Annex A )

applicability of industry codes and standards (2.1.1 and Item 1I.A of Annex A)

t o r n a d o d e s i g n b a s i s (2.1.2 and I t e m 1 I . F o f Annex A)

simplification of emergency planning requirements (2.1.3 and Item 1II.G of Annex A)

need for leakage control system for main steam isolation valves (2.3.1 and Item 1I.E of Annex A)

main steam isolation valve leakage rate (2.3.1 and Item 1I.E of Annex A)

simplification of postaccident sampling system (2.3.2 and Item 11.1 o f Annex A)

containment leak rate testing (2.5.1 and Item 1I.H of Annex A)

source term (2.5.2 and Item 1.A of Annex A)

hydrogen generation and control (2.5.3, 3.2.26 and Item 1.G of An- nex A)

safe shutdown requirements (2.5.6 and Item 1II.D of Annex A)

revised deficiency reporting requirements (Generic Safety Issue II.J.4.1) (3.2.3)

criteria for safety-related operator actions (Generic Safety Issue 8-37) (3.2.5)

diesel generator re1 iability (Generic Safety Issue B-56) (3.2.6)

allowable emergency core cooling system equipment outage periods (Generic Safety Issue B-61) (3.2.7)

Program Summary 4-4

P. 18.0-16

P. 18.0-17

P. lB.O-18

P.lB.O-19

P. 13.0-20

P.113.0-21

P. 113.0-22

P. 113.0-23

P. 111.0-24

P. 111.0-25

P. 111.0-26

P. 111.0-27

P. l t I .0-28

P. 1Ei.0-29

P. 1E1.0-30

P . 1E:.0-31

P . 1Em.O-32

P. l e .O-33

P. 18 .O-34

reactor coolant pump seal failures (Generic Safety Issue 23) (3 .2 .10)

bolting degradation or failure (Generic Letter 91-17) (Generic Safety Issue 23) (3 .2 .12 )

use of EPRI NP-5076 on good bolting practices for bolted joints (3 .2 .12)

bolting degradation or failure (Generic Safety Issue 29) (3 .2 .12)

effects of fire protection system actuation on safety-related equipment (Generic Safety Issue 57) (3 .2 .13 )

power-operated re1 i ef Val ve and block valve re1 i abi 1 i ty (Generic Safety Issue 70) (3 .2 .14 )

anticipated transients without scram (Generic Safety Issue 75) (3 .2 .16 and Item 1.6 o f Annex A )

unanalyzed reactor vessel thermal stress during natural convection cooldown (Generic Safety Issue 79) (3 .2 .17 )

control room habi tabi 1 i ty (3 .2 .18 )

low-temperature overpressure protection (Generic Safety Issue 94) (3 .2 .21 )

piping and use o f highly combustible gases in vital areas (Generic Safety Issue 106) (3.2.29)

dynamic qual i f i cat i on and test ng of 1 1 1 c arge-bore hydraul snubbers (Generic Safety Issue 113) (3 .2 .24 )

reliability, operability, and on-line testability of the final actuation contacts in protection systems (3.2.25)

essential service water pupp fgilures at multiplant sites (Generic Safety Issue 130) (3 .2 .29 ) .

guide1 ines for upgrading other! procedures (Generic Safety Issue HF 4 . 4 ) (3 .2 .33)

-

electronic display of proceduyes, use of mixed types of procedures from one control station to another, and use of active simulator to validate procedures ( 3 . 2 , 3 3 )

clarification of definition of local control stations (3 .2 .34 )

centralization of safety func t ions ( 3 . 2 . 3 4 )

development of a human factors verification and validation test plan ( 3 . 2 . 3 5 )

Program Summary 4-5

P. 16.0-35

P. 18.0-36

P. lB.O-37

P. 16.0-38

P. lB.O-39

P. lB.O-40

P. 18.0-41

P. 18.0-42

P. lB.O-43

P. 16.0-44

P. 16.0-45

P. lB.O-46

P. 16.0-47

P.lB.O-48

P. 18.0-49

P.16.0-50

P. 16.0-51

P. 16.0-52

P.16.0-53

P.lB.O-54

P.lB.O-55

P. 18.0-56

P. 18.0-57

P. 16.0-58

documentation of test activities for traceability and assurance that all human factors requirements are addressed during test and evaluation (3 .2 .35)

development of quantitative measures to assess human-system performance (3 .2 .35)

uniform damping values (3 .3 .1 )

modal combination of high-frequency modes for vi bratory loads (3 .3 .1 )

safety implications of control systems (Generic Safety Issue A-47)

heat exchanger testing (3 .3 .4 )

control of biofoul ing (3 .3 .4 )

zebra mussel fouling ( 3 . 3 . 4 )

mid-loop operation (Item 1.C of Annex A)

station blackout (Item 1.D of Annex A)

fire protection (Item 1.D of Annex A)

intersystem loss-of-coolant-accident (Item 1.E of Annex A)

core-concrete interaction - capability to cool core debris (Item 1.G of Annex A)

high-pressure core melt ejection (Item 1.1 of Annex A)

containment performance (Item 1.J of Annex A)

dedicated containment vent penetration (Item 1 .K of Annex A)

equipment survivability (Item 1 . L of Annex A)

inservice testing of pumps and valves (Item 1.N of Annex A)

electrical distribution (Item 1I.B o f Annex A)

seismic hazard curves (Item 1I.C of Annex A)

leak before break (Item 1I.D of Annex A)

classification of main steamline of BWB (Item 1I.E of Annex A)

containment bypass (Item 1I.G of Annex A)

level of detail (Item 1I.J of Annex A)

(3 .3 .3 )

Program Summary 4-6

I

P.lB.O-59 prototyping (Item 1I.K of Annex A)

P.lB.O-60 inspections, tests, analyses, and acceptance criteria (Item 1I.L of Annex A)

P.lB.O-61 reliability assurance program (RAP) (Item 1I.M of Annex A)

P.lB.O-62 site-specific PRAs (Item 1I.N of Annex A)

P.lB.O-63 severe-accident mitigation design alternatives (Item 11.0 of Annex A)

P.lB.O-64 generic rulemaking related to design certification (Item 1I.P of Annex A)

P.lB.O-65 regulatory treatment of non-safety systems (Item 1II.A of Annex A)

P.lB.O-66 definition of passive failure (Item 111.6 of Annex A)

P.lB.O-67

P.lB.O-68 control room habitability (Item 1II.E of Annex A)

P.1B.O-69 radionuclide attenuation (Item I1I.F of Annex A)

ChaD.;er 2 - Power Generation Svstems

P. 2 .o-1 safety valve design (3.4)

ChaDter 3 - Reactor Cool ant Svstem and Reactor Non-Safetv Auxi 1 i arv Systems

P .3 . 0- 1

P. 3.0-2

Chaoter 4 - Reactor Systems

P .4 . 0- 1 rod insertion capability after an'earthquake (2.2.6)

P .4 . O-2 inservice inspection of reactor pressure vessel (RPV) internal s (2.3.2)

P. 4 .(I-3 RPV thermocouples (3.3)

thermal-hydraulic stability of the Simplified Boiling Water Reactor (Item 1II.C of Annex A)

leak-testing of feedwater system valve that performs containment isolation function (5.5)

postaccident sampling system (7)

ChaDter 5 - Enqineered Safetv Svstems

P .5. (1 - 1 justification for 72-hour design-basi s period for control room habitability (2.1.1, 2.2, and 6.5)

P.5.C-2 need for activated charcoal filters (2.1.2)

P.5.C-3 timing of fission product release (2.1.3)


P.5.0-4

P.5.0-5

P.5.0-6

P.5.0-7

P.5.0-8

P.5.0-9

P. 5.0-10

P. 5 .O-11

P. 5.0-12

P. 5.0-13

P.5.0-14

P. 5.0-15

P .5.0- 16

P. 5.0-17

P .5.0- 18

P. 5.0-19

P .5.0-20

P. 5.0-21

P .5.0-22

P. 5 .O-23

P .5.0-24

/

[ I evaluation of aerosol fission product removal (2 .1 .6 )

secondary building fission product holdup and p ateout (2 .1 .7 and 6.4)

chemical form of iodine in containment (2 .1 .9 )

guidance on vendor-supplied information (2 .4 .2 )

identification of vital equipment (2 .5 )

use of carbon austenitic stainless steel for passive decay heat removal (PDHR) heat exchanger piping material (3 .3 )

diverse reactor protection system (RPS) input to control rods (3 .4 )

definition of the safe shutdown condition (4 .3 )

PDHR water pool capacity (4 .3 )

gravity drain tank for standby liquid control system (4.5)

separate connecting line for the two trains of the automatic depressurization system final stage (5.4)

elimination o f Type C leakage rate testing (6 .2 )

Type B air lock leak test interval requirements ( 6 . 3 )

dose consequence criteria for design-basis accidents (6 .4 )

maximum interval for Type C leakage rate testing (6 .3 )

need for safety-grade containment spray system and engineered safety features atmosphere cl eanup (2 .1 .6 and 6 .4 )

thyroid and beta skin radiation dose limits and credit for long- term use of breathing apparatus (6 .5 )

hydrogen concentration for PWR dry containment (6 .6 )

allowable compressive stress consistent with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (6 .6 )

safety-grade hydrogen recombi ners system (6.6)

containment steam bypass capability consistent with Standard Review Plan (7 .2)

ChaDter 6 - Buildinq Desiqn and Arranqement

P . 6 .0-1 classification o f non-safety-grade auxiliary systems as vital equipment (2 .3 .3 )

Program Summary 4-8

P. 6.0-2 t u rb ine -genera to r b u i l d i n g sa fe shutdown earthquake l o a d i n g condi- t i o n s ( 4 . 4 )

P .6.0-3 l o c a t i o n o f t h e c o n t r o l complex ( 4 . 5 . 4 )

P. 6.0-4 requi rements and acceptance c r i t e r i a f o r human f a c t o r s consider- a t i o n s ( 4 . 5 . 4 )

Chamer 7 - Fue l i nq and Re fue l i nq Systems

P . 7 . 0- 1 nondes t ruc t i ve t e s t i n g o f t h e spent f u e l pool l i n e r ( 2 . 3 . 1 )

P. 7 .O-2 i n s e r v i c e t e s t i n g requirements f o r t h e spent f u e l pool l i n e r ( 2 . 3 . 1 )

P. 7 .o-3 c r i t i c a l i t y of new f u e l i n new f u e l s torage f a c i l i t y ( 2 . 3 . 3 )

P. 7.0-4 r a d i o l o g i c a l consequences o f a f u e l hand l i ng acc ident ( 2 . 3 . 9 )

P. 7 .o-5 source term f o r a f u e l hand l ing acc ident ( 2 . 3 . 9 )

P. 7 .O-6 i n c o r r e c t water c l a r i t y requirement re fe rence ( 3 . 1 . 1 )

ChaD1;er 8 - P l a n t Coo l inq Water Systems

P. 8.0- 1 r e g u l a t o r y t rea tment o f non-sa fe ty - re la ted a c t i v e systems ( 2 . 2 , 3 .1 , 4 . 1 , 5, and 7 )

P.8.0-2 des ign requi rements f o r t h e f u e l pool c o o l i n g system ( 9 )

ChaD1:er 9 - S i t e S u m o r t Systems

P. 9.0-1 ba lance-o f -p lan t f i r e p r o t e c t i o n program ( 3 )

p. 9 -0 -2 independence o f v e n t i l a t i o n systems i n s i d e t h e containment t o prevent m i g r a t i o n o f smoke and h o t gases ( 3 . 3 . 1 )

P. 9 .o-3 requi rements f o r smoke-removal c a p a b i l i t y ( 3 . 3 . 1 )

P .9.0-4 s a f e t y c l a s s i f i c a t i o n f o r heat ing, v e n t i l a t i n g , and a i r c o n d i t i o n -

P .9 .0-5

i n g (HVAC) systems ( i n c l u d i n g pumphouse v e n t i l a t i o n system) ( 8 . 2 )

seismic Category I I / I c r i t e r i a f o r HVAC systems and components (8.2.1 and 8 .4 .1 )

P. 9 .O-6 use o f pos t f i l t e r i ns tead o f h i g h - e f f i c i e n c y p a r t i c u l a t e a i r f i l t e r s ( 8 . 2 . 1 )

P. 9 .o-7

P. 9 .O-8 c o n t r o l room h a b i t a b i l i t y f o l l o w i n g des ign-bas is acc ident ( 8 . 2 . 2 )

t h y r o i d and be ta s k i n r a d i a t i o n dose requi rements ( 8 . 2 . 2 )

P. 9 .o-9 omiss ion o f fans f o r warehouse areas ( 8 . 2 . 5 )

Program Summary 4-9

P.9.0-10

P.9.0-11 use of Uniform Building Code, Zone 2A (8.5)

P.9.0-12

omission of cooling coil supplied by chilled water for reactor cavity system (8.4.1)

power source and performance requirements following a loss of offsite power (8.5)

HVAC design for several areas not required to meet requirements specified in the General Design Criteria, Standard Review Plan, or regul at i ons (8.5)

P.9.0-13

ChaDter 10 - Man-Machine Interface Systems (M-MIS).

P. 10.0-1

P. 10.0-2

P. 10.0-3

P. 10.0-4

P. 10.0-5

-

P. 10.0-6

P. 10.0-7

P. 10.0-8

P. 10.0-9

P. 10.0-10

P. 10.0-11

P. 10.0-12

P . l o . 0- 13

P.10.0-14

P. 10.0-15

P. 10.0-16

P.10.0-17

independence of the software verification and validation review teams (3.1.2, 3.1.4, and 6.1.5)

use of commercial compilers for software used in safety systems (3.1.2)

dedication of commercial-grade software (3.1.2)

clarification of requirements for analysis and validation testing

use of rapid prototyping in the development and validation of functional specifications (3.1.3)

operator aids (3.4.5)

establishment and use of reliability and availability estimates (3 - 5)

definition of "practical" (3.5.3).

environmental conditions for equipment design including compatibility with tasks (3.5.3)

component reliability of M-MIS (3.5.4)

overall reliability of M-MIS (3.5.4)

sneak circuit analysis (3.5.4)

software maintenance (3.5.4)

minimum tests for continuous on-1 ine testing (3.6.1)

vulnerability o f power supplies for alarm systems (4.3.1)

guidance on criteria to establish priorities (4.3.4)

guidance on the maximum number of alarms (4.3.4)

of M-MIS (3.1.3)

Program Summary 4-10

P. 10 .0-18

P. 10.0-19

P. l o < 0-20

P. 10.0-21

P. 10.0-22

P. 10.0-23

a1 arm sequence recording (4.3.4)

valve position indication (4.4)

guidance on frequency allocation plan (4.6)

guidance on interference between communication systems and M-M equipment (4.6)

environmental conditions for minimally used local control stat (4.9.2)

guidance on inadvertent actuation of-control s at local control stations (4.9.2)

P.10.0-24

P. 10.0-25

P. 10.0-26

P. 10.0-27

P. 10.0-28

P. 10.0-29

P. 10.0-30

P. 10.0-31

P. 10.0-32

P. 10 .O-33

P. 10.0-34

P. 10.0-35

P . l o . 0-36

P . l o . 0-37

P. 10.0-38

P. 10.0-39

P.10.3-40

P. 10.3-41

5

ons

special physical security measures for the data transmission cable (5.2.1)

guidance on data system characteristics (5 .2.2)

propagation of common-mode failures through the data system (5.2.2)

expansion capability of multiplexers ( 5 . 2 . 3 )

re1 i abi 1 i ty of mu1 t i pl exing system (5.4)

software design aids and tools (6.1.1 and 6.1.5)

quality assurance requirements for safety-related software (6.1.2)

configuration management requirement for software (6.1.2)

software integrity (6.1.2)

guidance on software user documentation (6.1.2)

acceptance testing of commercially available software (6.1.2)

notification o f software errors or modifications of commercially delivered software products (6.1.2)

long-term configuration control of software (6.1.2)

clarification of top-down structured design approach (6.1.3)

guidance on convolution of software structure (6.1.3)

behavior of commerci a1 software when assumptions are violated (6.1.3)

guidance on memory protection (6.1.3)

use o f information by redundant safety channels (6.1.3)


P. 10.0-42

P. 10.0-43

P . l o . 0-44

P. 10.0-45

P.10.0-46

P. 10.0-47

P.10.0-48

P. 10.0-49

P . l o . 0-50

P. 10.0-51

P. 10.0-52

P. 10.0-53

P. 10.0-54

P.10.0-55

P. 10.0-56

P. 10.0-57

P. 10.0-58

P. 10.0-59

P. 10.0-60

P. 10.0-61

P. 10.0-62

P. 10.0-63

P. 10.0-64

P. 10.0-65

guidance on tools to improve re1 iabi (6.1.5)

definition of reasonable testing and considered (6.1.5)

specification of the level of divers 6.2.3, and 6.2.5)

ity and quality of software

degree of confidence to be

ty in safety systems (6.1.6,

guidance on performance of reliability eva

reference to I E E E 1050-1989 (6.2.2 and 6.2

compati bil i ty between M-MIS equipment and supply systems (6.2.2)

uation (6.1.6)

9)

ts external power

alarmed, self-diagnostic feature on clock update (6.2.3)

guidance on position of sensor isolation valves (6.2.5)

capaci tance-type pressure sensors (6.2.5)

minimal acceptance review criteria for isolation device (6.2.6)

voltage design of battery and dc system (6.2.8)

standards for surge suppression (6.2.8)

electromagnetic interference/radiofrequency interference (EMI/RFI) considerations for wiring shields (6.2.9)

specific grounding standards (6.2.9)

use of qualified isolators for wiring shields (6.2.9)

requirements for signal reconstruction (6.3.3)

use of interrupts (6.3.3)

redundancy of safety systems (6.3.4)

selection of automatic or manual control (7.2)

operation of plant by load dispatcher (7.3.2)

alternate means of inventory monitoring (7.13)

clarification of actuation logic self-testing (8.2

initiation of standby liquid control system (8.7)

3)

inclusion of four elements in human factors organizational structure (3.1.2 of Appendix B)


P. 10 .O-66

P. 10 -0-67

P. 10 .O-68

P. 10 0-69

P. 10,0-70

P. 10,0-71

P.10.0-72

P. 10.0-73

P . l o . 0-74

P. 10.0-75

P. 10.0-76

P.10.0-77

P. 10.0-78

P.10.0-79

P.10.0-80

P.10.0-81

P. 10.0-82

scheduling of human factors studies before start of control room design process (3.1.3 o f Appendix B)

guidance to the M-MIS designer to overcome past problems (3.1.4 of Appendix B)

guidance to improve interfaces between the operator and the plant (3.1.4 of Appendix B)

identification of human factors criteria, guidance, etc., that were used as the supporting bases for M-MIS requirements (3.2.1 of Appendix B)

traceabil ity o f human factors requirements to original source (3.2.2 of Appendix B)

method for establishing effective human factors requirements (3.2.3 o f Appendix B)

guidance on systems analysis (3.2.4 of Appendix B)

organization of plant information (3.2.4 of Appendix B)

configuration of operator’s workstation (3.2.5 of Appendix B)

human factors guidelines for new technology (3.3 of Appendix B)

illumination levels (3.3 of Appendix B)

maintenance procedures (3.4 o f Appendix B)

selection and qualification of plant personnel (3.5 of Appendix B)

training requirements for top-level personnel (3.6 of Appendix B)

human factors verification and validation test plan (3.7.1 o f Appendix B)

documentation of human factors test -activities (3.7.2 of Ap- pendix B)

team performance (3.7.4 of Appendix B)

ChaDter 11 - Electric Power Svstems

P . l l . D - 1 regulatory treatment of active non-safety systems (2.2.1, 3.2.2,

P. l l .9-2

5.2.1, 5.2.4, 5.2.5, and 8.2.1)

clarification of terminology describing the role of non-safety systems (2.2.1)

P.11.1)-3 loss of non-safety ac power during shutdown conditions (2.2.6)

Progr l rn Summary 4-13

P. 11 .O-4

P. 11 .O-5

P. 11 .O-6

P. 11 .O-7

P. 11 .O-8

P.11 .o-9

P. 11 .o-10

P. 11 .o-11

P. 11 .o-12

P. 11 .O-13

P. 11 .O-14

P. 11 .O-15

P.11.0-16

P. 11 .O-17

P.11.0-18

second offsite power supply circuit for permanent non-safety load buses during power operations (3.2.1)

consistency of Appendix A to Chapter 1 assumptions with Chapter 11 (3.2.3)

data provided in Appendix A to Chapter 1 relative to loss-of- offsite-power events (3.2.5)

non-safety ac electrical power systems during a smal 1 -break 1 oss- of-coolant accident (5.2.1)

clarification of revised requirements in Sections 1.5.2 and 5.3.1.1 (5.2.1)

applicability of the 2000-hour rating and its impact on peaking operation (5.2.1)

peaking operation of non-safety standby power sources (5.2.3)

consistency o f failure and unavailability rates in Appendix A to Chapter 1 (5.2.4)

justification for the use of independent self-contained cooling systems (5.2.6)

potential loss of dc buses (7.2.1)

load shedding for 72-hour battery endurance (7.2.3)

backup battery and battery charger (7.2.4)

transfer scheme for the safety dc power supply system (7.2.5)

battery and battery charger instrumentation and alarms (7.2.8)

lack of requirements for electric protective assemblies in RPS power for BWRs (7.2.9)

ChaDter 12 - Radioactive Waste Processins Systems

P.12.0-1 source term basis for designing radioactive waste systems and evaluation of offsite effluent radioactive nuclide concentration

basis for 2-minute-delay requirement for BWR turbine gland seal system exhaust (3.3.1)

production sources for essentially nonradioactive steam (3.3.1)

(2.2.2)

P.12.0-2

P.12.0-3

' P.12.0-4 discrepancy between Figure 12.3-1 and requirement in Chapter 13 (3.3.1)


P.12.0-5 use of post-filter downstream of charcoal adsorber in ventilation

P.12.0-6

exhaust systems (3.3.3)

guidance regarding direct piping from radioactive plant systems to sumps or waste collection tanks (BWR) (4.2)

requirements for liquid radioactive waste processing systems (LRWPS) filter housing and components (4.2)

P.12.0-7

P.12.0-8 requirements for LRWPS filters (4.2)

P.12.0-9 requirements for LRWPS ion exchangers (4.2)

- Chapter 13 - Main Turbine-Generator Systems

P.13.0-1 turbine missile generation probability effect on safety-related systems or components (3.1.4)

P.13.0-2 inservice inspection interval for turbine steam valves (3.3)

- CONFIRMATORY ISSUES

The following is a list of confirmatory issues for which the staff will ensure that EPRI meets its commitments to revise the Requirements Document for passive plant designs.

- Chapter 1 - Overall Reauirements

P.l.C-1 tornado wind speeds (4.5.2)

P.l.C-2 internal flooding design criteria (4.5.5)

P.l.C-3 compliance with Regulatory Guides 1.26 and 1.29 (9)

m m d i x A to ChaDter 1 - Key AssumDtions and Groundrules

None

m z n d i x B to ChaDter 1 - Licensinq and Requlatory Requirements and Guidance

P.1H.C-1 commitments to 10 CFR Part 21 and 50.55(e) (3.2.3)

P.ll3.C-2 Target Rock safetylrelief valves (3.2.5)

( & p t e r 2 - Power Generation Systems ~ None?

( & p t e r 3 - Reactor Coolant System and Reactor Non-Safety Auxiliary Systems

I P.3 C-1 an t ic ipa ted- t rans ients -wi thout-scrams events (5.5)

m b t e r 4 - Reactor Systems


None

ChaDter 5 - Enqineered Safety Systems

None


None

ChaDter 7 - Fuelinq and Refuelinq Systems

P.7.C-1 quality group classification of components for the new and spent fuel storage racks (2.3.1)

ChaDter 8 - Plant Coolinq Water Systems

None

ChaDter 9 - Site Surmort Systems.

P.9.C.1 guidance on the need for assessing the area interior to security detection equipment (5.2.7)

Chapter 10 - Man-Machine Interface Systems

P.1O.C-1 intention of use (1)

P.1O.C-2

P.1O.C-3 software common-mode failures (6.1.6, 6.2.5, and 8.2.1)

impact of support software on installed software (6.1.2)

P.1O.C-4 BWR depressurization system design change (8.6)

P.1O.C-5 manning of M-MIS that controls security functions (10.2.3)

ChaDter 11 - Electric Power Systems

P.11.C-1 isolation of safety systems from their non-safety sources (2.2.5)

P.1l.C-2 minimum starting voltages for valve actuator motors (2.2.8)

P.1l.C-3 requirements for main step-up transformers (3.2.6)

P.1l.C-4 harmonic distortion effect of adjustable speed drives (4.2.3)

P.1l.C-5 location of non-safety standby ac power sources (5.2.2)

P.1l.C-6

P.1l.C-7

P. 11 .C-8

sizing of thermal overload devices (6.2.1)

battery sizing criteria specified for 72-hour coping capability (7.2.2)

physical and electrical separation of the safety dc and low-vol tage


vital ac power supply system (7.2.5)

P. 11 .C-9

P.11.C-10 battery capacity margin (7.2.7)

non-safety dc power suppl ies for switchyard circuit protection and control equipment (7.2.6)

P.1l.C-11 normal and emergency lighting following design-basis events (8.2.1)

P.1l.C-12 intensity of emergency lighting system (8.2.2)

ChaDter 12 - Radioactive Waste Processins Systems

None

ChaDter 13 - Main Turbine-Generator Systems

P.13.C-1 compliance with Regulatory Guide 1.29 seismic design classification (3.1.1)

P.13.C-2 compliance with Standard Review Plan quality group classifications (3.1.2)


5 VENDOR- OR UTILITY-SPECIFIC ITEMS

During its review of EPRI's ALWR Utility Requirements Document, the staff identified items that were inadequately addressed in the document or were i s s u s that could not be addressed generically. These items will have to be satisfactorily resolved during the staff's review o f a vendor- or utility- specific application (i .e., an appl ication for final design approval and desilgn certification or a combined construction permit and operating 1 icense).

As discussed in Section 1 . 2 of this report, the Requirements Document has no legal or regulatory status and is not intended to demonstrate complete compliance with the Commission's regulations, regulatory guidance, or policies. It is not intended to be used as a basis for supporting final design apprsval and design certification for a specific design, nor is it to be used to substitute for any portion of the staff's review of future applications for final design approval and design certification. Specifically, satisfactory resolution of the items identified in this section for a vendor- or utility- specific application will not, by itself, support a finding that the application complies with the Commission's regulatory requirements. The staff will p e r f m n a complete licensing review of these applications using the Standard Revi2w Plan (NUREG-0800) and other appropriate Commission guidance. tory resolution of the vendor- or utility-specific items constitutes only one portion of the staff's review.

The iendor- or utility-specific items for the entire Requirements Document, provided by chapter or appendix with references to appropriate sections of the SER :hapter or appendix given in parentheses, are listed below. The designators in front of each issue provide a unique identifier for each issue. The lettlw "E" or "P" indicates that the issue applies to evolutionary or passive plant designs, respectively. The first number designates the chapter in which it i s identified. The letter that follows designates the appendix, if applicable. The letter V designates that i t is a vendor- or utility-specific item. Section 1 . 5 of each chapter or appendix.

Satisfac-

The final number provides the sequential number assigned to it in

5 . 1 Vendor- or Utilitv-SDecific Items Pertaininq to the Reauirements Document for Evolutionarv Plant Desiqns

The Following is a list of vendor- or utility-specific items that are identified in the SER (Volume 2 of this report) on the Requirements Document for evolutionary plant designs.

Chapter 1 - Overall Reauirements

E . l . ' J - 1

E . l . ' J - 2

scope of mitigation features ( 2 . 1 and 2 . 4 )

implementation of design characteristics intended to enhance accident resistance ( 2 . 2 )

bounding analysis by standard site design parameters ( 2 . 3 . 1 ) E.l.'J-3


E. 1 .V-4

E. 1 .V-5

E. 1 .V-6

E. 1 . V-7

E. 1 .V-8

E. 1 .V-9

E. 1 .V-10

E.1.V-11

E. 1 .V-12

E. 1 .V-13

E. 1 .V-14

E. 1 .V-15

E. 1 .V-16

E.l.V-17

E. 1 .V-18

E. 1 .V-19

E. 1 .V-20

E . l .V-21

E. 1 .V-22

E. 1 .V-23

E. 1 .V-24

E. 1 .V-25

E. 1 .V-26

E. 1 .V-27

selection of initiating events and their frequency categorization (2.3.2)

acceptance criteria for transient and accident analysis (2.3.2)

anticipated transient without scram response analysis (2.3.2)

acceptability of analytical codes and methodologies for safety analysis (2.5)

60-year plant life (3.3, 4.8.2, 8.2, and 11.3)

operation of PWR with a secured reactor coolant pump (3.5)

defense-in-depth analysis (3.5)

event response capability (3.5)

fuel burnup requirements (3.6)

extended operating life of control blades and control rod assem- b l i e s (3.6)

safety classification (4.3.1)

seismic qualification by experience (4.3.2 and 4.8.1)

non-seismic building structures (4.3.2 and 4.7.2)

structural design and construction codes (4.4 and 4.4.1)

elimination of operating-basis earthquake from design (4.4.3, 4.7.3, and Appendix B)

definition of support group (4.4.3)

use of Appendix N of ASME Code, Section I 1 1 (4.4.3 and 4.7.3)

analysis of vibratory loads with significant high-frequency input (4.4.3)

use of nonlinear analysis to account for gaps between pipes and - . piping supports (4.4.3)

probabilistic approach for changing existing combinations (4.5.1)

recurrence interval for wind loadings (4.5.2)

maximum ground water level (4.5.2)

precipitation for roof design (4.5.2)

snow loading (4.5.2)

oads and/or loading

Program Summary 5-2

E. 1.5-28

E. 1.5-29

E. 1 .'5-30

E. 1 .'!-31

E. 1 .'/-32

E. 1 .'I-33

E. 1 . '1-34 E. 1 .'1-35

E. 1 . \1-38

E. 1.11-39 E. 1 .11-40

E. 1 .v-41

E. 1 .\'-42

E. 1 . \'-43 E. 1. \'-44 E . 1. \'-45

E. 1 .\1'-46

E.l.1-47

E. 1 .U-48

E.1.V-49

E. 1 J - 5 0

E. 1.V-51

E. 1 .V-52

E. 1 . V-53 E. 1 .V-54

detailed quantification of soil parameters (4.5.2)

minimum margin against liquefaction (4.5.2)

external hazards eval uat i on (4.5.2)

number of full-stress cycles (4.5.2 and 4.8.1)

site-specific SSE (4.5.2)

power spectrum density function of the time history (4.5.2)

external impact hazards (4.5.2)

design temperature (4.5.2)

protection against surface vehicle bombs (4.5.3)

BWR safetylrel i ef Val ve 1 oads (4.5.4)

NUREG-1061 methodology and acceptance criteria for leak before break (4.5.5)

hydrodynamic loads from safety/relief valves (4.5.5)

suppression pool dynamic 1 oads (4.5.5)

design against internal-missile generation (4.5.5)

design of concrete containment (4.6.1)

load combinations for seismic Category I buildings and structures (4.6.1)

design o f seismic Category I steel structures ( 4 . 6 . 1 )

combination of pipe rupture loads with seismic loads for seismic Category I structures (4.6.1 and 4.6.1)

combination of loss-of-coolant-accident and SSE loads (4.6.1)

load combinations for safety-related portions of the plant (4.6.2)

dynamic analysis techniques (4.7.2)

methodology for generating design response spectra or time histories (4.7.2)

structural damping values (4.7.2)

masonry walls in Category I buildings (4.7.2)

use of expansion anchor bolts - compliance with Office of Inspec- tion and Enforcement Bulletin 79-02 (4.7.2 and 4.7.3)

Program Summary 5-3

E.1.V-55

E.1.V-56

E.1.V-57 use of ASME Code Case N-411 (4.7.3)

E.1.V-58

E.1.V-59 construction of core support structures (4.7.3)

E.1.V-60 fatigue design curves (4.7.3)

stability of shell-type structures under compression (4.7.2)

seismic evaluation and design o f small-bore p i p i n g (4.7.3)

use of ASME Code Cases N-411 and N-420 in same analysis (4.7.3)

I E.1.V-61 use of I E E E 323 (4.8.2)

E.1.V-62 environmental qualification of mechanical and electrical equipment

E.1.V-63 use of zinc to reduce radiation fields (5.2.7)

(4.8.2)

E.1.V-64 limits on nitrites, nitrates, and total halogens as chlorine

E.1.V-65

(5.2.8)

grinding controls for PWRs (5.3.1)

E.1.V-66

E.1.V-67 hardness limits for stainless steel (5.3.1)

effect of fabrication processes on intergranular stress corrosion cracking (5.3.1 and 5.3.1)

E.1.V-68 use of Alloy 600 and other alloys (5.3.1)

E.1.V-69 allowance for carbon and low-alloy-steel corrosion (5.3.1)

E.1.V-70 selection of seals, gaskets, and protective coatings (5.3.5)

E.1.V-71 aging of cable insulation and other electrical materials (5.3.6)

E.1.V-72 use of hydrogen water chemistry for the advanced BWR design (5.5.2)

E.1.V-73 PWR water chemistry (5.5.2)

E.1.V-74 submittal of operational reliabil-ity assurance program (O-RAP) (6)

E.1.V-75 organizational description for reliability assurance program (6.1)

E.1.V-76

E.1.V-77 reliability data bases (6.2)

analyses methods or models used in developing the reliability assurance program (6.2)

E.1.V-78 reliability, maintainability, and testability analyses (6.2)

E.1.V-79 apportionment of contributions of structures, systems, and components to core damage frequency (6.3)

Program Summary 5-4

E . 1 V-80

E. 1 I V-81

E . 1. V-82

E. 1. V-83

E.1.V-84

E.1.V-85

E. 1 .V-86

E. 1 .V-87

E . 1 .V-88

E. 1 .V-89

E . 1 .V-90

E. 1 .V-91

E . 1. V-92

E.1.V-93

E. 1. V-94

E. 1. V-95

E . 1. V-96

E. 1.5-97

E. 1 . '/-98 E. 1 .'/-99

E. 1 .'/-lo0

E. 1 .'/-lo1

E.1.V-102

E. 1 .\/-lo3

priority of safety in accident recovery (6.3) r e l a t i o n s h i p between s a f e t y and p roduc t i on a v a i l a b i l i t y (6.3)

e f f e c t o f l i m i t a t i o n s on r e f u e l i n g d u r a t i o n on p l a n t s a f e t y (6.3)

e f f e c t o f planned outage d u r a t i o n on p l a n t s a f e t y (6.3)

e f f e c t o f ma jor outage d u r a t i o n on p l a n t s a f e t y (6.3)

i n s p e c t i o n o f c o n s t r u c t i o n a c t i v i t i e s (7 and 11.13)

q u a l i t y assurance f o r non -sa fe ty - re la ted f a c i l i t i e s and systems (7)

i n s t a l l e d operating-phase s e c u r i t y system (7)

r e l i a b i l i t y o f modular c o n s t r u c t i o n ( 7 )

use o f I E E E P1025 P1023-1988/D5 and EPRI-2360 f o r guidance regard- i n g human f a c t o r s eng ineer ing (8.2)

i n s p e c t i o n and v e r i f i c a t i o n o f s e c u r i t y l o c k s r o b o t i c a l l y (8.3)

q u a l i t y assurance requirements f o r a l l equipment, s t r u c t u r e s , systems, f a c i l i t i e s o r so f tware t h a t have some s a f e t y importance o r has one importance (9)

compliance o f FDA/DC a p p l i c a t i o n s w i t h Commission's r e g u l a t i o n s and guidance (10)

i ssue r e s o l u t i o n f o r FDA/DC reviews (10) .

i nspec t i ons , t e s t s , analyses, and acceptance c r i t e r i a (10)

imp lementa t ion o f s i m p l i f i c a t i o n o b j e c t i v e (11.4)

implementat ion o f s t a n d a r d i z a t i o n o b j e c t i v e (11.5)

check v a l v e t e s t i n g methods (12.2.2)

f u l l - f l o w t e s t i n g o f check va lves (12.2.2)

q u a l i f i c a t i o n t e s t i n g o f a c t i v e and nonac t i ve motor-operated va lves (MOVs) (12.2.2)

t e c h n i c a l concerns rega rd ing MOVs (12.2.2)

l e a k r a t e t e s t i n g f o r i n d i v i d u a l containment i s o l a t i o n va lves (12.2.2)

i n s t r u m e n t a t i o n t o determine n e t p o s i t i v e s u c t i o n head d u r i n g a l l modes o f o p e r a t i o n (12.2.3)

t e s t i n g o f pump f l o w r a t e (12.2.3)

Program Summary 5-5

E.l.V-104 frequency and extent of disassembly and inspect re1 ated pumps (12.2.3)

ADDendix A to ChaDter 1 - PRA Kev AssumDtions and Groundrules

on of safety-

E.1A.V-1

E.1A.V-2

E.1A.V-3

E. lA.V-4

E.lA.V-5

E. lA.V-6

E.1A.V-7

E. lA.V-8

E.lA.V-9

E. 1A.V-10

E. lA.V-11

E.lA.V-12

E. lA.V-13

E.lA.V-14

E.lA.V-15

E.lA.V-16

E. lA.V-17

E. lA.V-18

E. lA.V-19

E. lA.V-20

E.1A.V-21

E. lA.V-22

E. 1A.V-23

E. lA.V-24

use of PRA in design

modeling of a PRA (1.6)

shutdown and low-power events (1.6)

external events (1.6, 3.3, and 6.1)

core damage frequency (1.7)

uncertainty treatment (1.9 and 6.1)

documentation of method of truncation and 2.5)

F accident seqi ences (1.10

low-frequency accident initiators leading to core damage (2.2)

mission time (2.10)

failure rate for components (2.11)

tornadoes and extreme winds (3.2)

external river flooding (3.2)

hurricanes and storm surges (3.2)

tsunami (3.2)

internal fires (3.2)

site-specific external events

internal flooding (3.2)

seismic hazards analysis (3.3)

core-damage-sequence binning (4.1)

plant damage state definition (4.2)

containment isolation assumptions and criteria (4.3)

in-pl ant sequence assessment (4.5)

containment event analysis (4.6)

details of uncertainty analysis (4.6)

Program Summary 5-6

E. 1A. V-25

E. 1A .V-26

E. 1A. V-27

E.1A.V-28

E. 1A. V-29

E. 1A. V-30

E.1A.V-31

E.1A.V-32

E.1A.V-33

E. 1A. V-34

E.1A.V-35

E. 1A.V-36

E. 1A.V-37

E. 1A. V-38

E.1A.V-39

E.1A.V-40

E. 1A.V-41

E . 1A.V-42

source term definition (4.7)

event tree binning (4.8)

risk measures related to containment performance (4.8)

use of mean values for characterization of risk results (5.1)

assessment of risk measures (5.2)

calculation of offsite consequences (5.2)

importance analysis for input to reliability assurance program

assessment of containment response (6.2)

source term (6.3)

scope and objective of human reliability analysis (HRA)(7.1)

process and criteria to confirm adequacy of human reliability analysis (HRA) (7.2)

impact o f advanced technologies on HRA (7.3)

function, task, timeline, and link analyses (7.3)

generic data sources (7.3)

performance shaping factors and their evaluation tools (7.3)

quantification methods for HRA (7.3)

loss of offsite power frequency (Annex A)

site data (Annex B)

(6.1)

ADDendix B to ChaDter 1 - Licensincl and Requlatorv Reauirements and Guidance

E.1B.V-1

E.1B.V-2 issue resolution for FDA/DC reviews (1.3)

compliance of FDA/DC applications with Commission’s regulations and guidance (1.3)

E.1B.V-3 elimination of missile provisions (2.1.2)

E.1B.V-4 dynamic seismic analysis of main steam piping and condenser (2.3.1.1 and Item 1I.E o f Annex C)

E.1B.V-5 main steamline classification (2.3.1.1 and Item 1I.E of Annex C)

E.1B.V-6 seismic analysis and plant walkdown of turbine building (2.3.1.1 and Item 1I.E of Annex C)

Program Summary 5-7

E.lB.V-7

E. 1B. V-8

E . lB.V-9

E. lB.V-10

E.lB.V-11

E. lB.V-12

E. lB.V-13

E.lB.V-14

E. lB.V-15

E. lB.V-16

E . lB.V-17

E.lB.V-18

E. lB.V-19

E. 1B .V-20

E. lB.V-21

E. lB.V-22

E . lB.V-23

E. lB.V-24

E. lB.V-25

E. lB.V-26

plateout considerations for main steam piping and valves (2.3.1.2 and Item 1II.F of Annex C)

reactor pressure vessel level instrumentation system (2.4.1)

source term (2.5.2.1, 2.5.2.2, Item I.Bof Annex A, and Item 1.A of Annex C)

compliance with Branch Technical Position MTEB 6.1 (2.5.2.2)

fission product cleanup analysis (2.5.2.2)

deletion of charcoal adsorbers (2.5.2.2)

dedicated containment vent penetration (2.5.3 and Item 1.K of Annex C)

decoupl ing of operating-basis earthquake (OBE) from safe shutdown earthquake (SSE) in seismic design of structures (Generic Safety Issue A-40) (3.2.7)

deletion o f OBE damping values in seismic design o f structures (Generic Safety Issue A-40) (3.2.7)

use of algebraic sum method for modal combination of high-frequency modes for vibratory loads (Generic Safety Issue A-40) (3.2.7)

use of spectral peak shifting techniques in lieu of spectral broadening (Generic Safety Issue A-40) (3.2.7)

plant-specific design and arrangement of control systems (Generic Safety Issue A-47) (3.2.9)

conformance to 10 CFR 50.34(f) hydrogen control requirements (Generic Safety Issues A-48 and 121) (3.2.10 and 3.2.46)

re1 iabil ity of emergency diesel generators (Generic Safety Is- sue B-56) (3.2.14)

resolution of Generic Safety Issues 2 and 110 (3.2.18 and 3.2.42)

resolution of Generic Safety Issue 15 (3.2.19)

independent reactor cool ant pump seal cool i ng during station blackout (Generic Safety Issue 23) (3.2.20)

resolution o f Generic Safety Issue 24 (3.2.21)

design details on threaded fasteners (Generic Safety Issue 29) (3.2.22)

reduction of biofouling in open-cycle service water and component cooling water systems (Generic Safety Issue 51) (3.2.23)

Program Summary 5-8

E. 111.V-27

E.1fl.V-28

E. 1El.V-29

E.1EI.V-30

E. 1E:.V-31

E. 1Es.V-32

E.1E .V-33

E. le .V-34

E. 18 .V-35

E. 1B. V-36

E.lB.V-37

E . 1B. V-38

E. lB.V-39

E.lB.V-40

E. lB.V-41

E.1B.V-42

E. 1B .V-43

E. 1B .V-44

E.1B V-45



equipment classification and vendor interface for reactor trip system components (Generic Safety Issue 75) (3.2.27)

2-week requirement for corrective maintenance (Generic Safety Issue 75) (3.2.27)

preventive maintenance and surveillance program for reactor trip breakers (Generic Safety Issue 75) (3.2.27)


cooldown rate in natural convection cooldown analysis (Generic Safety Issue 79) (3.2.29)

low-density storage racks in spent fuel pool for most recently dis- charged fuel (Generic Safety Issue 82) (3.2.30)

pl ant-speci f i c design and arrangement for control room heating, ventilating, and air conditioning (HVAC) system (Generic Safety Issue 83) (3.2.31)

design of emergency filter units (Generic Safety Issue 83) (3.2.31)

design details for control room capacity following a design-basis accident (Generic Safety Issue 83) (3.2.31)

design details for control room HVAC systems in the smoke removal mode (Generic Safety Issue 83) (3.2.31)


adequacy of 1 ow-temperature overpressure protection design (Generic Safety Issue 94) (3.2.34)

adequacy of BWR water level redundancy (Generic Safety Issue 101) (3.2.37)

interfacing system design details (Generic Safety Issue 105) (3.2.39)

inservice testing programs and technical specifications for appropriate pressure isolation valves (Generic Safety Issue 105) (3.2.39)

resolution of Generic Safety Issue 106 (3.2.4p)

environmental qualification and inservice inspection and testing o f large-bore hydraulic snubbers (Generic Safety Issue 113) (3.2.43)

Program Summary 5-9

E. lB.V-46 use of prestressed concrete containments (Generic Safety Issue 118) (3.2.44)

E.lB.V-47 reliability, operability, and on-line testability of protection system final actuation contacts (Generic Safety Issue 120) (3.2.45)

E.lB.V-48 operator training program and emergency operating procedures related t o initiating feed-and-bleed cooling (Generic Safety Issue 122.2) (3.2.50)

E.lB.V-49 auxiliary feedwater analyses (Generic Safety Issue 124) (3.2.52)

E.lB.V-50 operational aspects of electrical power reliability (Generic Safety Issue 128) (3.2.56)

E.lB.V-51 resolution of Generic Safety Issue 130 (3.2.57)


E.1B.V-53 resolution of Generic Safety Issue 135 (3.2.59)

E.1B.V-54 resolution o f Generic Safety Issue 142 (3.2.60)


E.lB.V-56 resolution of Generic Safety Issue 151 (3.2.62')

E.lB.V-57 assessment of safety service water system failure modes and contributions to core damage frequency and identification of dominant accident sequences (Generic Safety Issue 153 (3.2.63)

E.lB.V-58 resolution of Generic Safety Issue HF 4.4 (3.2.64)



Chapter 2 - Power Generation Systems

E.2.V-1 safety valve design (3.4)

E.2.V-2 attachment loads for safety and relief valves (3.4)

E.2.V-3 side stream condensate polisher (4.3)

E.2.V-4

Chapter 3 - Reactor Coolant System and Reactor Non-Safety Systems

E.3.V-1

E.3.V-2

condensate makeup system raw water pretreatment (6.4)

power supplies for power-operated relief valves (3.3)

pressurizer heater power source control design (3.4)

E.3.V-3 chemical and volume control system design (6.2)


U t e r 4 - Reactor Systems

E.4.V-1 reactor pressure vessel fatigue design criteria (2.3.2)

E.4. V-2 BWR thermal-hydraul ic stability performance during an anticipated transient without scram (4.2)

E.4. V-3 BWR nuclear and thermal-hydraul ic design for extended cycle operation (4.2)

E.4. V-4 effect o f electric protective assemblies on reactor protection system power supply requirements (5.3)

E.4.J-5 PWR thermal-hydraulic stability and xenon stability characteristics (7 2)

E .43 -6 PWR fuel design for load-following capability (7.2)

E.4.'11-7 60-year service life for control rod drive mechanisms (8.2)

ChaDter 5 - Enqineered Safety Systems

E.5.'1-1 containment performance criteria for severe accidents (2.1)

E.5.'/-2 metal-water reaction and hydrogen generation and control during a severe accident (2.3 and 6.5.1)

E.5.Y-3 fire protection (2.5)

E.5.V-4 diesel generator start time (3.2)

E.5.V-5 detailed LOCA analysis concerning core spray for BWRs (4.1)

E.5.V-6 safety classification of containment

E. 5. V-7 suppression pool bypass 1 eakage (4.5

E.5.V-8 suppression pool temperature-moni tor

E.5.V-9 intersystem LOCA (5.2)

E.5.V-10 operation o f RHR system with reduced tory (Generic Letter 87-12) (5.2)

E.5.V-11 shutdown risk (5.2)

E.5.V-12 feed-and-bleed capability (5.4)

spray system (4.4 and 7.2)

and 7.2)

ng system (4.6)

reactor cool ant system nven-

E.5.V-13 safety depressurization and vent system (5.4, 5.5, and 6.6.5)

E.5.\'-14 use o f remote manual valves on essential lines that are not part o f the engineered safety systems (6.2)

E.5.\'-15 Type C leak testing (6.2)

Program Summary 5-1 1

E.5.V-16 containment integrated leak rate testing (6.3.1)

E.5.V-17 Type A leak testing (6.3.1)

E.5.V-18 Type B testing of air locks (6.3.2)

E.5.V-19

E.5.V-20 Type C containment valve leak rate testing interval (6.3.3)

use of water in Type C containment leak rate testing (6.3.3)

E.5.V-21 control systems for radiolytically generated hydrogen (6.5.2)

E.5.V-22 design criteria for igniter system (6.5.3)

E.5.V-23 evaluation of igniter system (6.5.3)

E.5.V-24 method for determining load collapse of containment (6.6.1)

E.5.V-25 concrete containment analysis (6.6.1)

E.5.V-26 containment overpressure protection (6.6.3)

E.5.V-27 functionability of fission product control systems during a severe accident (6.6.4)

E.5.V-28 equipment survivability criteria for severe accidents (6.6.6)

E.5.V-29 accident management plan (6.6.8)

E.5.V-30 dynamic effects of pipe breaks during severe accidents (7.2)

E.5.V-31 main steam isolation valve leakage rate (7.2)

E.5.V-32 suppression pool design features (7.3)

E.5.V-33 containment leak rate (8.1)

E.5.V-34 postaccident pH control (8.2 and Appendix B to Chapter 1)


E.6.V-1 thermal growth of steel members (2.1)

E.6.V-2 inspectability of structural walls (2.1)

E.6.V-3 deviations from National Fire Protection Association codes and standards (2.3)

E.6.V-4 qualification criteria for fire barriers (2.3)

E.6.V-5 fire protection features in the heating, ventilation, and air conditioning (HVAC) design criteria (2.3)


E. 6. '1-6

E. 6 .'/-7

E. 6. \/-8

E. 6 .\1-9

E. 6. \!- lo E. 6 .\'-ll

E. 6. \'-12

E. 6. \'-13

E. 6.1'-14

E. 6 .\I-15

E. 6. \'-16

E. 6-\r -17

E. 6-\-18

E. 6 4 - 1 9

E.6.V-20

E. 6 .V-21

E.6.V-22

E. 6.V-23

E.6.V. 24

E. 6 .V-25

E.6.V-26

E. 6.V-27

E. 6 .V-28

compliance with the requirements of Three Mile Island (TMI) Action Plan Item II.B.2 (2.3)

details of shielding design and shielding computer codes (2.3, 2.4, and 4.2.8)

effect of site-specific topography on standard overall site arrangement (3.1)

flooding protection design requirements (3.3.1)

alternative seismic restraint devices (4.2.3)

piping and instrument line support design (4.2.4)

description of airborne radioactive material sources (4.2.5)

potential high-radiation areas, shielding, and measures for rninirniz- ing exposure (4.2.8 and 4.2.9)

review of coatings against SRP Section 6.1.2 (4.2.10 and 4.3.2)

use of epoxy-coated reinforcing bars at intake structures (4.2.11)

features to ensure H2 concentrations do not exceed detenation levels (4.3.2)

elimination of diagonal rebar in reinforced-concrete containment (4.3.2)

floor size for reactor vessel cavityldrywell (4.3.2)

design features that preclude potentially lethal radiation levels

containment access control (4.3.3 and 4.3.4)

details o f design of BWR reactor building (4.4.2)

details of design of PWR auxiliary building (4.4.3)

turbine-generator building seismic design loading (4.5.2)

details of design of BWR turbine-generator building (4 i5 .4 )

details o f design of radwaste facility (4.6.3)

details of emergency onsite power supply facility (4.6.4)

details of HVAC design for control complex (4.6.5)

details of design of technical support center (4.6.6)

(4.3.3)


ChaDter 7 - Fuelinq and Refuelinq Systems

E.7.V-1 quality group classification of components for new and spent fuel storage racks (3 .2 .1 )

E.7.V-2 radiological consequences of fuel handling accident (3 .2 .2 )

E.7.V-3 protection against tampering during refueling activities (3 .2.4)

E.7.V-4 design of the overhead bridge crane (6 .1 .2 )

E.7.V-5 radiological consequences of fuel cask drop acc

E.7.V-6 design of the fuel handling system (7 .1 .2 )

E.7.V-7 reactor disassembly and servicing equipment for

Chapter 8 - Plant Coolinq Water Systems

dent ( 6 . 5 )

BWRs (7 .5 )

E.8.V-1

E . 8 . V - 2 reduction o f surveillance testing (3.2)

E.8.V-3 availability of emergency power supply for the fuel pool cooling and

Chapter 9 - Site Support Systems

E.9.V-1 fire protection review (3 )

pump minimum flow line or recirculation line design (3 .2 )

I cleanup system following a design-basis accident (9 )

E.9.V-2 fire hazard analysis (3 .2 .2 )

E.9.V-3 smoke removal capability (3 .3 .1 )

E.9.V-4 security hardware on fire doors (3 .3 .1 )

E.9.V-5 separation of redundant shutdown equipment in the containment (3 .3 .1 )

E.9.V-6 control room cable fires (3 .4 .9 )

E.9.V-7 security area devitalized during unit shutdown (5 .1 )

E.9.V-8 operability of safety-related systems in areas with shared HVAC systems (8 .2 .1 )

E.9.V-9 criteria for design of HVAC ductwork (8 .2 .1 )

E.9.V-10 HVAC design for PWR auxiliary building (8 .2 .5 and 8 . 4 . 4 )

E.9.V-11 HVAC design for miscellaneous areas ( 8 . 2 . 6 )

E.9.V-12 charcoal filters in containment purge system (Branch Technical Position CSB 6-4, NUREG-0800) (8 .4 .2 )


E.9.V-13 design, equipment, and instrumentation for laboratories (9)

E.9.V-14 determination of airborne iodine concentration during an accident (Item I I I .D .3 .3 of NUREG-0737) (9)

Chapter 10 - Man-Machine In te r face Systems

E.10, V - 1

E. 10. V-2

E. 10. V-3

E. 10. V-4

E. 10. V-5

E. 10. V-6

E. 10. V-7

E.10.V-8

E.10.V-9

E. 10. V-10

E. 10. V-11

E.10.V-12

E. 10. V-13

E.10.V-14

E.10.V-15

E.10.V-16

E. 10. V-17

E.10.V-18

E. 10. V-19

E. 10. V-20

E. 10. V-21

E. 10. V-22

software protection (2.3)

level of automation (2.3)

review o f equipment used for displays to the operator (2.3)

methods to ensure operator alertness (2.3)

additional criteria for developing technology (2.3)

independence of verification and validation review teams

use of commercial compilers for software used in safety systems (3.1.2)

dedication of commercial-grade software (3.1.2 and 6.1.2)

use of commercial-grade equipment (3.1.2)

complexity of M-MIS (3.1.3)

clarification of requirements for analysis and validation testing

use of unproven techno1 ogy (3.2.2)

operator aids (3.4.5)

quantitative re1 i abi 1 i ty criteria (3.5)

establ i shment and use of re1 i abi 1 i ty and avai 1 abi 1 i ty estimates (3.5)

selection of equipment failure modes (3 .5.1 and 6.2.7)

maintenance frequency (3.5.2)

reliability analysis (3.5.4)

component reliability of M-MIS (3.5.4)

overall reliability of M-MIS (3.5.4)

minimum tests for continuous on-line testing (3.6.1)

automatic reconfiguration after failure detection (3.6.4)

(3.1.2)

Of M-MIS (3.1.3)


E. 10.V-23

E. 10. V-24

E.lO.V-25

E. 10.V-26

E. 10. V-27

E. 10.V-28

E.lO.V-29

E. 10. V-30

E.lO.V-31

E. 1O.V-32

E. 10.V-33

E.lO.V-34

E.lO.V-35

E.lO.V-36

E. 10.V-37

E. 10. V-38

E.lO.V-39

E.lO.V-40

E.lO.V-41

E. 10-V-42

E.lO.V-43

E. 10.V-44

E. 10.V-45

E. 10.V-46

surveillance period of automatic testing features (3.6.8)

automatic bypass initiation (3.6.10, 3.6.13, and 3.6.14)

module software concerns (3.7.4)

bypass and test lockouts during on-line repairs (3.7.6)

guidance on use of simulators and mockups (4.1.3)

vulnerability of power supplies for alarm systems (4.3.1)

a1 arm suppression techniques (4.3.3)

guidance on criteria to establish priorities (4.3.4)

guidance on the maximum number of alarms (4.3.4)

guidance on frequency allocation plan (4.6)

guidance on interference between communication systems and M-MIS equipment (4.6)

unauthorized access to equipment in remote shutdown stations (4.9.1)

guidance on inadvertent actuation of controls at local control stations (4.9.2)

design of emergency operations facility (4.9.4)

modification o f security boundaries during an emergency (4.9.4)

data storage methods (4.9.4)

compliance of perimeter intrusion alarm system with 10 CFR 73 .55 (h ) (5.2.1)

guidance on data system characteristics (5.2.2)

signal transport delay (5.2.5)

acceptability of digital-to-analog and analog-to-digital convertors (5.7)

software requirement specification (6.1.2)

verification of software (6.1.2)

documentation of testing and verification of commercially available software (6.1.2)

acceptance testing of commercially available software (6.1.2)


E.10.V-47

E.10.V-48

E. 1O.V-49

E.10.V-50

E. 1O.V-51

E.10.V-52

E.10.V-53

E. 10. V-54

E. 10. W-55

E.10.Y-56

E.10.V-57

E. 10. V-58

E. 1O.V-59

E.10.V-60

E.10.V-61

E.10.V-62

E. 13.V-63

E. 13.V-64

E. 10. V-65

E. 13.V-66

E. 13.V-67

E. 11). V-68

configuration control of software purchased through software clear- inghokes (6.1.2)

guidance on convolution of software structure (6.1.3)

behavior of commercial software when assumptions are violated (6.1.3)

guidance on memory protection (6.1.3)

-

of databases for redundant safety-related de separation (6.1.3)

definition dence (6.1

speci f i cat 6.2.3)

ices

of reasonable testing and sufficient degree of conf 5)

on of the level of diversity in safety systems (6.1

specific methods used to meet the requirement for diversity (6.1.6)

elimination of EM1 (6.2.2)

compatibility between M-MIS equipment and its external power supply systems (6.2.2)

signal validation methodology (6.2.2)

capacitance-type pressure sensors (6.2.5)

minimal acceptance review criteria for isolation device (6.2.6)

EMI/RFI considerations for wiring shields (6.2.9)

restoration state of control system components after loss of power (6.3.2)

setting resolution for control parameters (6.3.3)

requirements for signal reconstruction (6.3.3)

use of interrupts (6.3.3)

continuous self-testing of actuation logic (8.3.2)

radiation monitor placement, calibration frequency, and emergency power provisions (10.2..1)

compliance with Item II.F.1.3 of NUREG-0737 (10.2.1)

criteria for airborne radioactivity monitors (10.2.1)

Prolgram Summary 5-17

E. 10.V-69

E . 10.V-70

E. 10 .V-71

E. 10 .V-72

E.10.V-73

E. 1O.V-74

E. 10.V-75

E. 10.V-76 E.lO.V-77

E. 10. V-78

E. 10. V-79

E.lO.V-80

E.lO.V-81

14-day maintenance criteria for M-MIS for reactor protection system, plant control system, and plant information and monitoring systems (GSI 75) (Appendix B)

procedures t o assess unscheduled reactor shutdowns (GSI 75 ) (Appendix B)

safety imp1 ication of instrumentation and control systems (US1 A-47, GSI 76) (Appendix B )

inclusion of computer specialist on design and review teams (3.1 o f Appendix D)

establishment of Q-list and associated equipment list (GSI 75) (Appendix B)

handling of vendor interface (GSI 75) (Appendix B)

evaluate neutron monitoring system M-MIS (7.4)

reliable operation of reactor trip breaker (GSI 75) (Appendix B)

design of BWR water level instrumentation (GSI 101) (Appendix B)

operator training and emergency operating procedures concerning feed-and-bleed operations (GSI 122.2) (Appendix B)

human factors organization (3.1 of Appendix D)

acoustical environments in operating control areas (3.7.6 of Appendix D)

design reference documents include IEEE P1023/D5 (3.7.6 of Appendix D)

Chapter 1 1 - Electric Power Systems

E.ll.V-1 environmental qualification test criteria for electrical power system (2.2)

E.1l.V-2 safety classification o f loads (2.2.1)

E.1l.V-3 minimization of Class 1 E components (2.2.4)

E.1l.V-4 instrumentation and controls for electric motors (2.2.5)

E.1l.V-5 compliance with NFPA Codes and Standards (2.2.6)

E.11.V-6 integrity of electrical cable penetration seals during a fire

E.1l.V-7

E.1l.V-8

(2.2.6)

integrity of bus duct penetrations during a fire (2.2.6)

review o f IEEE standards not endorsed by regulatory guides (2.2.7)


E.1l.V-9 review of the actual setpoint criteria used for sizing thermal over1 oads (2.2.9)

E.11.V-10 limitation of total voltage distortion to 3 percent (4.2.4)

E.ll.V-11 effects of electrical faults on the coastdown capability of the reactor cool ant pumps and reactor internal pumps (4.2.5)

E.ll V-12 use of combustion turbine generator as alternate power source during shutdown (5.2.1)

E.ll V-13 continuous rating versus short-term rating for sizing the combustion turbine generator (5.2.3)

E.1l.V-14 inclusion of the pressurizer heaters in the diesel generator power analysis (5.2.4)

E.1l.V-15 continuous rating of the diesel generators to include emergency lighting (5.2.4)

E.1l.V-16 capability of the diesel generators to power safety buses in a protected bus configuration (5.2.5)

E.ll,V-17 emergency diesel engine starting system (5.2.6)

E.11. V-18 emergency diesel engine fuel oil storage and transfer system (5 .2.6)

E.1l.V-19 allowed outage time for load center (6.2)

E.ll,V-20 impact of loss of ac or dc bus on single-failure protection in safety-related systems (7.2.1)

E.1l.V-21 outage time for dc safety buses in a BWR plant design (7.2.2)

E.1l.V-22 common backup ac power sources for safety-related uninterruptible power suppl i es (7.2.4)

E.1l.V-23 design of the continuous ac lighting in safety-related areas and access routes outside the main control room (8.2.1 and 8.2.2)

E. 11 .V-24 method of integrating the emergency 1 ighting system with the normal lighting in the main control room (8.2.3)

E.1l.V-25 acceptability of lighting system for closed-circuit television system (8.2.4)

ChaDter 12 - Radioactive Waste Processinq Systems E.12.V-1 inputs and releases from the radioactive waste processing systems

(2.2.1)

E.12.V-2 use of demonstrated technology (2.2.1)

Progr,am Summary 5-19

E. 12. V-3

E.12.V-4

E.12.V-5

E. 12 .V-6

E. 12. V-7

E.12.V-8

E.12.V-9

E.12.V-10

E.12.V-11

offsite dose calculation manual (2.2.1)

fuel source term parameters for design of radioactive waste processing systems (2.2.2)

estimate of personnel radiation exposure (2.2.4)

control, monitoring and sampling of liquid and radioactive waste processing and effluent streams (2.2.9)

interface between BWR HVAC systems and GRWP systems (3.3.2)

use of HEPA filters downstream o f charcoal adsorbers (3.3.3) potentially explosive mixtures o f hydrogen and oxygen (3.3.4)

piping layout and design and operating procedures for filters and ion exchangers in liquid radioactive waste processing systems (4.2)

shipping container design (5.5)

Chapter 13 - Main Turbine-Generator Svstem

E. 13.V-1

E.13.V-2

E. 13 .V-3

E.13.V-4

E.13.V-5

E. 13.V-6

E. 13 .V-7

E. 13. V-8

E. 13 .V-9

E. 13 .V-10

E.13.V-11

60-year design life for major components of the main turbine- generator (2.2)

use o f seismic experience data base for seismic qualification (3.1.1)

performance and safety requirements for main turbine (3.1.3)

turbine maintenance program

effect of other duty cycles on long-term integrity of the turbine (3.1.4)

facture toughness properties of turbine casing material (3.1.4)

part-machining inspection of one-piece rotor (3.1.5)

need for prototype testing of new or significantly changed designs (3.1.6 and 4.1.1)

adequacy of turbine control system (3.3)

inservice inspection intervals for main stop and control valves and reheat stop and intercept valves (3.3)

seal clearances of gland seal system (3.4)


5.2 Vendor- or Utilitv-Specific Items Pertaininq to the Requirements Document for Passive Plant Desiqns

The following is a list of vendor- or utility-specific items that were identified in the DSER on the Requirements Document for passive plant designs.

Chapter 1 - Overall Reauirements P.1.J-1

P. 1 .,!-2

P. 1 .'I-3

P. l . l l -4

P. 1 . '1-5

P. 1 .'1-6

P. 1 .'1-7

P. 1 . \l-8

Q. 1 .\1-9

P. 1 .\!-lo

P.1.V-11

P. 1 .\'-12

P. 1 .\I-13

P. 1 .\'-14

P. 1 .\I-15

P . 1 . C'-16

P. 1 .\'-17

P. 1 .\'-18

,P. 1 . i -19

implementation of design characteristics intended to enhance accident resistance (2 .2 )

bounding analysis by standard site design parameters (2 .3 .1 )

selection of initiating events and their frequency categorization (2 .3 .2 )

acceptance criteria for transient and accident analysis (2 .3 .2 )

passive plant anticipated transient without scram response analysis (2 .3 .2 )

operator actions 72 hours after accident ( 2 . 3 . 2 )

use o f 72-hour design basis (2 .3 .2 )

technical basis for severe-accident management program and emergency operating procedures guidelines (2 .3 .4 )

acceptability of analytical codes and methodologies for safety analysis (2 .5 )

defense-in-depth analysis (2 .5 and 3 . 5 )

60-year plant life (3.3, 4.8 .2 , 8 . 2 and 11.3)

operation of PWR with a secured reactor coolant pump (3 .5 )

fuel burnup requirements (3 .6 )

extended operating life of control blades and control rod assemblies (3 .6 )

safety classification (4 .3 .1 )

seismic qualification by experience ( 4 . 3 . 2 and 4.8.1)

non-sei smic building structures (4 .3 .2 .3 and 4.7 .2 .10)

structural design and construction codes (4.4 and 4.4.1)

elimination of operating-basis earthquake from design (4 .4 .3 , 4 .7 .3 , and Appendix B)

Program Summary 5-2 1

P.1.V-20 definition of support group (4.4.3)

P.1.V-21 use of Appendix N of ASME Code, Section I 1 1 (4.4.3 and 4.7.3)

P.1.V-22 analysis of vibratory loads with significant high-frequency input (4.4.3)

P.1.V-23 use o f nonlinear analysis to account for gaps between pipes and piping supports (4.4.3)

P.1.V-24 probabilistic approach for changing existing loads and/or loading combinations (4.5.1)

P.1.V-25 recurrence interval for wind loadings (4.5.2.1)

P.1.V-26 maximum ground, water level (4.5.2.2)

P.1.V-27 precipitation for roof design (4.5.2.2)

P.1.V-28 snow loading (4.5.2.2)

P.1.V-29 detailed quantification of soil parameters (4.5.2.3)

P.1.V-30 minimum margin against liquefaction (4.5.2.3)

P.1.V-31 external hazards evaluation (4.5.2.3)

P.1.V-32 number of full-stress cycles (4.5.2.4 and 4.8.1)

P. 1 .V-33 si te-speci fic safe shutdown earthquake (SSE) (4.5.2.4)

P.1.V-34 power spectrum density function of the time history (4.5.2.4)

P.1.V-35 design temperature (4.5.2.7)

P.1.V-36 protection against surface vehicle bombs (4.5.3)

P.1.V-37 design against internal-missile generation (4.5.5)

P.1.V-38 design o f concrete containment (4.6.1.1)

P.1.V-39 load combinations for Category I buildings and structures (4.6.1.2)

P.1.V-40 design of Category I steel structures (4.6.1.2)

P.1.V-43 combination o f pipe rupture loads with seismic loads for seismic Category I structures (4.6.1.3 and 4.6.1.4)

P.1.V-42 combination of LOCA and SSE loads (4.6.1.7)

P.1.V-43 load combinations for safety-related portions of the plant (4.6.2)

P.1.V-44 dynamic analysis techniques (4.7.2.3)


P. 1. \'-45

P. 1 . \'-46

P. 1. \'-47

P. 1. \'-48

P. 1 .\'-49

P. 1 .\I-50

P. 1 .\I-51

P. 1 .\'-52

P. 1 .\I-53

P. 1.1'-54

P. 1. \'-55

P. 1. \'-56

P. 1.1'-57

P. 1 .\'-58

P. 1 .\I-59

P. 1 .\'-60

P. 1 .\'-61

P. 1 . \'-62

P. 1. \'-63

P. 1 .\I-64

P. 1 . \I-65

P. 1. \'-66

P. 1 .\'-67

P. 1.C'-68

P. 1.1'-69

methodology for generating design spectra or time histories (4.7.2.5)

structural damping values (4.7.2.6)

masonry walls in Category I buildings (4.7.2.7)

use of expansion anchor bolts - compliance with Office of Inspection and Enforcement Bulletin 79-02 (4.7.2.8 and 4.7.3)

stability of shell-type structures under compression (4.7.2.9)

use of ASME Code Cases N-411 and N-420 in same analysis (4.7.3)

use of ASME Code Case N-411 (4.7.3)

construction of core support structures (4.7.3)

design fatigue curves (4.7.3)

use of zinc to reduce radiation fields (5.2.7)

grinding controls for PWRs (5.3.1.1)

use of Alloy 600 (5.3.1.3)

effect of fabrication processes on intergranular stress corrosion cracking (5.3.1.8)

selection of seal, gaskets, and protective coatings (5.3.5)

aging of cable insulations and other electrical materials (5.3.6)

use o f hydrogen water chemistry for the advanced BWR design (5.5.2)

plant-specific reliability assurance program (6.5)

inspection of construction activities (7 and 11.13)

installed operating-phase security system (7 )

reliability o f modular construction (7)

inspection and verification of security locks robotically (8.3)

compliance o f FDA/DC applications with Commission's regulations and guidance (10)

issue resolution for FDA/DC reviews (10)

inspections, tests, analyses, and acceptance criteria (10)

implementation of simplification objective (11.4)


P.l.V-70 implementation of standardization objective (11.5)

P.l.V-71 inservice testing requirements for the essential non-safety-related components (12.2.1 and 12.2.3)

P.l.V-72 quarterly testing of pumps and valves (12.2.2)

P.l.V-73 check valve testing methods (12.2.2)

P.l.V-74 full-flow testing of check valves (12.2.2)

P.l.V-75 provisions to test hydraulically and pneumatically operated valves under design-basis differential pressure and flow (12.2.2)

qualification testing of active and non-active motor-operated valves (MOVs) (12.2.2)

P.l.V-76

P.l.V-77 technical concerns regarding MOVs (12.2.2)

P.l.V-78 leak rate testing for individual containment isolation valve

P. 1 .V-79

Appendix A to ChaDter 1 - PRA Key AssumDtions and Groundrules

(12.2.2)

frequency and extent of disassembly and inspection of safety-related pumps (12.2.3)

P. l A . V - 1

P. lA.V-2

P. lA.V-3

P. lA.V-4

P. lA.V-5

P. lA.V-6

P. 1A.V-7

P. lA.V-8

P. 1A.V-9

P. lA.V-10

long-term decay heat removal in the PRA (1.6)

justification of mission items and success criteria (2.10)

reliability data (2.11)

review of core-damage-sequence binning (4.1)

review of actual groupings of the accident sequences into plant damage states (4.2)

review of the evaluation of containment leakage paths (4.3)

computer codes for in-pl ant sequence assessment (4.4)

verification that the reference site parameters identified in Annex B are consistent with revised 10 CFR Part 100 (5.2)

differences in computer codes used for calculating offsite consequences (5.2)

source terms for representative accident sequences bounded by the physically based source term (6.3)


m e n d i x B to ChaDter 1 - Licensinq and Requlatorv Reauirements and Guidance P.1B.V-1 compliance of FDA/DC applications with Commission’s

regulations and guidance (1.2)

P.13.V-2 issue resolution for FDA/DC reviews (1.2)

P.13.V-3 dynamic seismic analysis of main steam piping and condenser (2.3.1 and Item 1I.E of Annex A)

Annex A)

Annex A)

and Item 1II.F of Annex A)

P.13.V-4 main steamline piping classification (2.3.1 and Item 1I.E of

P.113.V-5 seismic analysis of turbine building (2.3.1 and Item 1I.E of

P.113.V-6 plateout considerations for main steam piping and valves (2.3.1

P.lI3.V-7 reactor vessel level instrumentation system (2.4.1)

P. lI3.V-8

P. lI3.V-9

P. l I ! . V - l O

P.1D.V-11

P. 1D.V-12

P.1tl.V-13

P . l I l . V-14

P. 1El.V-15

P. 1El.V-16

dedicated containment vent penetration (2.5.4 and Item 1.K o f Annex A)

fission product leakage control (2.5.2 and Item 1II.F of Annex A)

fission product cleanup system (2.5.2 and Item 1II.F of Annex A )

automatic emergency core cooling system switch to recirculation (Generic Safety Issue 24) (3.2.11)

bolting degradation or failure (Generic Safety Issue 29) (3.2.12)

detached thermal sleeves (Generic Safety Issue 73) (3.2.15)

failure o f high-pressure coolant injection steamline without i so la - tion (Generic Safety Issue 87) (3.2.20)

adequate 1 eak test requirements for pressure i sol at i on Val ves in inservice testing programs and technical specifications (3.2.22)

identification and design of interfacing systems (3.2.22)

P.1El.V-17 implementation o f procedures for electrical power reliability (3.2.27)

P.1E8.V-18 leakage through electrical isolators in instrumentation circuits (Generic Safety Issue 142) (3.2.29)

P . l E .V-19 availability of chilled water systems and room cooling (Generic Safety Issue 143) (3.2.30)

P.1El.V-20 reliability of recirculation pump trip during an anticipated transient without seram (Generic Safety Issue 151) (3.2.31)


P.lB.V-21

P.lB.V-22 spectral peak shifting (3 .3 .1 )

ChaDter 2 - Power Generation Svstems

P.2.V-1

P.2.V-2 attachment loads for PWR safety and relief valves (3 .4 )

P.2.V-3 design adequacy of side stream polisher (4 .3 )

ChaDter 3 - Reactor Coolant System and Reactor Non-Safetv Auxiliarv Svstems

loss of emergency service water system (Generic Safety Issue 153) (3 .2 .32)

turbine bypass system flow capacity (3 .2 )

P.3.V-1

P .3. V-2

P.3.V-3

P.3.V-4

P.3.V-5

P.3.V-6

P.3.V-7

P.3.V-8

P.3.V-9

P. 3 . V-10

P. 3 .V-11

P. 3 . V-12

P. 3 . V- 13

P. 3 . V-14

vent and drain design (2 .1.2)

reactor coolant interface systems design (2 .1 .7 , 3 .1 , and 5.1)

specific performance requirements and acceptance criteria for active non-safety auxiliary systems (2.2, 6, and 8)

safety analyses of the abnormal conditions associated with loss o f a feed pump or load rejection (3 .2 )

acceptability of operation with reactor coolant pump(s) out of service (3 .2 )

low-temperature overpressure protection system design details (3 .3 )

design analyses to confirm the capability and reliability of the passive decay heat removal system ( 3 . 4 )

specific design requirements for reactor vessel level instrumentation system ~ (3 .5 )

design details of the man-machine interface system for steam generator water 1 eve1 control (4 .5 )

design details to mitigate excessive leakage of'main steam isolation valves (5 .4 )

design details for automatic reactor vessel overfill protection (5 .5 )

design details for PWR auxiliary systems, including chemical and volume control system, against criterion in Standard Review Plan Section 9.3.4 (6)

design details for reactor shutdown cooling pump seals (9 )

adequate vendor assessment of shutdown and low-power-operation risk (9 )


Chapter 4 - Reactor Svstems

P. 4 .\l-l

P. 4 .\'-2

P. 4. \'-3

P. 4. \'-4

P .4. \'-5

P. 4 .\I-6

P. 4.\'-7

P. 4. \'-8

P. 4 .ir-9

P. 4. \ ' - l o P.4. i-11

P.4.b-12

P. 4 .b-13

P.4.U-14

P.4.U-15

P.4.V-16

P.4.V-17

P.4.V-18

P. 4. V-19

P.4 .V-20

boration of water for fue handling and storage (2.2.8)

fuel assembly reconstitution (2.3.3)

condensation carryunder limitation (3.2)

experimental data for divided-chimney design (3.2 and 4.2.1)

BWR stability (2.2.4 and 4.2.1)

decay ratio limits and analysis methods and procedures (4.2.1)

LOCA analysis methodology (4.2.1 and 7.2.1)

1 oad-foll owing and maneuvering capabi 1 i ty (4.2.1 and 7.2.1)

methodology to achieve BWR stability (4.2.1)

fuel burnup requirements (4.2.2 and 7.2.2)

two-cycle fuel channel lifetime (4.2.4)

control rod assembly lifetime (4.2.6 and 7.2.3)

control rod scram time (5.2)

control rod assembly malfunctions in BWR accident analyses (5.2)

protection of scram pilot solenoid valves (5.3)

thermal shield removal (6.3)

reactor pressure vessel level instrumentation (6.3)

negative moderator temperature coefficient 1 imit (7.3)

fuel rod bow .penal ties (7.3)

control rod drive mechanism lifetime (8.2)

Chapter 5 - Enqineered Safetv Svstems

P.5.V-1

P.5.V-2

P.5.V-3

P.5.V-4 LOCA calculations justifying removal of core spray system (4.1)

P.5.V-5 manual standby liquid control system initiation (4.5)

decontamination factor for containment system (2.1.7)

challenge from inadvertent opening of the DPS (2.2)

safety-grade provisions for the fire protection system (2.3)


P. 5 . V-6

P.5.V-7

P.5.V-8

P.5.V-9

P. 5 . V-10

P. 5 .v-11

system design to minimize condensation water hammer (4 .2 )

in-containment refueling water storage tank boiling suppression (5 .3 )

justification for use of remote manual valve for containment isolation (6 .2 )

evaluation of the ignition system for combustible gas control (6 .6 )

reliability of power supplies for severe-accident equipment (6 .7 )

detailed discussions regarding design-basis-accident events (9 )


P.6.V-1 method for inspecting structural degradation (2 .1 .1 )

P.6.V-2 evaluation of the engineering backfill ( 2 . 1 . 1 )

P.6.V-3 use o f American National Standards Institute (ANSI) 10.4-1987

P.6.V-4 structural modules to be used in construction (2 .2 )

P.6.V-5

(2 .1.1)

redundant non-safety-grade auxiliary systems within the plant protected area (2 .3 .3 )

P.6.V-6 review of site-unique security and contingency plans (2 .3 .6 )

P.6.V-7 programs for controlling and storing toxic materials (2 .3 .7 )

P.6.V-8 shielding design requirements, shielding computer codes, and radioactive material sources (2 .4 )

P.6.V-9 use of ANSI/American Institute of Steel Construction N-690 (4 .1 .2 and 4.1 .3)

P.6.V-10 evaluation of potential high-radiation areas (4 .1 .7 )

P.6.V-11 design of the common basemat ( 4 . 1 . 9 )

P.6.V-12 reinforced-bar design criteria for vinyl-coated rebars (4 .1 .9 )

P.6.V-13 material control provisions inside the containment (4 .2 .2 and 4 .2 .3 )

P.6.V-14 control of access to the reactor containment (4 .2 .2 and 4 .2 .3 )

P.6.V-15 design features that preclude potentially lethal radiation levels

P.6.V-16 containment design details for aerosol and radioactive gases

( 4 . 2 . 2 )

(4 .2 .3 )


P.6.\'-17 use of ANSI/ASME NOG-1, 1983 (4.2.3)

P.6.\'-18 des gn of the control complex (4.5.4)

ChaDler 7 - Fuelinq and Refuelins Systems

P.7.C'-1 design of the overhead bridge crane (2.3.2)

P.7.k-2 high-radiation areas (2.3.7)

P.7.b-3 reactor disassembly and servicing equipment for BWRs (3.1.2

ChaDt er 8 - P1 ant Cool inq Water Systems

P.8.V-1

P.8.V-2 time-delay allowance for fuel pool cooling capability (9)

ChaDter 9 - Site Support Systems

design requirements for the chilled water system (8)

P.9.V-1

P.9.V-2

P.9.V-3

P.9.V-4

P .9 .v-5

P. 9.V-6

P. 9. v-7

P. 9 .V -8

P. 9. v,-9

separation of redundant shutdown equipment in the containment (3.3.1)

underfloor or ceiling control room cable fires (3.4)

security area devitalized during unit shutdown (5.1)

security for components that manipulate vital isolation valves (5.2.1)

non-safety-related auxiliary systems within the protected area (5.2.1)

sabotage vu1 nerabi 1 i ty analysi s (5.2.2)

charcoal filters in air filtration systems (8.2.4, 8.2.5, 8.3.3, 8.4.1, and 8.4.2)

safety classification of fuel facility ventilation supply subsystem (8.2.4)

safety classification of PWR auxiliary building ventilation supply subsystem (8.4.2)

ChaDtlzr 10 - Man-Machine Interface Systems (M-MIS)

P. l O . ' / - l acceptable interpretations o f requirements (1)

P.10.'/-2 software protection (2.3)

P.1O.V-3 level of automation (2.3)

P.1O.V-4 defense-in-depth and diversity analysis (2.3 and 4.5)


P. 10 .V-5

P. 10.V-6

P. 10 .V-7

P.lO.V-8

P. lo.v-9

P. lo.v-10

P.1O.V-11

P.1O.V-12

P. 10.V-13

P. 10.V-14

P.lO.V-15

P.lO.V-16

P.lO.V-17

P .lo. V- 18

P .lo. v-19

P. 10 .v-20

P .lo. v-2 1

P .lo. v-22

P .lo. V-23

P .lo. V-24

P .lo. V-25

P.lO.V-26

P. 10.V-27

P. 1O.V-28

P.lO.V-29

rev iew o f equipment used f o r d i s p l a y s t o t h e opera tor (2.3)

methods t o ensure opera tor a le r tness (2.3)

a d d i t i o n a l c r i t e r i a f o r develop ing technology (2.3)

independence o f v e r i f i c a t i o n and v a l i d a t i o n rev iew teams (3.1.2)

use o f commercial-grade equipment (3.1.2)

complex i ty o f M-MIS (3.1.3)

use o f unproven techno1 ogy (3.2.2)

q u a n t i t a t i v e r e l i a b i l i t y c r i t e r i a (3.5)

s e l e c t i o n o f equipment f a i l u r e modes (3.5.1 and 6.2.7)

maintenance frequency (3.5.2)

r e l i a b i l i t y ana lys i s (3.5.4)

automat ic r e c o n f i g u r a t i o n a f t e r f a i l u r e d e t e c t i o n (3.6.4)

s u r v e i l l a n c e p e r i o d o f automat ic t e s t i n g f e a t u r e s (3.6.8)

automat ic bypass i n i t i a t i o n (3.6.10, 3.6.13, and 3.6.14)

module so f tware concerns (3.7.4)

bypass and t e s t l ockou ts d u r i n g o n - l i n e r e p a i r s (3.7.6)

main c o n t r o l room s t a f f i n g (4.2)

alarm suppression techniques (4.3.3)

use o f "d ia l -up " te lephone-type p o r t a b l e r a d i o s f o r s e c u r i t y purposes (4.6.3)

unauthor ized access t o equipment i n (4.9.1)

computer room w i t h i n t h e main c o n t r (4.9.1)

des ign o f emergency opera t ions f a c i

m o d i f i c a t i o n o f s e c u r i t y boundaries

da ta s to rage methods (4.9.4)

remote shutdown s t a t i o n s

1 room s e c u r i t y boundary

i t y (4.9.4)

d u r i n g an emergency (4.9.4)

compliance o f per imeter i n t r u s i o n alarm system w i t h 10 CFR 73.55(h) (5.2.1 and 5.2.5)


P.10, V-30 signal transport delay (5.2.5)

P.lO,V-31 analog-to-digital and digital-to-analog converters (5.7)

P.lO,V-32 software requirement specification (6.1.2 and 6.1.6)

P.1O.V-33 verification of software (6.1.2)

P.lO.V-34 documentation of testing and verification of commercially available software (6.1.2)

P.lO.V-35

P.lO.V-36

P. 10.V-37

P. 10.V-38

P. 10.V-39

P.lO.V-40

P. 10.V-41

P. 10.V-42

P. 10.V-43

P. 10. V-44

configuration control of software purchased through software clearinghouses (6.1.2)

specific methods used to meet the requirement for diversity (6.1.6)

elimination of electromagnetic interference (6.2.2)

signal Val idation methodology (6.2.2)

restoration state o f control system components after loss of power (6.3.2)

setting resolution for control parameters (6.3.3)

neutron monitoring M-MIS (7.4)

selection of variables for automatic actuation (8.2.3)

radiation monitor placement, cal ibration frequency, and emergency power provisions (10.2.1)

compliance with Item II.F.1.3 of NUREG-0737 (10.2.1)

P.lO.V-45 criteria for airborne reactivity monitors (10.2.1)

P.lO.V-46 operating philosophy ( 2 o f Appendix B)

P.lO.V-47 use of mockups, prototypes, and simulators (2 of Appendix B)

ChaDt er 11 - .El ectri c Power Svstems

P . l l . V - 1 reliance on the non-safety electrical systems beyond 72-hour period (2.2.2)

P.1l.V-2 reliance on the non-safety electrical systems to achieve cold shutdown (2.2.2)

P.1l.V-3 applicable regulations and regulatory guidances that are not addressed in Chapter 11. (2.2.3)

minimization of Class 1E components (2.2.4) P . l l V - 4

Progr 3rn Summary 5-3 1

P.1l.V-5

P.1l.V-6 operating conditions of all plant loads for all relevant grid

use of revisions of I E E E standards not endorsed by the staff (2 .2 .7 )

conditions and the design of the bus voltage protection schemes (3 .2 .4 )

design of the standby power source starting system (5 .2 .6 ) P.1l.V-7

P.1l.V-8 design of the standby power source fuel oil storage and transfer system (5 .2 .6 )

P.1l.V-9 design of the electrical separation of dc and vital ac power supply systems (7 .2 .5 )

P.11.V-10 integration of the exterior lighting system with the closed-circuit tel evi si on system (8 .2 .4 )

Chapter 12 - Radioactive Waste Processinq Svstems

P.12.V-1

P. 12.V-2

requirements for radioactive waste processing systems and effluent p a t h s (2.2.8)

gaseous radioactive waste processing system hydrogen control design (3 .3 .4 )

P.12.V-3 design of dry waste shipping containers (5 .5 )

Chapter 13 - Main Turbine-Generator Svstems

P.13.V-1 use of seismic experience data base and analysis for SSE loading conditions (3 .1 .1 )

P.13.V-2 one piece rotor design part machining inspection requirements (3 .1 .5 )

P. 13.V-3 prototype testing of new or significantly changed turbine-generator designs (3 .1 .6 and 4 .1 .1 )


6 CONCLUSION

w u i r e m e n t s Document f o r Evol u t i o n a r v P1 an t Desiqns

Sub jec t t o t h e r e s o l u t i o n o f t h e i d e n t i f i e d ou ts tand ing p o l i c y i ssues and vendor- and u t i l i t y - s p e c i f i c i tems d iscussed i n t h e SER (Volume 2 o f t h i s r e p o r t ) , t h e s t a f f concludes t h a t t h e requi rements e s t a b l i s h e d i n t h e Require- ments Document f o r e v o l u t i o n a r y p l a n t designs (Volume 11) do n o t c o n f l i c t w i t h c u r r e n t r e g u l a t o r y guide1 ines and are acceptable. However, by themselves they do no t p r o v i d e s u f f i c i e n t i n f o r m a t i o n f o r t h e s t a f f t o determine i f t h e p l a n t des ign w i l l be adequate. Therefore, a p p l i c a n t s r e f e r e n c i n g t h e Requirements Document w i l l be r e q u i r e d t o demonstrate compliance w i t h t h e a d d i t i o n a l guidance p rov ided i n t h e Standard Review Plan (NUREG-0800), o r p rov ide j u s t i - f i c a t i o n f o r a l t e r n a t i v e means o f implement ing t h e assoc ia ted r e g u l a t o r y r e q u i remen t s . I n the s t a f f requi rements memorandum (SRM) o f August 24, 1989, t h e Commission i n s t r u c t e d t h e s t a f f t o p rov ide an a n a l y s i s d e t a i l i n g where t h e s t a f f proposes depar tu re f rom c u r r e n t r e g u l a t i o n s o r where t h e s t a f f i s s u b s t a n t i a l l y supplement ing o r r e v i s i n g i n t e r p r e t i v e guidance a p p l i e d t o c u r r e n t l y l i c e n s e d l i g h t water r e a c t o r s (LWRs). Appendix B t o Chapter 1 o f Volume 2 o f t h i s r e p o r t g i ves t h e s t a f f ' s regu la - t o r , y a n a l y s i s o f those issues i d e n t i f i e d f o r these designs. These issues have been addressed i n SECY-90-016 and SECY-91-078, and i n d r a f t Commission papers, " Issues P e r t a i n i n g t o Evo lu t i ona ry and Passive L i g h t Water Reactors and T h e i r Re1 a t i o n s h i p t o Cur ren t Regulatory Requirements," and "Design C e r t i f i c a t i o n and L i cens ing Issues P e r t a i n i n g t o Passive and E v o l u t i o n a r y Advanced L i g h t W a t w Reactor Designs," t h a t were issued on February 27 and J u l y 6, 1992, r e s p e c t i v e l y .

The s t a f f cons iders these t o be p o l i c y issues.

I n i t s SRMs dated June 26, 1990, and A p r i l 15, 1991, t h e Commission prov ided i t s d e c i s i o n s on SECY-90-016 and SECY-91-078 as they app ly t o e v o l u t i o n a r y designs . The February 27 and J u l y 6, 1992, d r a f t Commission papers have been forwarded t o the Adv isory Committee on Reactor Safeguards. The s t a f f w i l l i n c l u d e i t s views i n t h e f i n a l papers and document i t s f i n a l p o s i t i o n s be fo re seeking Cominission approval . When t h e s t a f f f i n a l i z e s these Commission papers, t h e Cominission w i l l complete i t s rev iew o f t h e b a s i s f o r t h e approach t h a t t h e s t a f f i s p ropos ing f o r those issues and, accord ing ly , may a t some f u t u r e p o i n t i n the rev iew determine t h a t such issues i n v o l v e p o l i c y ques t ions t h a t t h e Cominission may w ish t o cons ider . The approaches t o r e s o l v i n g these issues h a w n o t been rev iewed by t h e Commission, and, t h e r e f o r e , do n o t represent age'icy p o s i t i o n s .

Therefore, t h e s t a f f concludes t h a t EPRI's ALWR U t i l i t y Requirements Document f o r e v o l u t i o n a r y p l a n t des igns (Volume 11) s p e c i f i e s requi rements t h a t , sub, ject t o t h e r e s o l u t i o n o f t h e i d e n t i f i e d ou ts tand ing p o l i c y i ssues and venljor- and u t i l i t y - s p e c i f i c i tems, i f p r o p e r l y t r a n s l a t e d i n t o a des ign and cons t ruc ted and operated i n accordance w i t h the NRC r e g u l a t i o n s i n f o r c e a t t h e t i m e t h e des ign i s submit ted, should r e s u l t i n a nuc lea r power p l a n t t h a t

Pro l~ram Summary 6- 1

will have all the attributes required to ensure that there is no undue risk to the health and safety of the public or to,the environment. complying with existing regulations, such a facility would also be consistent with the Commission’s policies on severe-accident protection.

Reauirements Document for Passive P1 ant Desiqns

In addition to

Subject to the resolution of the identified outstanding issues and vendor- and utility-specific items listed in Sections 1.4 and 1.5 of each DSER chapter or appendix issued on April 24, 1992, the staff concludes that the requirements established in the Requirements Document for passive plant designs (Volume 111) do not conflict with current regulatory guidelines and are acceptable. However, b,y themselves they do not provide sufficient information for the staff to determine if the plant design will be adequate. Therefore, appli- cants referencing the Requirements Document will be required to demonstrate compliance with the additional guidance provided in the Standard Review Plan (NUREG-0800), or provide justification for alternative means of implementing the associated regulatory requirements.

In its August 24, 1989, SRM, the Commission instructed the staff to provide an analysis detailing where the staff proposes departure from current regulations or where the staff is substantially supplementing or revising interpretive guidance applied to currently licensed LWRs. policy issues. Appendix B to the DSER on Chapter 1 of the Requirements Document for passive plant designs gives that analysis. these issues to the Commission in draft Commission papers dated February 27 and July 6, 1992. When the staff finalizes these Commission papers, the Commission will complete its review of the basis for the approach that the staff is proposing for those issues and, accordingly, may at some future point in the review determine that such issues involve policy questions that the Commission may wish to consider. The approaches to resolving these issues have not been reviewed by the Commission, and, therefore, do not represent agency positions.

The s t a f f considers these to be

The staff forwarded

In addition, certain technical issues still have to be resolved before the staff can complete its review.

The following conclusions are based on the staff’s review as documented in the April 1992 draft SER (Volume 3 of this report). The final SER, scheduled to be issued in September 1993, will give the final results of the staff’s review of EPRI’s ALWR Utility Requirements Document for passive ALWR designs.

Therefore, on the basis of its review to date, the staff concludes that EPRI’s ALWR Uti1 ity Requirements Document for passive plant designs (Volume 111) specifies requirements that, subject to the resolution of the identified outstanding issues and vendor- and uti1 ity-specific items, if properly translated into a design and constructed and operated in accordance with the NRC regulations in force at the time the design is submitted, should result in a nuclear power plant that will have all the attributes required to ensure that there is no undue risk to the health and safety of the public or to the environment. In addition to complying with existing regulations, such a facility would also be consistent with the Commission’s policies on severe- accident protection.

-.

Program Summary 6-2

APPENDIX A

CHRONOLOGY OF CORRESPONDENCE

This appendix contains a chronological listing of routine licensing correspondence between the U . S . Nuclear Regulatory Commission (NRC) staff and the Electric Power Research Institute (EPRI) and other correspondence related to Project 669.

July 14, 1981

September 15, 1982

October 20, 1982

October 27, 1982

December 20, 1982

January 21, 1983

January 27, 1983

Prcgram Summary

Letter from J. C. Mark, Advisory Committee on Reactor Safeguards (ACRS) , to NRC submitting suggestions regarding potential safety improvements for incorporation into new designs for nuclear power plants. give appropriate priority and resources to developing safety requirements for future LWRs.

NRC should

Letter from S. Burstein, EPRI, to NRC discussing proposed NRC involvement in steering committee program for developing standardized LWR design as discussed at July 20, 1982, meeting . committee and program group.

Participation requi red i n pol icy

Letter from H. R. Denton, NRC, to EPRI, responding to September 15, 1982, letter requesting NRC participation in EPRI’s Standardized LWR Design Program. involvement in program to prove beneficial to both NRC and EPRI. Meeting requested.

NRC

Letter from C. 0. Thomas, NRC, to EPRI, summarizing October 21, 1982, meeting with E P R I regarding LWR standardization program. Development of LWR plant baseline designs and approach that could be used by vendors in obtaining certification of plant design emphasized in EPRI program.

Letter from S . Burstein, EPRI, to NRC, requesting that meeting between Utility Steering Committee for EPRI’s Standardized LWR Design Program and NRC Policy Committee be held on February 9, 1983, in Bethesda, Maryland, to discuss plans for standardized design. information enclosed.

Related

Letter from D. H. Moran, NRC, to EPRI, summarizing January 12, 1983, meeting with EPRI in Bethesda, Maryland, regarding review of licensing issues to be considered in LWR standardization program.

Letter from R . E. Nickell, EPRI, to NRC, confirming high- priority safety and licensing issues selected at meeting

A- 1

February 21, 1983

February 28, 1983

March 10, 1983

March 10, 1983

May 14, 1983

May 25, 1983

June 23, 1983

September 16, 1983

November 4 , 1983

Program Summary

on January 12, 1983, w i t h emphasis on r e v i s i o n o f Appendix R t o 10 CFR P a r t 50, decay heat removal (DHR), and h igh -s t reng th b o l t i n g . a t February 9, 1983, meet ing suggested.

Poss ib le i tems f o r d i scuss ion

L e t t e r from E P R I t o NRC, reques t ing meet ing between NRC P o l i c y Committee and U t i l i t y S t e e r i n g Committee f o r EPRI’s Standardized LWR Design Program.

L e t t e r f rom D. H. Moran, NRC, t o E P R I , summarizing February 14, 1983, meet ing w i t h E P R I i n Bethesda, Maryland, rega rd ing c u r r e n t s t a t u s and progress t o da te o f LWR s t a n d a r d i z a t i o n program.

L e t t e r f rom H. Denton, NRC, t o EPRI , responding t o February 21, 1983, reques t f o r meet ing between NRC P o l i c y Committee and U t i l i t y S tee r ing Committee f o r E P R I ’ s Standardized LWR Design Program. A p r i l 6, 1983, i n Bethesda, Maryland.

Meet ing t o be h e l d on

L e t t e r f rom D. H. Moran, NRC, t o E P R I , summarizing February 22, 1983, meet ing w i t h EPRI and Sol Levy Associates i n Palo A l t o , C a l i f o r n i a , r e g a r d i n g f i n a l p r i o r i t y sequence and r e l e g a t i o n o f nonapp l i cab le s a f e t y i ssues t o i n a c t i v e s t a t u s and c o n s o l i d a t i o n o f i n f o r m a t i o n on r e s o l u t i o n o f unresolved issues .

L e t t e r f rom S. Burs te in , NRC, t o E P R I , c o n f i r m i n g a c t i o n i tems r e s u l t i n g f rom meet ing w i t h U t i l i t y S t e e r i n g Committee f o r E P R I Standardized LWR P1 a n t Design Eva lua t i on Program on A p r i l 6, 1983, meet ing. Formal p l a n should be developed f o r coopera t i ve e f f o r t s between NRC and E P R I . Related i n f o r m a t i o n enclosed.

L e t t e r f rom D. H. Moran, NRC, t o E P R I , f o rward ing (1 ) d r a f t t a s k a c t i o n p lan , Rev is ion 2 f o r Task A-45 rega rd ing shutdown DHR requirements and ( 2 ) a paper on same s u b j e c t presented be fo re Swiss Federal I n s t i t u t e f o r Reactor Research i n l a t e A p r i l 1983.

L e t t e r f rom R. E. N i c k e l l , E P R I , t o NRC, acknowledging r e c e i p t of Rev is ion 2 o f d r a f t t a s k a c t i o n p l a n f o r Task A-45 rega rd ing shutdown DHR requirements and g i v i n g d e t a i l e d program p l a n discussed a t June 7-15, 1983, meetings. Related i n f o r m a t i o n enclosed.

L e t t e r f rom D. H. Moran, NRC, t o EPRI , summarizing August 11, 1983, meet ing w i t h U t i l i t y S t e e r i n g Committee f o r E P R I Standardized LWR P l a n t Design Eva lua t i on Program i n Bethesda, Maryland, rega rd ing progress o f s tandard ized des ign program.

L e t t e r f rom H. R. Denton, NRC, t o E P R I , f o rward ing rev iew o f E P R I l i s t o f gener i c i ssues n o t a p p l i c a b l e t o

A- 2

November 29, 1983

April 3, 1984

April 20, 1984

June 12, 1984

September 11, 1984

October 17, 1984

January 23, 1985

February 7, 1985

April 4, 1985

certification of future LWR standardized plants, as committed to at August 11, 1983, meeting. Results to be discussed at November 7, 1983, meeting.

Letter from D. H. Moran, NRC, to EPRI, summarizing November 7, 1983, meeting with EPRI Utility Steering Committee for LWR Standardized Plant Design Evaluation Program regarding corroboration of i ssues not appl i cab1 e to certification of future LWR plants. Agenda and viewgraphs enclosed.

Letter from D. H. Moran, NRC, to EPRI, summarizing February 16, 1984, meeting with EPRI in Bethesda, Maryland, regarding evaluation of LWR standardized plant design. List of attendees, agenda, and viewgraphs enclosed.

Letter from H. R. Denton, NRC, to ACRS, requesting that ACRS comment on correctness of categorization of generic safety and licensing issues, identifying which issues affect future LWR standardized plant designs.

Letter from D. H. Moran, NRC, to EPRI, summarizing NRC Policy Committee and EPRI Utility Steering Committee for LWR Standardization Plant Design Evaluation fifth quarterly meeting on May 23, 1984, in Bethesda, Maryland. List of attendees and viewgraphs enclosed.

Letter from J. C. Ebersole, ACRS, to NRC, discussing proposed categorization of generic safety and licensing issues, identifying which issues affect future LWR standardized plant designs. resolution would require considerable resource a1 1 ocat i on.

EPRI’s plan for issue

Letter from H. R. Denton, NRC, to ACRS, responding to ACRS September 11, 1984, letter regarding EPRI categorization of generic safety and licensing issues. Resource plan being drafted to consider resources required for appropriate review.

Letter from D. H. Moran, NRC, to EPRI, summarizing December 5, 1984, meeting with EPRI Utility Steering Committee for LWR Standardized Plant Design Evaluation. Commission approval of resource expenditures contingent on EPRI presentation showing rapid progress.

Transcript of Commission February 7, 1985, meeting in Washington, DC, regarding NRC briefing on standard design process. Pp 1-66. Supporting documentation encl osed.

Letter from S. Burstein, EPRI, to NRC, forwarding expanded wri teups for shutdown DHR, emergency core cooling system methodology, and high-strength bolting

Program Summary A-3

June 4 , 1985

October 15, 1985

October 15, 1985

December 2, 1985

December 3 , 1985

December 27 , 1985

January 24 , 1986

March 7 , 1986

March 24 , 1986

topics. April 11, 1985, NRC Policy Committee and EPRI Utility Steering Committee meeting requested.

Discussion of expanded format and agenda of

Letter from D. H. Moran, NRC, to EPRI, summarizing April 4 , 1985, meeting with EPRI Utility Steering Committee for LWR Standard Plant Design Evaluation in Bethesda, Maryland, regarding status of EPRI/industry ALWR program and of safety and 1 icensing issues.

Letter from D. H. Moran, NRC, to EPRI, summarizing August 27 and 28, 1985, working meetings in Bethesda, Maryland of EPRI and NRC project and task managers regarding listed generic safety issues affecting EPRI/industry ALWR Utility Requirements Document. remaining and forecast issues enclosed.

Agenda and summary of

Letter from D. H. Moran, NRC, to EPRI, summarizing September 12, 1985, quarterly meeting with EPRI Utility Steering Committee for LWR Standardized Plant Design Evaluation regarding status of ALWR program. viewgraphs, and list o f attendees enclosed.

Agenda,

Letter from D. H. Moran, NRC, to EPRI, summarizing October 28-30, 1985, working meetings with EPRI and NRC project managers regarding resolution of generic safety and licensing issues. Agenda, list of attendees, and summary of remaining and new issues enclosed.

Letter from C. F. Sears, EPRI, to NRC, forwarding draft "NRC Issue and Document Review Process for EPRIIIndustry Advanced LWR Program." Comments requested by January 15, 1986.

Letter from D. H. Moran, NRC, to EPRI, summarizing EPRI Utility Steering Committee for LWR Standardized Plant Design Evaluation and NRC Policy Committee meeting on December 3 , 1985, in Bethesda, Maryland, regarding status of ALWR program. List of attendees, agenda, and handouts enclosed.

Letter from T. P. Speis, NRC, to EPRI, forwarding comments on topical report submitted on December 3 , 1985, regarding ALWR Requirements Document, Chapter 1.

Letter from T. P. Speis, NRC, to EPRI, commenting on draft material regarding ALWR program and NRC review process. date and good base to develop NUREG on ALWR program.

Draft submittal good description of program to

Letter from C. F. Sears, EPRI, to NRC, forwarding "Advanced LWR P1 ant Optimization Subjects for Chapter 1 of Requirements Document." Optimization subjects

Program Summary A-4

addressed current regulations that might be overly conservative. Response requested by April 30, 1986.

June 5, 1986

June 6, 1986

June 10, 1986

June 30, 1986

July 8, 1986

July 8, 1986

Septmber 9, 1986

Octotier 15, 1986

Deceniber 9, 1986

Program S ummary

Letter from D. H. Moran, NRC, to EPRI, summarizing March 12, 1986, meeting with ALWR Utility Steering Committee regarding current status o f ALWR program. attendees, agenda, and related information enclosed.

List of

Letter from T. P. Speis, NRC, to EPRI, responding to request for review and opinion of "Advanced LWR Plant Optimization Subjects for Chapter 1 of Requirements Docu- ment." Optimization subjects in conformance with August 8, 1985 severe-accident policy.

Letter from D. H. Moran, NRC, to EPRI, forwarding trip report of attendance at May 5-7, 1986, working meetings with EPRI regarding draft Chapter 2 of Utility Requirements Document and NUREG-1197, "Advance Light Water Reactor Program - Program Management and Staff Review Methodology." Viewgraphs, agenda, and related information enclosed.

Letter from EPRI to NRC, forwarding executive summary of ALWR Utility Requirements Document.

Letter from C. F. Sears, EPRI, to NRC, discussing program to develop requirements for ALWRs to resolve generic safety and licensing issues applicable to future LWRs, through NRC, EPRI, and ALWR Utility Steering Committee efforts . Letter from C. F. Sears, EPRI, to NRC,_forwarding information on four safety and licensing issues from generic list of issues being evaluated for resolution via EPRI ALWR program, for comment.

Letter from D. H. Moran, NRC, to EPRI, summarizing ALWR Utility Steering Committee and NRC Policy Committee July 8, 1986, meeting in Bethesda, Maryland, regarding enclosed list of agenda items, including ALWR program support. Supporting documentation enclosed.

Letter from C. F. Sears, EPRI, to NRC, forwarding Chapter 2 of ALWR Utility Requirements Document." Chapter contained utility requirements for main and extraction steam, feedwater and condensate, chemical addition, condensate makeup purification, and auxiliary steam.

Letter from T . P. Speis, NRC, to EPRI, responding to July 8, 1986, request for comments on topic papers regarding Chapter 3 of ALWR Utility Requirements Document. mended that pump seal issue be resolved.

Recom-

A- 5

January 5, 1987 Letter from T. P. Speis, NRC, to EPRI, forwarding questions and comments resulting from review of EPRI ALWR Utility Requirements Document, Chapter 1. Response requested by February 28, 1987, to support NRC completion of draft safety evaluation report (DSER) on Chapter 1.

January 15, 1987 Letter from W. Kerr, ACRS, to NRC, summarizing ACRS 321st meeting on January 8-10, 1987, regarding improved safety requirements and objectives-for future LWRs.

February 2, 1987 Letter from T. P. Speis, NRC, to EPRI requesting that EPRI submit comments via ALWR Utility Steering Committee regarding ACRS recommendations on improved safety requirements and objectives for future LWR plants. Comments requested by February 27, 1987.

February 9, 1987 Letter from R. M. Bernero, NRC, to EPRI, forwarding ACRS January 15, 1987, letter presenting recommendations on improved safety design features for future LWRs. Com- ments in support of NRC and ACRS review of General Electric's Advanced Boiling Water Reactor (ABWR) requested by February 27, 1987.

February 26, 1987 Letter from H. R. Denton, NRC, to multiple recipients, forwarding executive summary of ALWR Utility Requirements Document. desired for plants of 1990s and beyond.

Document to provide specific attributes

March 2, 1987 Letter from L. J. Ybarrondo, NRC contractor, to NRC, forwarding "Independent Assessment of 'Advanced LWR Utility Requirements Document Chapter 1: Overall Requirements'".

March 5, 1987 Letter from D. H. Moran, NRC, to EPRI, summarizing October 15, 1986, quarterly meeting with ALWR Utility Steering Committee and NRC Policy Committee. List of attendees, agenda, and viewgraphs enclosed.

March 11, 1987 Letter from T. P. Speis, NRC, to EPRI, forwarding questions and comments from Sections 2 and 3 of Scientech Report SCIE-022-87 for January 1987. response to questions and comments requested by March 31, 1987. Report not enclosed.

EPRI

March 18, 1987 Letter from T. P. Speis, NRC, to EPRI, requesting additional information regarding review of ALWR Utility Requirements Document, Chapter 1. Response requested by April 15, 1987.

March 27, 1987 Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to independent review comments on ALWR Utility Requirements Document, Chapter 1, in response t o January 5, 1987, letter. Chapters 3 and 4 to be submitted in June 1987.

Program Summary A- 6

March 27, 1987

Apr 1 10, 1987

Apr 1 15, 1987

Letter from E. E. Kintner, EPRI, to NRC, forwarding comments on ACRS recommendations. for early dialogue with NRC and ACRS on important safety issues in ALWR program.

Letter from H. D. Curet, EPRI, to NRC, "Preliminary Comments on ALWR Utility Requirements Document, Chapter 2: Power Generation System."

Opportunity welcomed

Letter from D. A. Meneley, University of Fredericton, New Brunswick, to NRC, responding to H. Denton request for comments dated February 26, 1987, regarding executive summary of ALWR Utility Requirements Document. of regular measurement and reporting of reliability data for key safety systems at each plant recommended.

Program

April 22, 1987

April 23, 1987

April 29, 1987

May 5 , 1987

May 15, 1987

May 18, 1987

May 21, 1987

Letter from H. D. Curet, EPRI, to NRC, forwarding suggested modifications of Page 3 of Modification 2 to Contract NRC-03-86-057, "Independent Assessment of Advanced LWR Utility Requirements Documentation," and revised comments on Chapter 2 of ALWR Utility Requirements Document.

Letter from H. D. Curet, EPRI, to NRC, forwarding replacement pages 1-3 of "Preliminary Comments on Advanced LWR Utility Requirements Document, Chapter 2: Power Generation System."

Letter from M. M. Yedidia, EPRI, to NRC, forwarding abstracts for Chapters 2-13 of ALWR.Utility Requirements Document.

Letter from R. T. Dewling, NRC contractor, to NRC, responding to H. R. Denton request for views on ALWR Uti1 ity Requirements Document. n e e d e d t h a t ALWRs m a j o r s t e p f o r w a r d i n i m p r o v i n g p l a n t safety.

Convincing evidence

Letter from L. J. Ybarrondo, NRC contractor, to NRC, forwarding "Technical Evaluation of Chapter 2: Power Generation System o f Advanced LWR Utility Requirements Document . I' Letter from H. D. Curet, EPRI, to NRC, forwarding "Technical Evaluation of Chapter 2: Power Generation System of Advanced LWR Utility Requirements Document."

Letter from D. Crutchfield, NRC, to EPRI, requesting description o f quality assurance (QA) program applied during preparation of ALWR Utility Requirements Document and QA program to be applied t o ensure that future changes would be made in controlled fashion. NUREG-0800 and Regul atory Guide 1.28 enclosed.

Program Summary A- 7

May 27, 1987

May 31, 1987

June 1, 1987

June 5, 1987

June 5, 1987

June 9, 1987

June 12, 1987

June 15, 1987

June 18, 1987

June 22, 1987

June 30, 1987

Program Summary

Letter from P . H. Leech, NRC, to EPRI, requesting additional information on Chapter 2 of ALWR Utility Requirement Document. Scientech, Inc., provided in Enclosures 1 and 2 respectively. Response expected by September 1, 1987.

Comments and questions by NRC and

Letter from Duke Power Company, Westinghouse Electric Corporation, and CYGNA Energy Services, to NRC, forwarding EPRI NP-5159 "Guidelines for Specifying Inte- grated Computer-Aided Engineering Applications for Elec- tric Power Plants," . Letter to NRC forwarding comments on the executive summary for EPRI's ALWR Utility Requirements Document.

Letter from H. N. Berkow, NRC, to R.T. Dewling, acknowledging receipt of May 6, 1987, letter commenting on ALWR Utility Requirements Document executive summary. Comments forwarded to EPRI for consideration in preparing rest of Requirements Document.

Letter from H. N. Berkow, NRC, expressing appreciation for June 1, 1987, comments on ALWR Utility Requirements Document executive summary. Letter forwarded to EPRI for information.

Letter from H. D. Curet, EPRI, to NRC, forwarding revised page 4 of "Technical Evaluation of Chapter 2: Power Generation System of Advanced LWR Utility Requirements Document. 'I

Letter from P. H. Leech, NRC, to EPRI, requesting additional information on Chapter 2 of ALWR Utility Requirements Document, including description of live loading for valve packing developed by Atomic Energy of Canada, Limited (AECL) for CANDU plants, in accordance with May 27, 1987, request.

Letter from L. J. Ybarrondo, NRC contractor, to NRC, forwarding "Independent Assessment of Advanced LWR Utility Requirements Document," final monthly progress report for May 1987.

Letter from E. E. Kintner, EPRI, to NRC, forwarding Revisions 0 of Chapters 3 and 4 , of ALWR Utility Requirements Document for review.

Letter from P. Leech, NRC, to EPRI, summarizing Junq 17, 1987, meeting with EPRI representatives in Bethesda, Maryland, regarding briefing of new NRC reviewers o f ALWR Utility Requirements Document. viewgraphs enclosed.

List of attendees and

"Standard P1 ant Reviews Program P1 an''

A-8

June 30, 1987

June 30, 1987

July 9, 1987

July 20, 1987

July 22, 1987

September 17, 1987

Septeinber 24, 1987

September 24, 1987

Octoblv 6, 1987

October 12, 1987

November 13, 1987

Progrm Summary

Letter from E P R I , to NRC, forwarding R e v i s i o n 0 o f Chapter 4 of ALWR Utility Requirements Document.

Letter from EPRI, to NRC, forwarding Revision 0 of Chapter 3 of ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding six topic papers on generic issues pertaining to Chapters 1, 3, and 4, and optimization subject paper regarding Chapter 3 of ALWR requirements document.

Letter from P 1987, meeting Steering Comm attendees and

Letter from E May 21, 1987,

H. Leech, NRC, to EPRI summarizing July 1, with EPRI and Chairman o f ALWR Utility ttee regarding ALWR program. vi ewgraphs encl osed.

reauest to provide description o f QA

List of

E. Kintner, EPRI, to NRC, responding to

program appl i ed during preparation o f ALWR Uti 1 i ty Requirements Document. according to SRP Section 17.1 and Regulatory Guide 1.28.

QA program not called for

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 27 and June 12, 1987, questions and comments on Chapter 2 of ALWR Utility Requirements Docu- ment. Chapters 3 and 4 under NRC review. Chapter 5 should be available for review by end of November.

Letter from R. W. Hernan, NRC, to ACRS, forwarding September 24, 1987, DSER on Chapter 1 of ALWR Utility Requirements Document. Additional DSERs on subsequent chapters to be provided when complete. support ACRS October 6, 1987, meeting.

Information would

Letter from L. S. Rubenstein, NRC, to EPRI, forwarding DSER on Chapter 1 o f ALWR Utility Requirements Document. Findings generally favorable to ALWR program.

Transcript of ACRS Standardization of Nuclear Facilities Subcommittee meeting on October 6, 1987, in Washington, D.C. Pp 1-205. Supporting documentation enclosed.

Letter from J. C. Devine, EPRI, to NRC, commenting on proposed new SRP Section 3.6.3 regarding leak-before-break evaluation procedures. Proposed approach would severely 1 imi t appl i cat i on of leak-before-break technology for future plants and not achieve intent o f General Design Criterion 4 of Appendix A to 10 CFR Part 50.

Letter from P. H. Leech, NRC, to EPRI, requesting additional information on ALWR Utility Requirements

A- 9

Document, Chapters 3 and 4 . Schedule for review o f Chapters 3 and 4 based on receipt of response by December 31, 1987.

November 24, 1987 Letter from A. E. Scherer, Combustion Engineering, Inc., to NRC, forwarding proposed Advanced Reactor Severe Accident Program resolutions for four remaining NRCIIndustry Degraded Core Rulemaking (IDCOR) severe- accident issues. adopted in developing System 80+ design. Concurrence requested.

Six valid DOE/IDCOR resolutions to be

December 1, 1987 Letter from L. S. Rubenstein, NRC, to EPRI, forwarding list of significant technical issues that might arise during review of evolutionary standard ALWR plants. Comments requested.

December 8, 1987 Letter from E. E. Kintner, EPRI, to NRC, forwarding Revision 0 of ALWR Utility Requirements Document, Chap- ter 5, for review.

December 11, 1987 Letter from P. H . Leech, NRC, to EPRI, requesting additional information on Chapters 3 and 4 of ALWR Utility Requirements Document. Notification requested in case response delayed more than 2 to 3 weeks beyond December 31, 1987.

January 19, 1988 Letter from L. S . Rubenstein, NRC, to EPRI, forwarding status report on technical reviews of standardized plant designs for January 1988. Problem meeting listed input dates requested within 1 week of receipt of letter.

January 25, 1988 Letter from D. Crutchfield, NRC, to EPRI, forwarding information on ALWR performance goals. assembled after discussion with knowledgeable NRC staff members and during January 19, 1988, meeting with EPRI representatives.

Information

January 25, 1988 Letter from E. E. Kintner, EPRI, to NRC, forwarding additional information on Chapters 3 and 4 of ALWR Utility Requirements Document in response to all but two questions in November 13 and December 11, 1987 letters.

January 27, 1988 Letter from L. S. Rubenstein, NRC, to EPRI, requesting additional information regarding design goals addressing severe-accident releases. days of letter date.

Response requested within 30

February 5, 1988 Letter from Director, Office of Nuclear Reactor Regulation, to EPRI forwarding revised DSER on Chapter 1 of ALWR Uti 1 i ty Requirements Document.

February 18, 1988 Letter from L. S. Rubenstein, NRC, to EPRI, forwarding DSER on Chapter 2 ALWR Utility Requirements Document.

Program Summary A-10

February 18, 1988

Marctl 18, 1988

March, 25, 1988

March 28, 1988

April 4, 1988

April 6, 1988

May 13, 1988

May 31, 1988

June 3 , 1988

Utility requirements in Chapter 2 are in general agreement with NRC guidelines and regulatory requirements for power generation system involved.

Letter from L. S. Rubenstein, NRC, to Committee to Review Generic Requirements (CRGR), forwarding DSER on Chapter 2 of ALWR Utility Requirements Document and informing CRGR members and staff of progress in reviewing Requirements Document.

Letter from P. H. Leech, NRC, to EPRI, requesting additional information on Chapter 5 of ALWR Utility Requirements Document. review based on receipt of response by April 29, 1988. Notification expected if delay anticipated.

Letter from L. S. Rubenstein, NRC, to EPRI, forwarding safety evaluation of recommended modifications of Regulatory Guide 1.76, "Design Basis Tornado for Nucle r Power Plants." NRC interim position constitutes conservative reduction of design-basis winds for use by EPRI unti 1 revi sion avai 1 ab1 e.

Current schedule of

Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to Questions 36 and 40 to resolve NRC comments regarding ALWR low temperature overpressure protection requirements.

Letter from P. H. Leech, NRC, to EPRI, requesting additional information on Chapter 5 of ALWR Utility Requirements Document. by April 29, 1988.

Response to questions requested

Letter from E. E. Kintner, EPRI, to NRC, providing initial response to January 27, 1988, questions on ALWR Utility Requirements Document regarding implementation o f public safety criteria. (PRA) key assumptions and groundrules to be submitted in September 1988 to provide more detail.

Probabilistic risk assessment

Letter from L. S. Rubenstein, NRC, to EPRI, forwarding DSER on Chapter 3 of ALWR Utility Requirements Document. Document in general agreement with NRC guidelines based on January 25 and March 25, 1988, letters regarding reactor coolant system and nonsafety auxiliary systems.

Letter from Director, Office of Nuclear Reactor Regulation, forwarding DSER on Chapter 3 of ALWR Utility Requirements Document.

Letter from Director, Office of Nuclear Reactor Regulation, Director, forwarding DSER on Chapter 4 of ALWR Utility Requirements Document.


June 10, 1988

June 20, 1988

June 30, 1988

July 31, 1988

August 16, 1988

September 15, 1988

September 23, 1988

September 23, 1988

October 26, 1988

November 1, 1988

November 17, 1988

Letter from L. S. Rubenstein, NRC, to EPRI, forwarding DSER of Chapter 4 of ALWR Utility Requirements Document on the basis of commitments in January 25 and March 25, 1988, 1 etters.

Letter from L. S. Rubenstein, NRC, to ACRS, forwarding DSERs on Chapters 3 and 4 o f ALWR Utility Requirements Document. Effort to continue until final SER issued in 1991.

Report by Fauske and Associates, Inc., "Technical Support for Hydrogen Control Requirement for EPRI Advanced LWR Requirements Document Task 8.3.5.4," Advanced Reactor Severe Accident Program.

Letter from R. Stiger, International Technology Corp., to NRC, "Technical Basis for EPRI Advanced LWR Requirements Document Assumption on Del ayed Fi ssi on Product Re1 ease (Task 8.3.5.2)," Advanced Reactor Severe Accident Pro- gram.

L e t t e r from E. E. Kintner, EPRI, to NRC, forwarding responses to March 18 and April 4, 1988, requests for additional information on Chapter 5 of ALWR Utility Requirements Document on station blackout.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to NRC Comments 430.1, 430.2, 430.3, and 430.4, regarding ALWR Utility Requirements Document and two reports referenced in responses to NRC Comments 480.5 and 450.1.

Letter from W. 0. Long, NRC, to EPRI, regarding low- temperature overpressure protection for ALWRs.

Letter from C. Y . Cheng, NRC, to EPRI, clarifying position on reactor vessel survei 11 ance program and 1 ow temperature overpressure protection in Chapters 3 and 4 of ALWR Utility Requirements Document.

Letter from W. 0. Long, NRC, to EPRI, clarifying two statements in May 13, 1988, DSER on Chapter 3 of ALWR Utility Requirements Document concerning chemical and volume control system.

Letter from W. 0. Long, NRC, to EPRI, summarizing October 26, 1988, meeting with EPRI in Rockville, Maryland, regarding NRC recommendations for design of ALWR electrical system. Meeting handout enclosed.

Letter from W . 0. Long, NRC, to EPRI, advising that open issue regarding suppression pool loads in Section 3 . 0 . D . 2 of Chapter 1 of ALWR Requirements Document needed further


November 18, 1988

November 22, 1988

December 23, 1988

December 30, 1988

January 9, 1989

January 11, 1989

January 31, 1989

February 2, 1989

February 6, 1989

February 23, 1989

Program Summary

cl ari f i cati on. leak-before-break proposal consistent with General Design Criterion 4 of Appendix A to 10 CFR Part 50.

Section not cl ear on whether proposed

Letter from E. E. Kintner, EPRI, to NRC, forwarding Revision 0 o f ALWR Utility Requirements Document Chapter 6, for review.

Letter from D. Crutchfield, NRC, to EPRI, forwarding information regarding scope of safety analysis report for future standardized design applications and NRC review of applications.

Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Utility Requirements Document, Chapter 12.

Letter from E.E. Kintner, EPRI, to NRC, forwarding Revision 0 o f ALWR Utility Requirement Document, Chapter 8.

Letter from B. Lee, Nuclear Management and Resources Council (NUMARC), to NRC, clarifying NUMARC position on need for NRC rulemaking on severe reactor accidents for ALWRs. Rulemaking not necessary and may be coun- terproduct i ve.

Letter from E'. E. Kintner, EPRI, to NRC, forwarding ALWR Requirements Document, Chapter 9, for use in chapter review.

Letter from D. G. Harrison, Idaho National Engineering Laboratory, to NRC, forwarding "Interim External Events Integration for EPRI Advanced LWR Requirements Document WBS 4 . 3 . 3 . "

Letter from W . 0. Long, NRC, t o EPRI, discussing proposed resolutions of Generic Safety Issues I.F.1, "Expand QA List," and II.E.5 "Classification of Instrumentation and Electrical Equipment," and suggesting that EPRI reaffirm commitment in its December 3 , 1985, letter in order to resolve i ssues . Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Utility Requirements Document Chapter 13. Chapter incorporated lessons learned from evaluation of LWR techno1 ogy as appl i ed to generation of electricity . Letter from W . 0. Long, NRC, to EPRI, requesting additional information on ALWR Utility Requirements Document, Cha ter 6. level o f deta 1 consistent with Regulatory Guide 1.70 for safety analys s report.

Information should be provided on

A- 13

February 28, 1989

March 21, 1989

March 22, 1989

March 22, 1989

March 30, 1989

March 30, 1989

March 30, 1989

April 3, 1989

April 5, 1989

April 10, 1989

April 11, 1989

Program Summary

Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Utility Requirements Document, Chapter 7.

Letter from W . 0. Long, NRC, to EPRI, summarizing meeting on March 15, 1989, with EPRI and contractors in Palo Alto, California, regarding ALWR issues. List of attendees and handouts encl osed.

Letter from W . 0. Long, NRC, to EPRI, requesting additional information on ALWR Utility Requirements Document, Chapters 8, 9, 12, and 13. Response requested within 60 days.

Letter from W . 0. Long, NRC, to EPRI, requesting review of Chapter 7 of ALWR Utility Requirements Document. Schedule for review given.

Letter from E. E. Kintner, EPRI, to NRC, responding to open issue in DSER on Chapter 4 of ALWR Utility Requirements Document regarding capability of reactor pressure vessel to withstand multiple natural circulation cooldowns and to similar comment by ACRS on August 9, 1988.

Letter from E. E. Kintner, EPRI, to NRC, submitting additional information sufficient to resolve Generic Safety Issue (GSI) II.E.6.1, "In Situ Testing of Valves," ALWR Utility Requirements Document addressed subissues of GSI II.E.6.1. Concurrence on resolution of issue requested.

Letter from E. E. Kintner, EPRI, to NRC, responding to March 25, 1988, letter regarding optimization issue on tornado design in Chapter 1 of ALWR Utility Requirements Document. 2 and Figure 2 of safety evaluation for ALWR.

Agreed to use tornado design criteria in Table

Letter from E. E. Kintner, EPRI, to NRC, forwarding additional information on outstanding issues in DSERs on Chapters 2 , 3 , and 4 of ALWR Utility Requirements Docu- ment.

Letter from E. E. Kintner, EPRI, to NRC, regarding GSIs I.F.l and II.F.5.

Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Requirements Document, Chapter 11. EPRI believed major objectives had been achieved and concerns about station blackout would be virtually eliminated.

Letter from EPRI, to NRC, forwarding Revision 0 of ALWR Utility Requirements Document, Chapter 11.

A-14

April 28, 1989

May 16, 1989

May 17, 1989

May ;!4, 1989

June 8, 1989

June 16, 1989

July 3, 1989

July 3, 1989

July 14, 1989

July 19, 1989

August 18, 1989

Program Summary

Letter from W . 0. Long, NRC, to EPRI, requesting additional information on ALWR Utility Requirements Document, Chapters 6, 7, 8, and 9 in accordance with March 22, 1988, request for additional information.

Letter from W . 0. Long, NRC, to EPRI, discussing ALWR Utility Requirements Document regarding tornado design. Position on tornado missile barriers given in SRP Sections 3.5.1.4, 3.5.2, and 3.5.3 and in Regulatory Guides 1.76 and 1.117.

Letter from W . 0. Long, NRC, to EPRI, giving notice of June 22, 1989, meeting with ALWR Utility Steering Committee in Rockville, Maryland, to discuss ALWR issues including program overview, hydrogen generation and ignition, and design-basis-accident source term assumptions.

Letter from W . Long to EPRI requesting additional information on Chapter 7, 8, 11, and 12 of the ALWR Requirements Document.

Letter from W . 0. Long, NRC, to EPRI, requesting additional information on GLs-81-12 and 86-10.

Letter from W . 0. Long, NRC, to EPRI, regarding GSIs II.G.l and II.G.23 pertaining to pressurizer equipment electric power and threat of reactor coolant pump seal 1 oss of-cool ant-accident.

Letter from E. E. Kintner, EPRI, to NRC, forwarding Revision 0 of ALWR Utility Requirements Document, Appendix A to Chapter 1. be used in performing PRA.

Appendix to provide guidance to

Letter from E. E. Kintner, EPRI, to NRC, forwarding Chapter 6 o f ALWR Utility Requirements Document.

Letter from J. L. Blaha, NRC, to EPRI, forwarding DSER on Chapter 1 of ALWR Utility Requirements Document and actual and estimated review schedule for ALWR project. Estimated dates preliminary and subject to revision after EPRI and NRC discussions.

Letter from T. Kenyon, NRC, to EPRI, summarizing June 22, 1989, meeting with EPRI representatives of nuclear power industry and source term and related issues for ALWR program. List o f attendees and handouts enclosed.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to March 22 and April 28, 1989, letters regarding Chapters 6, 7, 8, 9, 12,' and 13 of ALWR Utility Requirements Document.

A-15

September 15, 1989

October 19, 1989

October 26, 1989

November 6, 1989

November 28, 1989

December 22, 1989

December 26, 1989

January 18, 1990

January 31, 1990

. February 5, 1990

February 22, 1990

Program Summary

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 24, 1989, request for additional information on Chapters 7, 8, 11, and 12 of ALWR Utility Requirements Document. Power suppl ies of emergency response facility system to be designed in accordance with NUREG-0696.

Letter from E. E. Kintner, EPRI, to NRC, forwarding additional information-on Chapters 6, 9, and 11 of ALWR Utility Requirements Document in response to June 8, 1989, request. ALWR to be in compliance with NUREG-0800 regarding enhanced fire protection.

Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Utility Requirements Document, Chapter 10, and Topic Paper, "Reactor Pressure Vessel Level Instrumentation for PWRs.

Letter from J. L. Blaha, NRC, to SECY, forwarding correspondence between EPRI and NRC on Chapter 5 o f ALWR Uti1 ity Requirements Document.

Letter from T. J . Kenyon, NRC, to E P R I , requesting information on reactor safeguards within 60 days o f date of letter.

Letter from E. E. Kintner, EPRI, to NRC, forwarding responses t o requests for additional information on Chapters 1, 6, 7, 8, 9, 11, and 13 of ALWR.

Letter from E. E. Kinter, EPRI, t o NRC, regarding review priorities and process.

Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to requests for additional information on Chapters 6, 7, 8, 9, 12, and 13 of ALWR Utility Requirements Document regarding physical security, insider and outsider sabotage threats, and controlled access to the containment.

Report by Fauske and Associates, Inc., "Technical Support for Hydrogen. Control Requirement for EPRI Advanced LWR Requirements Document."

Letter from E. E. Kintner, EPRI, to NRC, forwarding "Technical Support for Hydrogen Control Requirement for EPRI Advanced LWR Requirements Document."

Letter from E. E. Kintner, EPRI, to NRC, forwarding updated Revision 0 of ALWR Utility Requirements Document, Appendix A to Chapter 1. Section 3.2.2, "Earthquake," for NRC review.

Document reissued to submi t

A-16

Febriiary 27, 1990 Letter from K. M. Carr, NRC, to EPRI, respond December 26, 1989, letter expressing concerns progress of NRC review of ALWR Utility Requir Document.

Febrc-ary 28, 1990

March 16, 1990

March 29, 1990

March 31, 1990

April 10, 1990

June 20, 1990

June 30, 1990

July 3, 1990

July 3 , 199C

July 13, 1990

Progr'im Summary

ng to regarding ments

Letter from C. L. Miller, NRC, to EPRI, forwarding DSER Chapter 5 on ALWR Utility Requirements Document, including severe-accident prevention and mitigation, hydrogen generation, control source term issues, and station blackout.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on Chapters 8 and 13 of ALWR Utility Requirements Document regarding remote positive indication of correct alignment of manual isolation valve and cast pressure-retaining parts of stop valves, respectively.

Letter from E. E . Kintner, EPRI, to NRC, regarding treatment of generic safety issues for ALWR requirements.

Letter from EPRI to NRC, forwarding Volume I, "ALWR Policy and Summary of Top-Tier Requirements," of ALWR Utility Requirements Document.

Letter from T. J. Kenyon, NRC, to EPRI, requesting additional information on Volume I1 of ALWR Utility Require- ments Document on the basis of staff's review of Appendix A to Chapters 10 and 11. days of letter date.

Response requested within 60

Letter from T. J. Kenyon, NRC, to EPRI, summarizing December 6, 1989, meeting with EPRI regarding Chapter 10 of Volume I1 of ALWR Utility Requirements Document. of attendees and viewgraphs enclosed.

List

Report by Fauske and Associates, Inc. , "Technical Support for Debris Coolability Requirements for Advanced LWRs in Utility-EPRI LWR Requirements Document."

Letter from S. H. Smith, Nuclear Power Oversight Commit- tee, to NRC, advising NRC that U.S. utility industry vitally interested in timely certification of both evolutionary and passive ALWR designs currently under review by NRC. schedules important.

Maintaining current certification review

Letter from E. E. Kintner, EPRI, to NRC, forwarding "Technical Support for Debris Coolability Requirements for Advanced LWRs in Utility-EPRI LWR Requirements Document. I'

Letter from T. J. Kenyon, NRC, to EPRI, requesting information needed to complete review of design criteria.

A-17

July 23, 1990

July 23, 1990

August 2, 1990

August 2, 1990

August 8, 1990

August 13, 1990

August 15, 1990

August 22, 1990

August 23, 1990

August 27, 1990

Program Summary

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to April 10, 1990, request for additional information on Chapters 1, 10, and 11 of ALWR Utility Requirements Document.

Letter from EPRI, to NRC, forwarding "Prevention of Early Containment Failure Due to High Pressure Melt Ejection and Direct Containment Heating for Advanced LWRs."

Letter from T. J. Kenyon, NRC, to EPRI, requesting additional information on Chapter 10 of Volume I1 o f ALWR Utility Requirements Document regarding electrical system.

Letter from T. J. Kenyon, NRC, to EPRI, summarizing May 31, 1990, meeting with EPRI, NUMARC and standardized plant vendors regarding Volume I1 of ALWR Utility Requirements Document. List of attendees enclosed.

Letter from T. J. Kenyon, NRC, to EPRI, summarizing May 31, 1990, meeting with EPRI regarding source term t o be used for future LWR. List o f attendees and viewgraphs encl osed.

Letter from J. M. Taylor, NRC, to S . H. Smith, Nuclear Power Oversight Committee, responding to July 3, 1990, letter to Chairman K. Carr regarding timely certification of two evolutionary passive designs. Sufficient time needed by NRC staff to identify and evaluate issues to ensure that proposed ALWR designs provide adequate protection.

Letter from T. J. Kenyon, NRC, to EPRI, forwarding comments regarding unresolved and generic safety issues addressed in Appendix B to Chapter 10 of Volume I1 of ALWR Utility Requirements Document.

Letter from T . J. Kenyon, NRC, to EPRI, requesting additional information on Volume I1 of ALWR Utility Re- quirements Document regarding design criteria. requested within 60 days of letter.

Response

Letter from T. J. Kenyon, NRC, to EPRI, forwarding summary of July 16, 1990, meeting with EPRI regarding hydrogen generation and containment performance. attendees, NRC slides, and EPRI presentation enclosed.

List of

Letter from T. J. Kenyon, NRC, to EPRI, forwarding revised request for additional information on Chapter 1 of Volume I1 of ALWR Utility Requirements Document.

A- 18

August 30, 1990

August 31, 1990

September 7, 1990

September 30, 1990

October 12, 1990

October 18, 1990

October 29, 1990

November 7, 1990

November 29, 1990

December 6, 1990

December 21, 1990

Letter from T. 3 . Kenyon, NRC, to EPRI, requesting additional information on ALWR Utility Requirements Document, Volume I, and Chapters 1, 6, and 10 of Volume I 1 to complete review of design criteria.

Letter from EPRI, to NRC, submitting annotated Revision 1 of Volume I 1 to ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding proprietary Revision 1 of Volume I 1 (ALWR evolutionary plant) and Revision 0 of Volume I 1 1 (ALWR passive plant) of ALWR Utility Requirements Document for safety evaluation review. March 29, 1991, letter from EPRI.

Documents no longer withheld per

Report by D. E. Leaver and L. P. Tenera, "Licensing Design Basis Source Term Update for Evolutionary Advanced LWR. I'

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to July 13 and August 2, 1990, requests for additional information on Chapters 7, 10, and 13 of ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding "Licensing Design Basis Source Term Update for Evol uti onary Advanced LWR. I'

Letter from W. R. Sugnet, EPRI, to NRC, forwarding tables listing NRC rules and regulatory guidance not in Appendix B to Chapter 1 of Volumes I 1 and I 1 1 of Advanced LWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding additional information on Appendix A to Chapter 1 and Chapter 12 o f ALWR Utility Requirement Document.

Transcript of closed-session meeting with ALWR Utility Steering Committee on November 29, 1990, with NRC. Pp 64-223. Supporting information enclosed. Transcript no longer proprietary per March 29, 1991 letter from EPRI.

Letter from E. E. Kintner, EPRI, to NRC, forwarding proprietary response to August 30, 1990, request for additional information on human factors considerations. Response addressed questions on ALWR Utility Requirements Document, Volume I, and Chapters 1, 6, and 10 of Volume 11. Enclosure no longer withheld per March 29, 1991, 1 etter from EPRI . Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on unresolved and generic safety issues and addressing


December 21, 1990

January 9, 1991

January 15, 1991

January 25, 1991

January 25, 1991

January 28, 1991

January 30, 1991

January 31, 1991

February 1, 1991

February 4, 1991

Program Summary

questions on Volume I1 of ALWR Utility Requirements Document. 1991, letter form EPRI.

Response no longer withheld per March 29,

Letter from T. H. Boyce, NRC, to EPRI, forwarding notification of January 14, 1991, meeting with EPRI in Rockville, Maryland, to discuss seismic issues for ALWR Util ity Requirements Document.

Letter from W. R. Sugnet, EPRI, to NRC, forwarding proprietary information on ALWR seismic design evaluation program. March 29, 1991, letter from EPRI. Letter from C. L. Miller, NRC, to EPRI, forwarding DSER on Chapters 6, 7 , 8, 9 , 12, and 13 of Volume I1 of ALWR Util ity Requirements Document.

Letter from T. J. Kenyon, NRC, to EPRI, requesting additional information on Volumes I1 and I11 of ALWR Util ity Requirements Document regarding quality assurance. Document no l o n g e r w i t h h e l d p e r March 29, 1991 letter from EPRI.

Enclosures no longer withheld per

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on reactor vessel level instrumentation system and addressing concerns related to Volume I1 of ALWR Utility Requirements Document. March 29, 1991 letter from EPRI.

Response no longer withheld per

Letter from W. R. Sugnet, EPRI, to NRC, forwarding comparison between Volumes 11, and I11 of ALWR Utility Requirements Document. March 29, 1991 letter from EPRI.

Enclosures on longer withheld per

Letter from W. R. Sugnet, EPRI, to NRC, forwarding data base for tracking open issues in DSER for Chapters 1-5 o f Volume I1 of ALWR Utility Requirements Document. sure no longer withheld per March 29, 1991 letter from EPRI.

Enclo-

Report by Jack R. Benjamin and Associates, Inc., "Ad- vanced LWR Seismic Design and Evaluation Program." Report no longer withheld per March 29, 1991 letter from EPRI.

Letter from W. R. Sugnet, EPRI, to NRC, forwarding Revision 1 of "Development of Seismic Hazard Input for Advanced LWR Seismic PRA," in response to NRC request at meeting on January 14, 1991.

Letter from D. Crutchfield, NRC, to EPRI, providing preliminary views on Revision 0 of Volume I11 of ALWR

A-20

Febriary 7, 1991

Febriary 7, 1991

Febriary 22, 1991

Febriary 28, 1991

Febriary 28, 1991

Febriary 28, 1991

Marct 1, 1991

March 1, 1991

March 1, 1991

Utility Requirements Document regarding criteria to be used in design for combustible gas control. longer withheld per march 29, 1991 letter from EPRI.

Document no

Letter from T. H. Boyce, NRC, to EPRI, providing summary of January 14, 1991 meeting with EPRI in Rockville, Maryland, regarding seismic issues for ALWR Utility Requirements Document. March 29, 1991 letter from EPRI.

Summary no longer withheld per

Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to questions raised at meeting on November 29 and 30, 1990. Numbering scheme for responses also enclosed. 1991 letter from EPRI.

Responses no longer withheld per March 29,

Letter from W. R. Sugnet, EPRI, to NRC, forwarding EPRI NP-5159, "Guidelines for Specifying Integrated Computer- Aided Engineered Applications for Electric Power Plants," and EPRI NP-5639, "Guide1 ines for Piping System Reconciliation (NCIG-05, Revision l)."

Letter from W. R. Sugnet, EPRI, to NRC, forwarding Sections 1-5 of Chapter 1, Volume I 1 of ALWR Utility Requirements Document to propose level of information to be witheld from public disclosure. withheld per March 29, 1991 letter from EPRI.

Document no longer

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I 1 1 of ALWR Utility Requirements Document.

Report by M. W . McCann, Jack R. Benjamin and Associates, Inc., Revision 1 of "Development of Seismic Hazard Input for Advanced LWR Se-i-smi c PRA.

Letter from J. H. Wilson, NRC, to E P R I , requesting additional information Volume I 1 1 of ALWR Utility Requirements Document regarding quality assurance. Document no longer withheld per March 29, 1991 letter from EPRI.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I 1 1 of ALWR Utility Re- quirements Document regarding physical security and safeguards requirements. Document no longer proprietary per March 29, 1991 letter from EPRI.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I 1 1 of ALWR Utility Requirements Document regarding plant systems. no longer withheld per March 29, 1991 letter from EPRI.

Document

Progr,am Summary A-21

March 1, 1991

March 1, 1991

March 1, 1991

March 1, 1991

March 8 , 1991

March 14, 1991

March 15, 1991

March 15, 1991

March 19, 1991

March 28, 1991

Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on ALWR Utility Requirements Docu- ment regarding radiation protection and health physics. Document no longer withheld per March 29, 1991 letter from EPRI.

Letter from J . H. Wilson; NRC, to EPRI, providing summary of February 11 and 12, 1991, meetings with EPRI in Rock- ville, Maryland, to discuss NRC staff review of Volume I1 of ALWR Utility Requirements Document. List of attendees encl osed.

Letter from J . H. Wilson, NRC, to EPRI, correcting summary of February 11 and 12, 1991, meetings in Rockville, Maryland, t o discuss NRC staff review of Volume I 1 of ALWR Utility Requirements Document.

Letter from W . R. Sugnet, EPRI, to NRC, forwarding industry list of technical issues central to design o f ALWR passive pl ants.

Letter from 3 . H. Wilson, NRC, to E P R I , requesting additional information on Volume I11 of ALWR Utility Re- quirements Document regarding reactor systems. Document no longer withheld per March 29, 1991 letter from EPRI.

Letter from E. E. Kintner, EPRI, to NRC, providing perspective of progress on work on ALWR Utility Require- ments Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to January 25, 1991, request for additional information on quality assurance. withheld per March 29, 1991 letter from EPRI.

Response no longer

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to requests for additional information on quality assurance. March 29, 1991 letter from EPRI.

Enclosure no longer withheld per

Letter from T. H. Boyce, NRC, to EPRI, providing summary of March 5 and 6, 1991, meetings with Combustion Engineering and EPRI in Rockville, Maryland, regarding NRC staff review of Volume I1 of ALWR Utility Re- quirements Document. List of attendees and agenda enclosed.

Letter from T. H. Boyce, NRC, to EPRI, providing summary of March 28, 1991, meeting with EPRI in Rockville, Maryland, to discuss NRC staff review of Volume I1 of ALWR Uti 1 i ty Requirements Document.


Apri. 3, 1991

Apri' 3, 1991

Aprii 5, 1991

April 8, 1991

April 17, 1991

April 17, 1991

April 22, 1991

April 24, 1991

April 24, 1991

April 25, 1991

April 25, 1991

Program Summary

Letter from C. L. Miller, NRC, to EPRI, forwarding DSER on Chapter 11 of Volume I1 of ALWR Utility Requirements Document.

Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document. review of design criteria addressed during review of Chapter 9, "Site Support System. I'

Information needed to complete

Letter from J . H. Wilson, NRC, to EPRI, concluding that regulations and guidance not applicable to design requirements should be added to "applicable" list in Appendix B to Chapter 1 of ALWR Utility Requirements Document. Information given in October 29, 1990, letter should be added to next revision of document.

Letter from J. H. Wilson, NRC, to EPRI, requesting EPRI's position on changes to criteria in Volume I11 of ALWR Uti1 ity Requirements Document that apply to evolutionary designs.

Letter from J. H. Wilson, NRC, to EPRI , requesting information on Volume I11 of ALWR Utility Requirements Document regarding human factors.

Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding safeguards.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding reactor systems.

Letter from K. M. Carr, NRC, to EPRI, responding to March 14, 1991, letter noting significant progress made in last year on ALWR Utility Requirements Document. Agreed that review o f advanced passive reactors presented different kind of challenge.

Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding structural engineering concerns.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 o f ALWR Utility Requirements Document regarding materials and chemcial engineering concerns.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to requests for additional information on

A-23

Chapters 4 and 7 of Volume I11 of ALWR Utility Require- ments Document regarding reactivity physics, fuel, and fuel storage criticality.

April 26, 1991 Letter from E. E . Kintner, EPRI, to NRC, forwarding Revision 2 of Volume I1 and Revision 1 of Volume I11 of ALWR Utility Requirements Document in response to NRC concerns regarding qual i ty assurances, human factors, generic and unresolved safety issues, and open issues in DSERs . Changes resul ti ng from el imi nat i on of operat i ng- basis earthquake a1 so i ncl uded.

1 30, 1991 Letter from EPRI, to NRC, submitting copies of Revision 1 of Volume I1 of ALWR Utility Requirements Document; and Chapter 1 and Appendix B to Chapter 1 updated through Revision 2.

1 30, 1991 Letter from EPRI, to NRC, transmitting revised pages issued as Revision 2 of Volume I1 and Revision 1 of Volume I11 of ALWR Utility Requirements Document.

April 30, 1991 Letter from EPRI, to NRC, transmitting "Position Paper on Standardi zat i on. "

May 6, 1991 Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding materials engineering concerns.

May 6, 1991

May 8, 1991

May 8, 1991

May 8, 1991

May 9, 1991

Letter from D. Crutchfield, NRC, to EPRI, clarifying issues discussed in SECY-90-016.

Letter from E . E. Kintner, EPRI, to NRC, response to April 8, 1991, letter regard changes to Volumes I1 and I11 of ALWR Ut Requirements Document.

Letter from E. E. Kintner. EPRI, to NRC,

forwarding ng consistent 1 ity

forwarding di ti onal response to March 8, 1991; request for a

information on Chapter 4 of Volume I11 of ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to April 5, 1991, letter regarding regulations and guidance not applicable to ALWR design requirements.

Letter from E. E. Kintner, EPRI, to NRC, forwarding report describing plant containment performance criteria for evolutionary plant and rationale for their selection. Information provided technical support for Volume I1 of ALWR Utility Requirements Document.


May S, 1991

May 13, 1991

May 13, 1991

May 13, 1991

May 1 3 , 1991

May 13, 1991

May 17, 1991

May 1 7 , 1991

May l 'r, 1991

May l ? , 1991

Progriim Summary

Letter from E. E . Kintner, EPRI, to NRC, forwarding matrix approach to containment performance criteria for evolutionary plants and rationale for their selection.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to March 1, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding measures to be taken to prevent high operational and post-shutdown radiation levels in reactor coolant system piping.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to February 28, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding system reliability, quality assurance, and seismic qualification or capability of non-safety- grade active systems.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to March 1 , 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding resolution of GSI A-29 for advanced reactors.

Letter from E. E. Kintner, EPRI, to NRC, forwarding partial response to March 1, 1991, request for additional information on Volume I1 of ALWR Utility Requirements regarding fire protection.

Letter from EPRI, to NRC forwarding response to March 1 , 1991, request for additional information on Volume I11 o f ALWR Utility Requirements Document.

Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Requirements Document regarding mechanical engineering concerns.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding instrumentation and control s .

Letter from J . H. Wilson, ,NRC, to EPRI, forwarding page 15 omitted from April 2 2 , 1991, request for additional information regarding review of Chapter 5 of Volume I11 of ALWR Utility Requirements Document.

Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding reactor systems.

A-25

May 17, 1991

May 22, 1991

May 22, 1991

May 22, 1991

June 5, 1991

June 13, 1991

June 24, 1991

July 1, 1991

July 1, 1991

July 2, 1991

Letter from J . H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Require- ments Document submitted on September 7, 1990 regarding safeguards considerations.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on Volume I11 of ALWR Utility Requirements Document regarding plant systems.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on Volume I11 of ALWR Requirements Document regarding radi at i on protection.

Letter from G. Bockhold, EPRI, to NRC, forwarding updated open issues tracking system for Chapters 1-9, 11, and 12 of Volume I1 of ALWR Utility Requirements Document.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on reactor systems to complete review o f Volume I 1 o f ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to April 17, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding human factors.

Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to April 3 and 17, 1991, requests for additional information on ALWR Utility Requirements Document regarding safeguards concerns.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to requests for additional information on Volume I11 of ALWR Utility Requirements Document regarding compliance of leak-before-break methodology with acceptance criteria in statement of considerations for final rule, General Design Criterion 4 o f Appendix A t o 10 CFR Part 50.

Letter from E. E. Kintner, EPRI, to NRC, forward response to April 22, 1991, request for additional information on Section 6.2.2 of Chapter 1 of Volume I11 o f ALWR Requirements Document regarding reliability requirements for non-safety critical systems.

Letter from E . E. Kintner, EPRI, to NRC, forwarding response to April 2 4 , 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding safety-margin basis requirements.


July 8, 1991

July 22, 1991

July 22, 1991

July 22, 1991

Augujt 1, 1991

Augu;t 1, 1991

August 1, 1991

August 12, 1991

Augu;t 16, 1991

August 19, 1991

August 19, 1991

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume 111 of ALWR Utility Requirements Docu- ment regarding emergency pl anni ng.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume 111 o f ALWR Utility Requirements Document regarding methodology for dedicating commercial software for designing safety-re1 ated systems.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding site security interfaces in plant security systems and key-locked controls.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume 111 of ALWR Utility Requirements Document regarding safes, single-action valves, security barriers, and intrusion detection systems.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to April 22, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding reactor systems.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding mechanical engineering regarding materi a1 and chemical engineering concerns.

Letter from E. E. Kintner, E P R I , t o NRC, forwarding final response to May 17, 1991, request for additional information Volume I11 of ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to June 5, 1991, request for additional information on Volume I1 o f ALWR Utility Requirements Document regarding reactor systems.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding human factors.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding safeguards.


August 20, 1991

August 29, 1991

September 5, 1991

September 5, 1991

September 6, 1991

September 11, 1991

September 11, 1991

September 23, 1991

October 1, 1991

October 2, 1991

October 2, 1991

October 3, 1991

Letter from T. J. Kenyon, NRC, to EPRI, providing summary of July 17, 1991, meeting to discuss effects of changes in source term on ALWR designs.

Letter from D. Crutchfield, NRC, to EPRI, requesting additional information on Volume I11 o f ALWR Utility Requirements Document regarding Question 210.40.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I1 of ALWR Utility Require- ments Document regarding shutdown risks.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding shutdown risks.

Letter from J. H. Wilson, NRC, to EPRI, transmitting open issues from staff review of Appendix A to Chapter 1 of Volume I1 of ALWR Utility Requirements Document.

Letter from D. E . Leaver, Tenera, L.P. (formerly Tenera Corporation), to NRC, providing additional information for NRC consideration on several ALWR source term matters that came up at August 1991 meeting.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document regarding unresolved and generic safety issues.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Appendix A to Chapter 1 of Volume I11 of ALWR Utility Requirements Document.

Letter from G. Bockhold, EPRI, to NRC, forwarding Volumes I1 and I11 of ALWR Utility Requirements Document consisting of hand-marked page changes in accordance with NRC request.

Letter from J. H. Wilson, NRC, to EPRI, summarizing August 27, 1991, meeting with utilities in Rockville, Mary1 and, regarding development of updated source term for LWRs.

Letter from J. H. Wilson, NRC, to EPRI, summarizing August 14 and 15, 1991, meetings with EPRI in Rockville, Maryland, to discuss issues associated with staff review o f Volume I11 ALWR Utility Requirements Document.

Letter from R. Chambers, Idaho National Engineering Labo- ratory, to NRC, forwarding two copies of Advanced Reactor Severe Accident Program report "Interim External Events Integration for EPRI Advanced LWR Requirements Document


October 8, 1991

Octoher 9, 1991

Octotier 10, 1991

Octotier 10, 1991

Octotier 17, 1991

Octotler 23, 1991

October 30, 1991

November 4, 1991

November 4, 1991

November 6, 1991

Program Summary

WBS 4.3.3," DOE-10-10227, by 0. G. Harrison in response to request by J. D. Trotter of EPRI to support review of Project 669.

Letter from R. C. Pierson, NRC, to EPRI, forwarding DSER on Chapter 10 of Volume I1 of ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to August 19, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document. Battery rooms to be locked and alarmed, and isolation zone lighting to be designed to permit observation.

Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume I11 of ALWR Utility Requirements Document submitted on September 7, 1990, regarding the reliability assurance program.

Letter from J. H. Wilson, NRC, to EPRI, forwarding open issues from review of Section 6 of Chapter 1 of Volume I1 o f ALWR Utility Requirements Document.

Letter from G. Bockhold, EPRI, to EPRI, forwarding response to questions raised during teleconference regarding ALWR seismic hazard curve.

Letter from J. H. Wilson, NRC, to EPRI, forwarding cor- rections for pages 4-10 and 4-11 of DSER on Chapter 10 of Volume I1 of ALWR Utility Requirements Document transmitted by letter dated October 8, 1991.

Letter from J . H. Wilson, NRC, to EPRI, summarizing September 26, 1991, meeting with EPRI in Rockville, Maryland, to present results o f research by NRC and contractors concerning development of updated source term. List of attendees provided.

Letter from R. C. Pierson, NRC, to EPRI, forwarding DSER on Appendix A to Chapter 1 of Volume I1 of ALWR Utility Requirements Document. Open issues required resolution.

Letter from G. Bockhold, EPRI, to NRC, forwarding information regarding errors and resolutions in response to request for additional information on ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response t o August 29, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding mechanical engineering concerns.

A-29

November 15, 1991

November 20, 1991

November 25, 1991

November 27, 1991

December 2 , 1991

December 6, 1991

December 10, 1991

December 16, 1991

December 18, 1991

December 20, 1991

Program Summary

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to August 19, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding human factors.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to September 11, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding unresolved and generic safety issues.

Letter from E. E. Kintner, EPRI, to NRC, forwarding Revision 3 of Volume I1 of ALWR Utility Requirements Document . Letter from E. E. Kintner and J. J. Taylor, EPRI, to NRC, proposing agenda of meetings at senior levels of NRC staff and Utility Steering Committee to discuss remaining generic safety issues in ALWR Utility Requirements Docu- ment as discussed at workshop on November 4 and 5, 1991.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to September 5 , 1991, request for additional information on Volume I1 of ALWR Utility Requirements Document regarding shutdown risk.

Letter from FPRI , to NRC, forwarding ALWR positions on central i ssue? pertaining to evol uti onary pl ant identified during July 1991 meeting with NRC. included containment performance, core debris coolability, and seismic hazard.

Issues I

Letter from J. D. Trotter, EPRI, to NRC,forwarding pen- and-ink changes to Chapters 5, 6, 9, 10, and 11 of Volume I1 of ALWR Utility Requirements Document regarding security and suggesting working-1 eve1 meeting during week of January 6, 1992, to resolve any outstanding concerns.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to September 5, 1991, request for additional information on Chapters 4 and 5 of Volume I1 of ALWR Utility Requirements Document regarding scope of PRA for operating conditions when plant is at power and general initiating events.

Letter from G. Bockhold, EPRI, to NRC, advising that modification of Figure 12.3-1 in Chapter 12 of Volume I1 of ALWR Utility Requirements Document was inadvertently omitted from list of changes submitted to NRC in June and October 1991.

Letter from D. Crutchfield, NRC, to EPRI, identifying issues pertaining to evolutionary and passive plant designs. approaches to resolving issues.

Information provided to initiate discussion of

A-30

December 20, 1991

December 21, 1991

December 26, 1991

January 9, 1992

January 9, 1992

January 10, 1992

January 10, 1992

January 22, 1992

January 24, 1992

January 24, 1992

Program Summary

I

Letter from J. D. Trotter, EPRI, to NRC, summary of technical rationale by C. Neg high-eff i ciency particulate air f i 1 ters waste off-gas system.

Letter from J. D. Trotter, EPRI, to NRC, DroDosed minor revisions of ALWR Utility

forwarding n regarding n radioactive

forwarding Requirements

Dochment addressing open issues pertaining to reactor pressure vessel materi a1 s .

Letter from D. Crutchfield, NRC, to EPRI, discussing preparation of final SER on Volume I1 of ALWR Utility Requirements Documents and requesting that EPRI submit all responses and positions by January 31, 1992, so that s t a f f could complete i t s t echn ica l eva lua t ions .

Letter from E. E. Kintner, EPRI, to'NRC, forwarding response to October 10, 1991, request for additional information on open issues resulting from NRC review of Section 6 of Chapter 1 of Volume I1 of ALWR Utility Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to October 10, 1991, request for additional information on Volume I11 of ALWR Requirements Document.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to August 16, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding electrical systems.

Letter from G. Bockhold, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume I11 of ALWR Utility Requirements Document regarding core support structures because initial response dated August 1, 1991, was inappropriate.

Letter from D. Crutchfield, NRC, to EPRI, discussing January 30, 1992, meeting with EPRI in Palo Alto, California, concerning resolution of policy and technical issues associated with ALWR Utility Requirements Document . Letter from J. H. Wilson, NRC, t o EPRI, summarizing January 8, 1992, meeting with EPRI in Rockville, Maryland, to discuss EPRI-proposed changes to security requirements.

Letter from E. E. Kintner, EPRI, to NRC, forwarding response to September 23, 1991, request additional information on Appendix A to Chapter 1 of Volume I11 of ALWR Utility Requirements Document.

A-3 1

January 24, 1992 L e t t e r f rom G, Bockhold, EPRI , t o NRC, fo rward ing responses t o open and con f i rma to ry i ssues i n DSER on Appendix A t o Chapter 1 o f Volume I 1 o f ALWR U t i l i t y Requirements Document. summaries f o r each i ssue and markups o f pages i n Volume I 1 pages.

Enclosure i nc luded bo th da ta base

January 24, 1992 L e t t e r f rom E. E. K in tne r , EPRI , t o NRC, submi t t i ng changes t o Chapters 5, 6, 9, 10, and 11 o f Volume I 1 o f ALWR U t i l i t y Requirements Document t h a t d e a l t w i t h safeguards-re la ted requi rements. Changes should be considered i n f i n a l SER on Volume 11.

January 24, 1992 L e t t e r f rom G. Bockhold, EPRI , t o NRC, fo rward ing responses t o t h r e e issues i n DSER on Chapter 11 o f Volume I 1 o f ALWR U t i l i t y Requirements Document. Enclosure i nc ludes bo th da ta base summaries f o r each i ssue and markups o f pages i n Volume 11.

January 28, 1992

February 3, 1992

L e t t e r f rom G. Bockhold, EPRI , t o NRC, f o rward ing f i r s t p a r t o f responses t o open and con f i rma to ry i ssues (99 o f 126 issues) i n DSER on Chapter 10 o f Volume I 1 o f ALWR U t i l i t y Requirements Document. be r e f l e c t e d i n Rev is ion 4 o f Volume 11.

Minor proposed changes t o

L e t t e r f rom G. Bockhold, E P R I , t o NRC, f o rward ing second p a r t o f responses t o open and c o n f i r m a t o r y i ssues i n DSER on Chapter 10 o f Volume I 1 o f ALWR U t i l i t y Requirements Document and s t a t i n g t h a t responses had been g i ven f o r a l l b u t f o u r o f t h e issues. Responses t o remain ing f o u r i ssues t o be prov ided l a t e r t h i s month.

February 3, 1992 L e t t e r f rom G. Bockhold, E P R I , t o NRC, fo rward ing changes t o Volume I 1 o f ALWR U t i l i t y Requirements Document t h a t addressed DSER open and con f i rma to ry i ssues o r o t h e r concerns. Volume 11.

Changes t o be i nco rpo ra ted i n t o Rev is ion 4 o f

February 4 , 1992 L e t t e r f rom G. Bockhold, E P R I , t o NRC, fo rward ing proposal f o r maintenance feed t o t o p - t i e r e l e c t r i c a l bus a c t i o n .

February 10, 1992 L e t t e r f rom T. U. Marston, EPRI , t o NRC, suggest ing s t a f f - l e v e l meetings be h e l d i n l a t e February and e a r l y March o f 1992 t o exped i te q u a l i t y c l o s u r e o f DSER issues p e r t a i n i n g t o chapters on man-machine i n t e r f a c e systems and PRA i n Volume I 1 o f U t i l i t y Requirements Document.

February 11, 1992 L e t t e r f rom J . D. T r o t t e r , EPRI, t o NRC, d i scuss ing m i sce l 1 aneous i tems f o r P r o j e c t 669 i n c l u d i ng r e s o l u t i on o f steam genera tor tube r u p t u r e s by i n c r e a s i n g des ign pressure.


February 18, 1992

February 27, 1992

March 2, 1992

March 3, 1992

March 3, 1992

March 9, 1992

March 10, 1992

March 19, 1992

Marcti 19, 1992

March 30, 1992

March 31, 1992

Program Summary

Letter from G. Bockhold, EPRI, to NRC, forwarding responses to four open issues in DSER on Chapter 10 of Volume I 1 of ALWR Utility Requirements Document in response to October 10, 1991, request for additional information and containing both data base and road-map summaries for each issue and markups of pages in Volume 11.

Letter from D. Crutchfield, NRC, to EPRI, forwarding

policy issues on evolutionary and passive plant designs. Positions supersede those in February 20, 1991 letter.

e draft Commission paper describing major technical and

Letter from T. J. Kenyon, NRC, to EPRI, summarizing January 30, 1992 senior management meeting with EPRI and nuclear industry representatives on technical issues for evolutionary and passive ALWRs. presentations enclosed.

List of attendees and

Letter from G. Bockhold, EPRI, to NRC, forwarding February 1992 status report. apparent in area of policy issues.

Significant progress

Letter from G. Bockhold, EPRI, to NRC, forwarding proposed changes to Chapter 1, Appendix B to Chapter 1, and Chapters 10 and 1 1 of Volume I 1 of ALWR Utility Requ i remen t s Document . Letter from T. J. Kenyon, NRC, to EPRI, issuing errata to summary of senior management meeting on January 30, 1992 on technical issues for evolutionary and passive ALWRs.

Letter from J. D. Trotter, EPRI, to NRC, forwarding final draft of position paper for passive plant system classification and requirements.

Letter from E . E . Kintner, EPRI, to NRC, forwarding position paper on passive plant system classification and requirements,

Letter from G. Bockhold, EPRI, to NRC, forwarding changes to Volume I 1 of ALWR Utility Requirements Document per continuing discussions with NRC.

Letter from G. Bockhold, EPRI, to NRC, submitting changes to Chapter 11 of ALWR Utility Requirements Document in response to NRC concerns associated with electrical distribution policy issue.

Letter from G. Bockhold, EPRI, to NRC, forwarding changes to Volume I 1 1 o f ALWR Utility Requirements Document addressing open issues on performance requirement for turbine exhaust boot and design of radial and thrust bearings.

A-33

A p r i l 3, 1992

A p r i l 7, 1992

A p r i l 9, 1992

A p r i l 9, 1992

A p r i l 9, 1992

A p r i l 17, 1992

A p r i l 17, 1992

A p r i l 24, 1992

A p r i l 30, 1992

May 1, 1992

May 5, 1992

Program Summary

L e t t e r f rom G. Bockhold, EPRI , t o NRC, f o rward ing comments on d r a f t NUREG/CR-5747 "Est imate o f Rad ionuc l ide Release C h a r a c t e r i s t i c s i n t o Containment Under Severe Acc ident Cond i t ions . "

L e t t e r f rom G. Bockhold, E P R I , t o NRC, s u b m i t t i n g response t o NRC reques t f o r a d d i t i o n a l i n f o r m a t i o n on second source o f power t o non-safety loads r e q u i r e d e x c l u s i v e l y f o r u n i t ope ra t i on as disocussed i n Sec t ions 3.3 and 4.2 o f Chapter 11 o f DSER.

L e t t e r f rom G. Bockhold, EPRI , t o NRC, on d i v e r s i t y against common-mode software failures. Listed elements prov ided i n Chapter 10 o f ALWR U t i l i t y Requirements Document.

L e t t e r f rom W . Borchardt , NRC, t o E P R I , summarizing March 27, 1992 meeting w i t h E P R I i n Denver, Colorado on major i ssues r e s u l t i n g f rom NRC rev iew o f ALWR U t i l i t y Requirements Document f o r e v o l u t i o n a r y and pass ive designs. L i s t o f attendees, meeting agenda handouts, and s l i d e s enclosed.

L e t t e r f rom G. Bockhold, E P R I , t o NRC, s u b m i t t i n g summary o f methods & assumptions used i n development o f ALWR 80 th p e r c e n t i l e meteoro log ica l database.

L e t t e r f rom G. Bockhold, EPRI , t o NRC, s u b m i t t i n g changes t o Chapter 3 o f Volume I 1 o f ALWR U t i l i t y Requirements Document concerning t e s t a b i l i t y o f t h i r d feedwater i s o l a t i o n va l ve i n bwrs. S i m i l a r change w i l l be made t o Volume 111.

L e t t e r f rom E. E. K in tne r , E P R I , t o NRC, f o rward ing Rev is ion 4 t o Volume I 1 o f ALWR U t i l i t y Requirements Document.

L e t t e r f rom D. C r u t c h f i e l d , NRC, t o E P R I , f o rward ing DSER on Volume I11 o f ALWR U t i l i t y Requirements Document.

L e t t e r f rom D. C r u t c h f i e l d , NRC, t o E P R I , d i scuss ing t o p i c s under c o n s i d e r a t i o n f o r i n c l u s i o n i n Commission paper t o d iscuss a d d i t i o n a l i ssues on f u t u r e r e a c t o r designs.

L e t t e r f rom G. Bockhold, E P R I , t o NRC, f o rward ing response t o open i ssue on t h e use o f phys ica l l y -based source te rm on Volume I 1 o f t h e ALWR U t i l i t y Requirements Document . L e t t e r f rom G. Bockhold, EPRI , t o NRC, f o rward ing m o d i f i c a t i o n s t o ALWR U t i 1 i t y Requirements Document t o address general concern o f NRC on v u l n e r a b i l i t y o f ALWRs d u r i n g shutdown & low power ope ra t i on .

A-34

May 5 , 1992

May 8, 1992

May 13, 1992

May 15, 1992

May 28, 1992

May 26, 1992

June 4, 1992

June 15, 1992

June 23, 1992

J u l y 2, 1992

J u l y 6, 1992

J u l y 6, 1992

L e t t e r f rom G. Bockhold, EPRI , t o NRC, f o rward ing d r a f t o f PWR and BWR pass ive p l a n t system c l a s s i f i c a t i o n . Enclosure submi t ted t o f a c i l i t a t e p r e p a r a t i o n s f o r May 14, 1992 meet ing on r e g u l a t o r y t rea tmen t o f nonsafe ty systems.

L e t t e r f rom G. Bockhold, E P R I , t o NRC, f o rward ing EPRI's d r a f t p o s i t i o n s on a d d i t i o n a l t e c h n i c a l and p o l i c y i ssues on pass ive and e v o l u t i o n a r y p l a n t designs i n p a r a l l e l w i t h U t i l i t y S tee r ing Committee rev iew.

L e t t e r f rom E. E. K i n t n e r , E P R I , t o NRC, f o rward ing Rev is ion 3 t o Chapters 1-13 o f Volume I11 o f t h e ALWR Utility Requirements Document. L e t t e r f rom D. C r u t c h f i e l d , NRC, t o E P R I , f o rward ing d r a f t o f SER on Volume I 1 o f ALWR U t i l i t y Requirements Document.

L e t t e r f rom T. J. Kenyon, NRC, t o E P R I , summarizing meet ing w i t h E P R I on May 6, 1992 on in-containment f i s s i o n produc t removal mechanisms f o r e v o l u t i o n a r y and pass ive p l a n t designs.

L e t t e r f rom T. 3. Kenyon, NRC, t o E P R I , summarizing meet ing w i t h E P R I on March 20, 1992 on EPRI's i n i t i a l work on r e l i a b i l i t y - b a s e d t e c h n i c a l s p e c i f i c a t i o n s f o r pass ive ALWRs. i n c l uded.

L i s t o f at tendees and E P R I p r e s e n t a t i o n

L e t t e r f rom E. E. K i n t n e r , E P R I , t o NRC, responding t o February 27, 1992 l e t t e r on major t e c h n i c a l and p o l i c y i ssues f o r e v o l u t i o n a r y and pass ive p l a n t des igns

L e t t e r f rom G. Bockhold, E P R I , t o NRC, f o rward ing A p r i l and May 1992 s t a t u s r e p o r t

L e t t e r f rom D. C r u t c h f i e l d , NRC, t o EPRI , d i s c u s s i n g EPRI-proposed o p t i m i z a t i o n s u b j e c t on s i m p l i f i c a t i o n o f o f f - s i t e emergency p lann ing f o r ALWRs u s i n g pass i ve s a f e t y systems.

L e t t e r f o r G. Bockhold, E P R I , t o NRC, f o rward ing da ta base f o r i ssues i d e n t i f i e d on Chapters 2-10 o f Volume I 1 1 o f t h e ALWR U t i l i t y Requirements Document.

L e t t e r f rom G. Bockhold, EPRI , t o NRC, f o rward ing d i scuss ion o f t e rm 'decommissioning' as te rm r e l a t e s t o ALWR U t i 1 i t y Requirements Document.

L e t t e r f rom D. C r u t c h f i e l d , NRC, t o E P R I , f o rward ing d r a f t commission paper, "Design C e r t i f i c a t i o n and L i cens ing P o l i c y Issues P e r t a i n i n g t o Passive and E v o l u t i o n a r y Advanced L i g h t Water Reactor des igns . "


July 17, 1992 Letter f r om G. Bockhold, EPRI, to NRC, forwarding revised data base for issues on Chapter 9 of Volume I11 o f the ALWR Utility Requirements Document. balance-of-plant fire protection program, independence o f ventilation system inside containment, and requirements for smoke removal capability.

Issues include

July 7, 1992 Letter from T.G. Hiltz, NRC, to EPRI, summarizing meeting with EPRI on June 11, 1992 on generic system classification, reliability-based technical specifications, and shutdown risk considerations for ALWRs.

August 3, 1992 Letter from G . Bockhold, EPRI, to NRC, forwarding supporting information on charcoal filters per teleconference with NRC.


APPENDIX B

I REFERENCES

Atomit: I n d u s t r i a l Forum (A IF ) , AIF/NESP-020, "Compendium o f Design Features t o Reduc 2 Occupat ional Radi a t i on Exposure a t Nucl ear Power P1 an ts . 'I

--- , , \ I F Study, 6.4-12.

Committee on t h e B i o l o g i c a l E f f e c t s o f I o n i z i n g Rad ia t i on , "The E f f e c t on Popul i t i o n o f Exposure t o Low Level o f I o n i z i n g Rad ia t i on , " J u l y 1980

E l e c t p i c Power Research I n s t i t u t e ( E P R I ) , E P R I NP-309, "Human Fac to rs Review o f Nu1:lear Power P l a n t Con t ro l Room Design."

--- , I P R I NP-1081, "Re fue l i ng Outage Water C l a r i t y Improvement Study."

--- , I IPRI NP-1982, "Eva lua t i on o f Proposed Con t ro l Room Improvements Through A n a l y j i s o f C r i t i c a l Operator Ac t i ons . "

--- , I IPRI NP-2294, "Guide t o Design o f Secondary Systems and T h e i r Components To Min imize Oxygen-Induced Cor ros ion . "

--- , E P R I NP-2360, "Human Fac to rs Methods f o r Assessing and Enhancing Power P1 a n t Mai n t a i nabi 1 i t y . I' --- , E P R I NP-2411, "Human Eng ineer ing Guide f o r Enhancing Nuc lear Con t ro l Room. 'I

--- , IrPRI NP-2777.

--- , I i P R I NP-3448, "A Procedure f o r Reviewing and Improv ing Power P l a n t Alarm S y s t e m . 'I

--- , I iPRI NP-3659, "Human Fac to rs Guide f o r Nuc lear Power P l a n t C o n t r o l Room Devel opment. It

--- , IIPRI NP-3701, "Computer-Generated D i s p l a y Guide1 i n e s " (Volumes 1 and 2 ) .

--- , IlPRI NP-3784, "A Survey o f t h e L i t e r a t u r e on Low-Alloy S t e e l Fastener Cor ros ion i n PWR Power P1 ants , 'I J. S. H a l l , December 1984.

--- , IIPRI NP-4350, "Human Eng ineer ing Design Gu ide l i nes f o r M a i n t a i n a b i l i t y . "

--- , I:PRI NP-4762-SR.

--- , I:PRI NP-4947-SR, "BWR Hydrogen Water Chemistry Gu ide l i nes , " 1987.

Progrirm Summary B- 1

--- , E P R I NP-5067, "Good B o l t i n g P rac t i ces , A Reference Manual f o r Nuclear Power P1 an t Maintenance Personnel , It Volume 1 : Volume 2: "Small Bo1 t s and Threaded Fasteners, 'I 1990.

"Large Bo1 t Manual , I' 1987 and

--- , E P R I NP-5159, "Guidel i n e s f o r S p e c i f y i n g I n t e g r a t e d Computer-Aided Engineered A p p l i c a t i o n s f o r E l e c t r i c Power P lan ts . "

--- , E P R I NP-5283-SR-A, "Gu ide l i nes f o r Permanent BWR Hydrogen Water Chemistry I n s t a l 1 a t i o n s , " September 1987.

--- , E P R I NP-5479, " A p p l i c a t i o n Gu ide l i nes f o r Check Valves i n Nuc lear Power P1 ants . 'I

--- , E P R I NP-5639, "Guidel i n e s f o r P i p i n g System R e c o n c i l i a t i o n (NCIG-05, Rev is ion l ) . "

--- , E P R I NP-5652, "Guidance f o r t h e U t i l i z a t i o n o f Commercial Grade I t e m s i n Nucl ear S a f e t y Re1 ated Appl i c a t i ons (NCIG-07). It

--- , E P R I NP-5693.

--- , E P R I NP-5769, "Degradat ion and F a i l u r e of B o l t i n g i n Nuc lear Power P l a n t s , " R. E. N i c k e l l , P r i n c i p a l I n v e s t i g a t o r , Volumes 1 and 2, A p r i l 1988.

--- , E P R I NP-5960, "PWR Pr imary Water Chemistry Guidel ines , " Rev is ion 1.

--- , E P R I NP-5989, " E f f e c t s o f Control-Room L i g h t i n g on Operator Performance, A P i l o t Emp i r i ca l Study."

--- , E P R I NP-6202, " M a t e r i a l S p e c i f i c a t i o n f o r A l l o y X-750 i n LWR I n t e r n a l Components . I' --- , E P R I NP-6209, " E f f e c t i v e P l a n t Labe l i ng and Coding."

--- , E P R I NP-6316, "Gu ide l i nes f o r Threaded-Fastener A p p l i c a t i o n i n Nuc lear Power P lan ts , " Looram Engineer ing, Inc. , J u l y 1989.

--- , E P R I NP-6433.

--- , E P R I NP-6559.

--- , EPRI NP-6628.

--- , E P R I NP-6748.

--- , E P R I NP-7077, Rev is ion 2.

--- E P R I NP-7183-SLY "SHARP 1, A Revised Systemat ic Human A c t i o n R e l i a b i l i t y Procedure. It

--- , EPRI RP-2184-7.

--- , E P R I RP-2705-7.

Program Summary B- 2

--- , EPRI TR-100259, "An Approach to the Analysis of Operator Actions in Probabilistic Risk Assessment," G. W . Parry, et al., Draft, November 1991.

Fauske and Associates, Inc., "Process for Evaluating Accident Management Capabi 1 it i es . --- , "Technical Support for the Hydrogen Control Requirement for the EPRI Advaticed Light Water Reactor Requirements Document," June 1988.

Federal Guidel ines on Dam Safety.

General Electric Company, NEDO-22155, "Generation and Mi tigation o f Combustible Mixtures in Inerted Mark I Containments."

--- , NEDO-31643 P (proprietary).

--- , NED0 31858.

I1 1 unii nat i on Engineering Society, "IES Lighting Handbook. I'

National Research Council, "Estimated Probabilities of Extreme Winds," 1988.

Nuclear Construction Issues Group (NC1G)-07, "Guidance for the Utilization of Commclrci a1 Grade I tems in Nucl ear Safety Re1 ated Appl i cat i ons. I'

--- , NCIG-14 (EPRI NP-6628), "Procedure for Seismic Evaluation and Design of Small Bore Piping," April 1990.

Nuclc,ar Safety Analysis Center (NSAC), NSAC-39, "Verification and Validation for $,afety Parameter Di spl ay Systems. I'

--- , NSAC-147, "Losses of Off-Site Power at U.S. Nuclear Power Plants; Through 1989. I'

Nuclear Utilities Management and Resource Council (NUMARC), NUMARC-87-00.

--- , NUMARC Containment Integrity Working Group Report, February 1988.

--- , "Process for Evaluating Accident Management Capabilities."

INDUSTRY CODES AND STANDARDS

American Concrete Institute (ACI) , 318, "Building Code Requirements for Reinforced Concrete."

American National Standards Institute (ANSI) , 10.4, "Guidel ines for the Verification and Validation of Scientific and Engineering Computer Programs for the Nuclear Industry," 1987.

, 35.1. ---

--- , A13.1-1981 (Reaffirmed 1985) , "Scheme for the Identification of Piping Systems. I'

Program Summary 8-3

--- 9 A58.1-1982, "Minimum Design Loadings for Buildings and Other Structures. I'

--- , C96.1.

--- , MCll.1-1976 (ISA-57.3), "Quality Standard for Instrument Air Systems."

--- , N45.2.1, "Cleaning of Fluid Systems and Associated Components During Construction Phase of Nucl ear Power P1 ants. I'

--- , N45.2.2, "Packaging, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power P1 ants. I'

--- , N45.2.3, "Housekeeping During the Construction Phase of Nuclear Power Plants."

--- , N45.4-1972, "Leakage Rate Testing of Conta Reactors. I'

--- , N101.2-1980, "Protective Coatings (Points) Reactor Containment Faci 1 it i es, 'I ,1980. --- , N101.4-1972, "Quality Assurance for Protec Nuclear Facil i ties. 'I

nment Structures for Nuclear

for Light Water Nuclear

ive Coatings Applied to

--- , 235.1-1972, "Accident Prevention Signs, Specification for"

--- , 286.1.

--- , "Leak Rate Testing of the Containment Structures for Nuclear Reactors."

American National Standards Institute/American Concrete Institute (ANSI/ACI), 349, "Code Requirements for Nuclear Safety-Re1 ated Structures. I'

American National Standards Institute/American Institute of Steel Construction (ANSI/AISC), N-690, "Specification for the Design, Fabrication. and Erection of Steel Safety-Related Structures for Nuclear Facilit

American National Standards Institute/American Nuclear 2.3, "Standard for Estimating Tornado and Extreme Wind Nucl ear Power Sites, 'I 1980.

--- , 2.5-1984, "Standard for Determining Meteorologica Power Sites. It

es," Chjcago, Illin

Society (ANSI/ANS), Characteristics at

Information at Nuc

--- , 2.8-1981, "Standard for Determining Design Basis Flooding at Power Reactor Sites. I'

i s .

ear

--- , 2.12, "American Nuclear Society Guidelines for Combining Natural and Man- Made Hazards at Power Reactor Sites."

--- , 3.1, "Selection, Qual i f i cat i on and Training for Nucl ear Power P1 ants. I'

--- , 3.3-1988.

Program Summary B-4

--- , 18.1, "American Na t iona l Standard Rad ia t i on Source Term f o r Normal Oper ' i t i on o f L i g h t Water Reactors. I'

--- , 51.1, "Nuclear Sa fe ty C r i t e r i a f o r t h e Design o f S t a t i o n a r y PWR P l a n t s . "

--- , 52.1, "Nuclear Safety C r i t e r i a f o r t h e Design o f S t a t i o n a r y BWR P l a n t s . "

--- , 55.1, " S o l i d Rad ioac t i ve Waste Processing Systems f o r L i g h t Water Cooled Reactor P1 an ts . I' _ _ _ , 55.4, "Gaseous Rad ioac t i ve Waste Processing Systems f o r L i g h t Water Reactor P1 an ts . I'

--- , 55.6, " L i q u i d Rad ioac t i ve Waste Processing Systems f o r L i g h t Water Reactor P1 an ts . I'

_ _ _ , 56.2, "Containment I s o l a t i o n P rov i s ions f o r F l u i d Systems A f t e r a LOCA," 1976 and 1984.

, 56.7. ---

--- , 56.8, "Containment System Leakage T e s t i n g Requirements," 1987

--- , 57.1, "Design Requirements f o r L i g h t Water Reactor Fuel Hand l ing S y s t m s . I'

--- , 57.2, "Design Requirements f o r L i g h t Water Reactor Spent Fuel Storage F a c i l i t i e s a t Nuc lear Power P lan ts . "

--- , 57.3, "Design Requirements f o r New Fuel Storage F a c i l i t i e s a t L i g h t Water React o r P1 an ts . I'

--- , 58.1, " P l a n t Design Aga ins t M i s s i l e s . "

--- , 58.2, "Design Bas is f o r P r o t e c t i o n o f L i g h t Water Nuc lear Power P l a n t s Aga ins t t h e E f f e c t s o f Pipe Rupture."

--- , 58.8-1984, "Time Response Design C r i t e r i a f o r S a f e t y Re la ted Opera tor Ac t i ons a t Nuc lear Power P lan ts . "

--- , 58.9, " S i n g l e F a i l u r e C r i t e r i a f o r L i g h t Water Reactor Safe ty -Re la ted F l u i d Systems."

--- , 59.2-1985, "Sa fe ty C r i t e r i a f o r HVAC Systems Located Outs ide Pr imary Containment. I'

American N a t i o n a l Standards I n s t i t u t e / A m e r i c a n Soc ie ty o f Mechanical Engineers

--- , 46-1, "Code on Nuclear A i r and Gas Treatment," 1989.

--- , B31.1, "Power P ip ing . "

(ANSI/ASME), 3.3-1988.

Program Summary B- 5

--- , B.31.1, Appendix 2, "Non-Mandatory Rules for the Design of Safety Valve Instal 1 ations. 'I

--- , N509, "Nuclear Power Plant Air Cleaning Units and Components," 1989.

--- , N510, "Testing of Nuclear Air-Cleaning System," 1989.

--- , NOG-1-1983, "Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) . I '

--- , NQA-1.

--- , NQA-2. --- , OM-6, "Inservice Testing of Pumps."

--- , OM-10, "Inservice Testing of Valves."

American National Standards Institute/Institute of Electrical and Electronics Engineers (ANSI/IEEE) , 387, "IEEE Standard Criteria for Diesel Generator Units Appl i ed as Standby Power Suppl i es f o r Nucl e a r Power Generat i ng Stations , 'I

--- , 730 , "Software Qual i ty Assurance P1 ans. 'I

, 828. ---

--- , 829, "Software Test Documentation."

--- , 982.1-1988, "IEEE Standard Dictionary of Me Software. I'

ures to Produ e R 1 iabl

--- , 982.2-1988, "IEEE Guide for the Use of IEEE Standard Dictionary of Measures To Produce Re1 i ab1 e Software. I'

--- , 1012-1986, "IEEE Standard for Software V&V Plan."

--- , 1042.

--- , 1063-1987, "Standard for Software Users Documentation." --- , ANS-7-4.3.2-1982, "Application Criteria for Programmable Digital Computer Systems in Safety Systems of Nuclear Power Generating Stations."

American Nuclear Society (ANS), 5.1, "Decay Heat Power in Light Water Reactors," La Grange Park, Illinois, October 1975 and October 1979.

--- , 18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."

American Society for Testing and Materials (ASTM), A262 Practice E, Modified Strauss Test.

--- , A 708 Strauss Test.

Program Summary B-6

, I 800. ---

--- , I 3803, "Standard Test Methods for Radiological Testing of Nuclear-Grade Gas-Pnase Absorbents."

--- , I 3842.80.

--- , E-185-82, "Standard Recommended Practices for Surveillance Tests for Nuclear Reactor Vessel s , I' 1982.

, E-813. ---

Ameri:an Society of Civil Engineers ( A S C E ) , 4-1986, "Seismic Analysis of SafetpRelated Nuclear Structures and Commentary on Standard fo r Seismic Analysis of Safety-Related Nuclear Structures," September 1986.

--- , 7-1988 (formerly ANSI A58.1).

Ameri:an Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code) , Section I I I , "Nucl ear Power P1 a n t Components. 'I

--- , Section 111, Appendix 11, Paragraph 11-1430.

--- , Section 111, Appendix N.

--- , Section 111, Division 1, "Nuclear Power Plant Components, with Appendices . I' --- , Section 111, Division 2 , "Code for Concrete Reactor Vessels and Containments."

--- , Section 111, Subsection CC-3720.

--- , Section 111, Subsection NB/NC/ND-llOO(a).

--- , :Section 111, Subsection NCA-1140.

--- , Section 111, Subsection NF, "Component Supports."

--- , Section 111, Subsection N G , "Core .Support Structures."

--- , Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components. I'

--- , Section XI, Subsection IWV 3421-3427(a).

--- , Code Case N-397.

--- , Code Case N-411, "Alternative Damping Values for Seismic Analysis of Classes 1, 2 , and 3 Piping Sections."

--- , Code Case N-420, "Linear Energy Absorbing Supports fo r Subsection NF, Classes 1, 2 , and 3 Construction, Section 111, Division 1."

I Progrtim Summary B- 7



--- , NCA-1140.

--- , NQA-2AY P a r t 2.7, "Qual i t y Assurance Requirements o f Computer Software f o r Nucl ear Fac i 1 i t y Appl i c a t i ons. I'

--- , NQA 2.7.

--- , PTC 6, "Steam Turbines Performance Tes t Code."

--- , PTC 6.1, " A l t e r n a t i v e Procedure f o r T e s t i n g Steam Turbir)e."

--- , TDP-2.

I n s t i t u t e o f E l e c t r i c a l and E l e c t r o n i c s Engineers (IEEE), 279, " C r i t e r i a f o r P r o t e c t i o n Systems f o r Nuclear Power Generat ing S ta t i ons . "

--- , 308-1980, " C r i t e r i a f o r Class 1E Power Systems f o r Nuclear Power Generat ing S t a t i o n s . "

--- , 323-1974, " I E E E Standard f o r Q u a l i f y i n g Class 1E Equipment f o r Nuclear Power Generat ing S t a t i o n s . "

--- , 334, "Standard f o r Type Tes ts o f Continuous Duty Class 1E Motors f o r Nucl ear Power Generat i ng S t a t i o n s . I'

--- , 338-1977, " C r i t e r i a f o r t h e P e r i o d i c T e s t i n g o f Nuc lear Power Generat ing S t a t i o n Sa fe ty Systems."

--- , 344-1987, "Recommended P r a c t i c e s f o r Seismic Q u a l i f i c a t i o n o f Class 1E Equipment f o r ' Nuc lear Power Generat ing S ta t i ons . "

--- , 352, "Guide f o r General P r i n c i p l e s o f R e l i a b i l i t y A n a l y s i s o f Nuclear Power Generat ing S t a t i o n P r o t e c t i o n Systems."

--- , 383, "Type Tes t o f Class 1E E l e c t r i c Cables, F i e l d Sp l ices , and Connections f o r Nuc lear Power Generating S t a t i o n s . "

--- , 384, " I E E E T r ia l -Use Standard C r i t e r i a f o r Separa t ion o f Class 1E Equipment and C i r c u i t s , " 1974.

--- , 472-1974.

, 519.

--- , 577, "Requirements f o r R e l i a b i l i t y Ana lys i s i n t h e Design and Opera t ion of Safety Systems f o r Nuclear Power Generat ing S t a t ons. I'

---

Program Summary B-8

--- , 603, "Trial-Use Standard Criteria for Safety Systems for Nuclear Power Generating Stations."

--- , 741, "IEEE Standard Criteria for the Protection of Class 1 E Power Systems and Kquipment in Nuclear Power Generating Stations."

--- , 765-1983, "IEEE Standard for Preferred Power Supply for Nuclear Power Generating Stations."

--- , 828-1983, "IEEE Standard for Software Configuration Management Plans."

--- , 981.1.

--- , 981.2.

--- , 982.2-1988, "IEEE Guide for the Use of IEEE Standard Dictionary o f Measures To Produce Re1 i ab1 e Software. I'

--- , 1008-1987, "IEEE Standard for Software Unit Testing."

--- , 1042.

--- , 1050-1989, "IEEE Guide for Instrumentation and Control Grounding in Generating Stations."

--- , C37-90.1-1989.

--- , C62.41-1980, "IEEE Guide for Surge Voltages in Low-Voltage AC Power Ci rcui ts. 'I

--- , P1023/05, "Guide for the Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Power Generating Stations."

Instrument Society of America (ISA), 67.15, Draft RP 67.04, Part 11, "Mett,odology for the Determination of Setpoints for Nuclear Safety-Re1 ated Instr8umentation."

Insulated Power Cable Engineers Association (IPCEA), 5-61-402.

International Electrotechnical Commission (IEC), 880-1986, "Software for Compiters in the Safety Systems of Nuclear Power Stations."

Nat i c nal El ectri cal Code.

Naticnal Fire Protection Association (NFPA), 13, "Standard for the Instal- laticn of Sprinkler Systems."

, 70. ---

--- , 90A-1989, "Installation of Air Conditioning and Ventilating Systems."

--- , 101, "Life Safety Code."

Progr,am Summary B- 9

--- , 803, " F i r e P r o t e c t i o n f o r L i g h t Water Nuc lear Power P lan ts . "

Tubu lar Heat Exchangers Manufacturers A s s o c i a t i o n (THEMA), T-2.41.

Un i fo rm B u i l d i n g Code.

U.S. Department o f Defense (DOD), DOD-MIL-HDBK-217Ey " R e l i a b i l i t y P r e d i c t i o n o f E l e c t r o n i c Equipment."

--- , DOD-MIL-HDBK-263.

--- , DOD-MIL-HDBK-338, " E l e c t r o n i c Re1 i a b i l i t y Design Handbook."

--- , DOD-MI L-HDBK-472 , "Mai n t a i nabi 1 i ty P r e d i c t i o n . I'

--- , DOD-STD-781 , "Re1 i a b i 1 i t y Tes t Methods , P1 ans , and Environment o f Engi n e e r i ng Development , Qual i f i c a t i on , and Produc t ion . I'

--- , DOD-STD-1399.

--- , DOD-STD-1629AY "Procedures f o r Per fo rming a F a i l u r e Modes E f f e c t s and C r i t i c a l i t y Analys is . I'

Program Summary B-10

APPENDIX C

LIST OF ABBREVIATIONS

The - fo l l ow ing i s a l i s t o f abb rev ia t i ons used throughout t h i s r e p o r t and t h e DSER

ac AAC ABWR AC I AC RS A /D ADS AEOD AFW AHU A I F A I S C ALARi\ ALWR AN S A N S I A00 ARSAI' ASCE ASD ASME ASTM ATW S AWS

BE B E I R BOP BT P B&W BWG BWR BWROCi

c- I c- I1 CAE CAGS CAS CCDF CCFP C C I C CCTV

a l t e r n a t i n g c u r r e n t a l t e r n a t e ac advanced b o i l i n g water r e a c t o r American Concrete I n s t i t u t e Adv isory Committee on Reactor Safeguards analog t o d i g i t a l automat ic depressur i z a t i on system O f f i c e f o r A n a l y s i s and E v a l u a t i o n o f Opera t iona l Data aux i 1 i a r y feedwater a i r hand l i ng u n i t Atomic I n d u s t r i a l Forum, Inc . American I n s t i t u t e o f S t e e l Const ruc t ion as low as i s reasonably achievable advanced l i g h t water r e a c t o r American Nuclear Soc ie ty American Na t iona l Standards I n s t i t u t e a n t i c i p a t e d opera t i ona l occurrence Advanced Reactor Severe Acc ident Program American Soc ie ty o f C i v i l Engineers a d j u s t a b l e speed d r i v e American Soc ie ty o f Mechanical Engineers American Soc ie ty f o r Tes t i ng and M a t e r i a l s a n t i c i p a t e d t r a n s i e n t ( s ) w i t h o u t scram American Welding Soc ie ty

b e s t est imate B i o l o g i c a l E f f e c t s o f I o n i z i n g Rad ia t ion , Committee on t h e balance o f p l a n t branch t e c h n i c a l p o s i t i o n Babcock and Wi lcox B r i t i s h Wire Gauge b o i 1 i ng water r e a c t o r B o i l i n g Water Reactor Owners Group

Category I Category I 1 computer-aided eng ineer ing compressed a i r and gas system c e n t r a l a larm system .

complementary cumula t ive d i s t r i b u t i o n f u n c t i o n c o n d i t i o n a l containment f a i l u r e p r o b a b i l i t y core coo l an t i n v e n t o r y c o n t r o l c l o s e d - c i r c u i t t e l e v i s i o n

Program Summary c- 1

ccw ccws CDF CDWS CE C ET CFR CIV CMEB

CMT COL c PU C RA C RD CRDM CRGR cs css CT CTS I cvc cvcs cws D/A DBA DBT DC DC DCH DDACS DEGB DHR DNBR DOD

DS DSER DWST

ECCS ECWS EDG EFU EFW EFWS EFWST EM I EMS EOF EOP E PA E PA

D-RAP

component cooling water component cooling water system core damage frequency chi 11 ed water system Combustion Engineering, Inc. comtainment event tree Code of Federal Requlations containment isolation valve former NRC Office of Nuclear Reactor Regulation Chemical Engineering Branch core makeup tank combined construction and operating 1 icense control processing unit control rod assembly control rod drive control rod drive mechanism Committee to Review Generic Requ containment spray containment spray system combustion turbine condensate treatment systems inf chemical and vol ume control

rements

uent

chemical and volume control system circulating water system

digital to analog des i gn-bas i s accident design-basis tornado design certification direct current direct containment heating digital data acquisition and control system doubl e-ended gui 11 ot i ne break decay heat removal departure from nucl eate boi 1 i ng ratio Department of Defense design reliability assurance program drywell spray draft safety evaluation report demi neral i zed water storage tank

emergency core cooling system essential chilled water system emergency diesel generator emergency f i 1 ter un i t emergency feedwater emergency feedwater system emergency feedwater storage tank electromagnetic interference environmental monitoring system emergency operations facility emergency operating procedure electric protective assembly Environmental Protection Agency

Program Summary c-2

EPG EPRI ERF ESF ESFAS ES I ESW ESWS

FDA FIVE FN FPCCS FPLC FPLCS FPS FSER

GDC GE GI G IMCS GIP GL G PM GRW PS GS I

HCLPF HEPA HF HPC I HPI HRA HVAC HWC

IASCC I &C ICC ICs I CST IDCOR IE I &E I EB I EC I EEE I ES I GA IGSCC I IT I LRT I MS

emergency procedures guideline Electric Power Research Institute emergency response facility engineered safety feature(s) engineered safety feature actuation system Energy Systems Group essential service water e s s e n t i a l serv ice water system

final design approval fire vunerability evaluation ferrite number fuel pool cooling and cleanup system fission product 1 eakage control fission product leakage control system fire protection system final safety eval uat i on report

general design criterion(a) General Electric Company generic issue generic issues management control system Generic Implementation Procedure generic 1 etter gal 1 on ( s ) per mi nute gaseous radioactive waste processing system generic safety issue

high confidence/low probability o f failure h i g h - e f f i c i en cy part i cu 1 at e a i r human factors high-pressure coolant injection high-pressure injection human reliability analysis heating, ventilating, and air conditioning hydrogen water chemistry

irradiation-assisted stress corrosion cracking instrumentation and controls inadequate core cooling. integrated control system influent to condensate storage.tank Industry Degraded Core Rulemaking Office of Inspection and Enforcement Office of Inspection and Enforcement Office of Inspection and Enforcement bulletin International Electrotechnical Commission Institute of Electrical and Electronics Engineers Illumination Engineering Society intergranul ar a t t a c k intergranular stress corrosion cracking incident investigation team integrated leak rate test information management system

Program Summary c-3

IN INPO I PCEA I PE I RWST IS1 I SLOCA I ST I TAAC

LBB LBHS LCO LCS LCS LDB LDR LLNL LOCA

LOFT LOOP LRB LRFD LRW PS LTOP LWR

MAA P MCC MCPR MC R MF MFW M-MI M-MIS MOV M PA MSIV MSIVLCS MTC MWD/MTU MWSE MWSG MWST

NCC \

NC IG NDE NDT NECWS NESWS NFPA NNS

NRC information notice Institute of Nuclear Power Operations Insulated Power Cable Engineers Association individual plant evaluation in-containment refueling water storage tank inservice inspection intersystem loss-of-cool ant accident inservice testing inspections, tests, analyses, and acceptance criteria

leak before break large-bore hydraulic snubber limiting condition(s) for operation 1 eakage control system local control station licensing design basis 1 oad def i ni ti on report Lawrence Livermore National Laboratory 1 oss-of-cool ant accident

1 oss-of-fl uid test l o s s o f o f f s i t e power licensing review basis load and resistance factor design liquid radioactive waste processing system low-temperature overpressure protection light water reactor

modular accident analysis program motor control center minimum critical power ratio main control room moderate frequency main feedwater man-machine interface man-machine interface system(s) motor-operator Val ve mu1 ti pl ant action main steam isolation valve main steam isolation valve leakage control system moderator temperature coefficient megawatt-day(s) per metric ton of uranium makeup water systems effluent makeup water to steam generator makeup water storage tank

natural convection cooldown Nuclear Construction Issues Group nondestructive examination nil ductility temperature nonessential chilled water system noneseential service water system National Fire Protection Association Non-nucl ear safety

Program Summary c-4

NPHS NPRDS NPSH NRC NS NSAC NSSS NUMARC N/VT

OBE ODCM

OSHA 0-RAP

PAP PASS PCCS PDHR PDHRS PGA PGC PGP P I N P I V PM PMF PMP PORV PRA PSD PS F P S I S PVC PW R PWSCC

QA

RA I RAM RAP RCA R C I C RC P RCPB RCS RF I RF PY RG RH R R I P RISCC ROM

normal power heat s i n k nuc lea r power p l a n t r e l i a b i l i t y da ta system n e t p o s i t i v e s u c t i o n head U . S . Nuc lear Regu la to ry Commission non-seismic Nuc lear Sa fe ty Ana lys i s Center n u c l e a r steam supply system Nuc lear U t i 1 i t y Management and Resources Counci 1 neut ron(s ) /square meter

ope ra t i ng -bas i s earthquake o f f s i t e dose c a l c u l a t i o n manual ope ra t i ons r e l i a b i l i t y assurance program Occupat ional Safe ty and Hea l th A d m i n i s t r a t i o n

personnel access p o r t a l pos tacc iden t sampling system pass ive containment c o o l i n g system pass i ve decay heat removal pass i ve decay heat removal system peak ground accel e r a t i on power genera t i on complex procedures genera t i on package p r o j e c t i n f o r m a t i o n network p ressure i s o l a t i on Val ve p r e v e n t i v e maintenance probab le maximum f l o o d probab le maximum p r e c i p i t a t i o n power-operated r e l i e f va l ve p r o b a b i l i s t i c r i s k assessment power spectrum d e n s i t y performance shaping f a c t o r pass i ve s a f e t y i n j e c t i o n system p o l y v i n y l c h l o r i d e p ressu r i zed water r e a c t o r p r imary water s t r e s s c o r r o s i o n c rack ing

qual i t y assurance

reques t f o r a d d i t i o n a l i n f o r m a t i o n random-access memory r e l i a b i l i t y assurance program r a d i o l o g i c a l c o n t r o l area r e a c t o r co re i s o l a t i o n c o o l i n g r e a c t o r coo lan t pump r e a c t o r c o o l a n t p ressure boundary r e a c t o r c o o l a n t system r a d i ofrequency i n t e r f e r e n c e r e a c t o r f u l l - p o w e r yea r r e g u l a t o r y gu ide r e s i d u a l heat removal r e a c t o r i n t e r n a l pump rad ia t i on - induced s t r e s s c o r r o s i o n c rack ing read-on ly memory

Progumam Summary c-5

RPS R P V RSDC RTB RTD

;kT RV RVLIS RWCS RWCU

SAFDL SAR SAS SBWR SCA S C C SDV SDV SDVS SECY SEP SER S FA SFC SG SG I SGOF SGTR SGTS SHARP S I S I S SLC SLCS SMB SOER SPDS SQAP SQUG S RM S ROA SRP S RV SSAR ssc SSE ss I ssw STCP SWAP sws

reactor protection system reactor pressure vessel reactor shutdown cooling reactor t r i p breaker resistance temperature detector nil d u c t i l i t y temperature reactor t r i p system reactor vessel reactor vessel level instrumentation system reactor water cleanup system reactor water cleanup

specified acceptable fuel design l imi t safety analysis report secondary a1 arm s t a t i on simplified boiling water reactor sneak c i r c u i t analysis s t r e s s corrosion cracking safety depressurization and vent scram discharge volume safety depressurization and vent system NRC Office of the Secretary (of the Commission) Systematic Eva1 ua t i on Program safety eval ua t i on report s i ngl e-f a i 1 ure anal ysi s s i ngl e-fai 1 ure c r i t e r i on steam generator safeguards information steam generator over f i l l steam generator tube rupture safety gas treatment system systematic human action r e l i a b i l i t y procedure safety injection safety injection system standby liquid control standby l iquid control system safety margin basis s ign i f icant operating event report safety parameter di spl ay system Sofware Qual i ty Assurance Program Seismic Qual i f ica t ion U t i l i t i e s Group s t a f f requirements memorandum(a) safety-related operator action S t a n d a r d Review Plan (NUREG-0800) safetylrel i ef Val ve standard safety analysis report s t ruc tures , systems, and components safe shutdown earthquake so i l - s t ruc ture interaction safety service water Source Term Code Package Service Water Assistance Program service water system

Program Summary C-6

TBCCWS TD I THERP TID TIP TM I

TS TSC

TMI-2

UBC UHS u PS U RS us I UT uv V V & V

ws

turbine building component cooling water system Transameri ca Del aval , Inc. technique for human error rate prediction technical information document traversing in-core probe Three Mile Island Three Mile Island Unit 2 technical specification(s) technical support center

Uniform Building Code ultimate heat sink uninterruptible power supply ultimate rupture strength unresolved safety i ssue ultrasonic test undervol tage

volt verification and validation

wetwell spray

Progr i m Summary c-7

APPENDIX D

PRINCIPAL CONTRIBUTORS

NRC Personnel

H. Asliar R . Borchardt B. Bo \ -d in ick J . Briimmer K. Carnpe C. Carpenter R. Cai-uso T. Chmdrasekaran T. Cheng M. Chiramal 0. Chopra R. Coi-reia R. Dube R. Eclcenrode A. E l -Bass ion i F. El. ;awila T. Es:;ig E. Fox J . Gal 1 agher R. Ga' l lo G. Georgiev C . Goodman J . Guo B. Hai-din C . Hi i ison T. H i ' l t z G. H s i i J. Joyce F . Kai i tor G. K e l l y S . K i l n T. K i l n L. Kopp J . K u d r i c k J . Lazevni c k E . Le(? J . Lee S . Lee A . L e ' i i n J . Le,/ ine C . L i Y . C . L i J . Lyons M. Ma' l loy J . M a r t i n

Review Area

s t r u c t u r a l eng ineer ing p r o j e c t management 1 egal counsel mechanical eng ineer ing p r o b a b l i s t i c r i s k assessment re1 i ab i 1 i t y assurance systems p l a n t systems s t r u c t u r a l eng ineer ing i n s t r u m e n t a t i o n and c o n t r o l s e l e c t r i c a l systems human f a c t o r s , r e l i a b i l i t y assurance p h y s i c a l s e c u r i t y human f a c t o r s p r o b a b l i s t i c r i s k assessment research ( p l a n t systems) r a d i a t i o n p r o t e c t i o n emergency p l anni ng i n s t r u m e n t a t i o n and c o n t r o l s opera tor l i c e n s i n g m a t e r i a l s engineer ing human f a c t o r s p l a n t systems research r a d i a t i o n p r o t e c t i o n p r o j e c t management r e a c t o r systems i n s t r u m e n t a t i o n and c o n t r o l s emergency p l anni ng p r o b a b l i s t i c r i s k assessment s t r u c t u r a l eng ineer ing p r o j e c t management r e a c t o r systems p l a n t systems e l e c t r i c a l systems mechanical engineeer ing r a d i a t i o n p r o t e c t i o n s t r u c t u r a l eng ineer ing r e a c t o r systems meteorology p l a n t systems mechanical eng ineer ing p l a n t systems p r o j e c t management r a d i a t i o n p r o t e c t i o n

Progr i m Summary D- 1

E. McKenna B. Mendel sohn J . Monninger J . Moore D. N o t l e y R. P a l l a K. Parczews k i L. P h i l l i p s R. Pichumani T. Pohida T. P o l i c h F. R i n a l d i R. Rothman M. Rubin G. Schwenk J . Sharkey P. Shea L. Shotk in B. Siege1 F. Skopec D. Smith P. Sobel L. S o f f e r I . S p i c k l e r J . Spraul J . Stewart D. Terao D. Thatcher G. Thomas C. T i n k l e r J . Tsao M . Waterman J. Wigginton F. W i t t R. Woods P. Worthington

Program Summary

qual i t y assurance phys i ca l s e c u r i t y p l a n t systems 1 egal counsel p l a n t systems p r o b a b l i s t i c r i s k assessment chemical engi n e e r i ng r e a c t o r systems s t r u c t u r a l eng ineer ing i ns t rumen ta t i on and c o n t r o l s re1 i abi 1 i t y assurance s t r u c t u r a l eng ineer ing geosciences r e a c t o r systems r e a c t o r systems re1 i abi 1 i t y assurance l i c e n s i n g ass is tance research ( p l an t systems) p r o j e c t management r a d i a t i on p r o t e c t i on human f a c t o r s geosciences research (source t e r m ) r a d i a t i o n p r o t e c t i o n qual i t y assurance ins t rumen ta t i on and c o n t r o l s mechanical eng ineer ing e l e c t r i c a l systems r e a c t o r systems research m a t e r i a l eng ineer ing i ns t rumen ta t i on and c o n t r o l s r a d i a t i o n p r o t e c t i o n chemical eng ineer ing research (gener ic s a f e t y i ssues) research ( p l a n t systems)

D- 2

APPENDIX E

COMMISSION PAPERS APPLICABLE TO ADVANCED LIGHT WATER REACTORS

SECY-77-439, " S i n g l e F a i l u r e C r i t e r i o n , " August 17, 1977.

SECY-136-228, " I n t r o d u c t i o n o f R e a l i s t i c Source Term Est imates I n t o L icens ing , " Augusi 6, 1986.

SECY-38-147, " I n t e g r a t i o n Plan f o r Closure o f Severe Acc ident Isues," May 25, 1988.

SECY-88-203, "Key L i c e n s i n g Issues Associated With DOE-Sponsored Advanced Reactor Designs," J u l y 15, 1988.

SECY-89-012, " S t a f f Plans f o r Acc ident Management Regulatory and Research Progr'ims, I' January 18, 1989.

SECY-89-013, "Design Requirements Related t o t h e E v o l u t i o n a r y Advanced L i g h t Water Reactors (ALWRs)," January 19, 1989.

SECY-39-153, "Severe Acc ident Design Features o f t h e Advanced B o i l i n g Water Reactor (ABWR)," May 10, 1989.

SECY-139-228, " D r a f t Sa fe ty Eva lua t ion Report on Chapter 5 o f t h e Advanced L i g h t Water Reactor Requirements Document," J u l y 28, 1989.

SECY-119-341, "Updated L i g h t Water Reactor (LWR) Source Term Methodology and Poten.;i a1 Regul a t o r y Appl i c a t i ons, 'I November 6, 1989.

SECY-!lO-016, " E v o l u t i o n a r y L i g h t Water Reactor (LWR) C e r t i f i c a t i o n Issues and T h e i r R e l a t i o n s h i p t o Cur ren t Regulatory Requirements," January 12, 1990.

SECY-!l0-241, "Level o f D e t a i l Required f o r Design C e r t i f i c a t i o n Under P a r t ! j2," July 11, 1990.

SECY-!l0-307, " Impacts o f Source Term Timing on NRC Regulatory P o s i t i o n s , " Augus.; 30, 1990.

SECY-!l0-313, "S ta tus o f Acc ident Management Programs and Plans f o r Implementa- t i o n , " September 5, 1990.

SECY-!l0-329, "Comparison o f t h e General E l e c t r i c Advanced B o i l i n g Water Reactor (ABWR) Design and t h e E l e c t r i c Power Research I n s t i t u t e ' s (EPRI's) Advanced L i g h t Water Reactor (ALWR) Requirements Document," September 20, 1990.

SECY-!l0-341, " S t a f f Study on Source Term Update and Decoupl ing S i t i n g f rom Design," October 4, 1990.

Progr im Summary E- 1

SECY-90-353, "L icens ing Review Basis f o r t h e Combustion Engineer ing, Inc. , System 8 0 t E v o l u t i o n a r y L i g h t Water Reactor," October 12, 1990.

SECY-90-377, "Requirements f o r Design C e r t i f i c a t i o n Under 10 CFR P a r t 52," November 8, 1990.

SECY-90-406, " Q u a r t e r l y Report on Emerging Technica l Concerns," December 17, 1990.

SECY-91-074, "Pro to type Decis ions f o r Advanced Reactor Designs;" March 19, 1991.

SECY-91-078, "Chapter 11 o f t h e E l e c t r i c Power Research I n s t i t u t e ' s (EPRI's) Requirements Document and A d d i t i o n a l E v o l u t i o n a r y L i g h t Water Reactor (LWR) C e r t i f i c a t i o n Issues," March 25, 1991.

SECY-91-135, "Conclusions o f t h e P r o b a b i l i s t i c Seismic Hazard Stud ies Conducted f o r Nuclear Power P l a n t s i n t h e Eastern U n i t e d Sta tes , " May 14, 1991.

SECY-91-178, " Inspec t ions , Tests, Analyses, and Acceptance C r i t e r i a ( ITAAC) f o r Design C e r t i f i c a t i o n s and Combined Licenses," June 12, 1991.

SECY-91-210, " Inspec t ions , Tests, Analyses, and Acceptance C r i t e r i a (ITAAC) Requirements f o r Design Review and Issuance o f a F i n a l Design Approval ," J u l y 16, 1991.

SECY-91-229, "Severe Acc ident M i t i g a t i o n Design A l t e r n a t i v e s f o r C e r t i f i e d Standard Designs," J u l y 31, 1991.

SECY-91-262, "Reso lu t ion o f Selected Technica l and Severe Acc ident Issues f o r E v o l u t i o n a r y L i g h t Water Reactor (LWR) Designs," August 16, 1991.

SECY-91-272, "Role o f Personnel and Advanced Cont ro l Rooms i n Fu ture Nuclear Power P lan ts , " August 27, 1991.

SECY-91-273, "Review o f t h e Vendor's Test Programs To Support t h e Design C e r t i f i c a t i o n o f Passive L i g h t Water Reactors," August 27, 1991.

SECY-91-292, " D i g i t a l Computer Systems f o r Advanced L i g h t Water Reactors," September 16, 1991.

SECY-91-348, " Issuance o f F i n a l Rev is ion t o Appendix J t o 10 CFR 50, and Reslated F i n a l Regulatory Guide 1 .XXX (MS 021-5)," October 25, 1991.

SECY-92-030, " I n t e g r a l System T e s t i n g Requirements f o r Westinghouse's AP600 P l a n t , " January 27, 1992.

SECY-92-037, "Need f o r NRC-Sponsored Conf i rmatory I n t e g r a l System T e s t i n g of t h e Westinghouse AP600 Design," January 31, 1992.

SECY-92-053, "Use o f Design Acceptance C r i t e r i a Dur ing 10 CFR P a r t 52 Design C e r t i f i c a t i o n Reviews," February 19, 1992.

Program Summary E- 2

SECY-!l2-092, "The Containment Performance Goal, Ex terna l Events Sequences, and t h e D f ! f i n i t i o n o f Containment F a i l u r e f o r Advanced L i g h t Water Reactors," March 17, 1992.

SECY-!l2-120, "NRC S t a f f Reviews f o r t h e Westinghouse AP600 and t h e General E l e c t r i c S i m p l i f i e d B o i l i n g Water Reactor (SBWR) Designs," A p r i l 7, 1992.

SECY-!)2-127, "Revised Acc ident Source Terms f o r L i g h t Water Nuclear Power Plant:;," A p r i l 10, 1992.

SECY-!l2-133, " D r a f t Sa fe ty Eva lua t ion Reports f o r Volume I and Volume 111 o f t h e E ' l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor Requirements Document," A p r i l 14, 1992.

SECY-!I2-134, "NRC C o n s t r u c t i o n I n s p e c t i o n Program f o r E v o l u t i o n a r y and Advanced Reactors Under 10 CFR P a r t 52," A p r i l 15, 1992.

SECY-!l2-170, "Rulemaking Procedures f o r Design C e r t i f i c a t i o n , " May 8, 1992.

SECY-!J2-172, " F i n a l Safe ty E v a l u a t i o n Report f o r Volume I 1 o f t h e E l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor Requirements Document," May 12, 1992.

SECY-!J2-211, "NRC Conf i rmatory I n t e g r a l System T e s t i n g f o r t h e General E l e c t r i c SBWR Design," June 5, 1992.

SECY-!J2-214, "Development o f Inspec t ions , Tests, Analyses, and Acceptance C r i t e r i a (ITAAC) f o r Design C e r t i f i c a t i o n s , " June 11, 1992.

SECY-!)2-219, "NRC-Sponsored Conf i rmatory T e s t i n g o f t h e Westinghouse AP600 Design," June 16, 1992.

Progr un Summary E-3

APPENDIX F

REPORT BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

P r o g r a m Summary

UNITED STATES NUCLEAR REGULATORY COMMISSION

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555

August 18, 1992

The Honorable Ivan Selin Chis irman U. ::. Nuclear Regulatory Comniission Washington, D.C. 20555

I Dedr Chairman Selin:

SU3JECT: ELECTRIC POWER RESEARCH INSTITUTE ADVANCED LIGHT WATER REACTOR UTILITY REQUIREMENTS DOCUMENT -- VOLUME 11, EVOLUTIONARY PLANTS

During the 387th and 388th meetings of the Advisory Committee on Re3ctor Safeguards, July 9-11 and August 6-8, 1992, we reviewed the NR3 staff's Final Safety Evaluation Report (FSER) for Volume I1 of th2 Electric Power Research Institute's (EPRI) Advanced Light Water Resctor (ALWR) Utility Requirements Document (URD) for Evolutionary Plants. Our Subcommittee on Improved Light Water Reactors held meetings on June 17-18 and July 27, 1992, to review this subject. During these meetings, we had the benefit of discussions with representatives of the NRC staff and EPRI. We also had the benefit of the documents referenced.

In the early 1980s, EPRI established the ALWR program to-support the United States utility industry efforts to ensure a viable nuclear power generation option for the 1990s and beyond. The objective of the program was to ensure that future nuclear power plants would be safer, simpler, more robust with greater margins, more easily operated and maintained, and more certain of being constructed and licensed without delays. This was accomplished using utility experience, by establishing design philosophy, producing design criteria and guidance to achieve the objective, and addressing the policies and regulations of the NRC.

The EPRI ALWR URD is a compendium of technical requirements for design, construction, and performance of ALWR nuclear power plants for the 1990s and beyond.

0 Volume I, "ALWR Policy and Summary of Top-Tier Requirements,Il is a management-level synopsis of the URD, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant.

The URD consists of three volumes:

Prog,*am Summary F- 1

The Honorable Ivan Selin 2 August 18, 1992

0 Volume 11, llALWR Evolutionary Plant, It consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1350 Mwe).

0 Volume 111, 'IALWR Passive Plant," contains utility design requirements for passive nuclear power plants (approximately 600 Mwe).

We have followed the development of the EPRI ALWR program from its inception and offered suggestions regarding safety improvements on several occasions. We also held numerous subcommittee and Committee meetings to consider and discuss the development of the EPRI URD program and the NRC staff's reviews.

The staff's review of the URD was conducted as described in "REG- 1197. As noted therein, the staff used NUREG-0800, "Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants,tt for review guidance. In addition, the staff's review reflects the requirements of 10 CFR 52, the Commission's policy statements on severe accidents, and the safety goals.

Although the SRP was used by the staff as guidance, the level of detail in the EPRI submittal did not permit a review of its completeness. (The SRP was written to support the review of safety analysis reports on specific plant designs for which a significant amount of design and construction information was available.) The staff conducted its review with the understanding that EPRI design criteria would meet all current regulations, except where deviations were identified. The staff's review of the URD focused primarily on determining whether the EPRI criteria did or did not conflict with current regulatory requirements.

In its review of Volume I1 of the URD, the staff identified a number of issues that will require additional information before the staff can reach a final conclusion. Initially, the staff divided the outstanding issues into three categories: (1) open policy issues on which the staff has proposed a position, but for which the Commission has not yet provided guidance, (2) open issues that must be satisfactorily resolved before the staff can complete its review of the URD, or (3) confirmatory issues for which the staff will ensure follow up of commitments in the URD.

At this date the staff has identified 21 open policy issues that are included in a draft Commission paper, ItIssues Pertaining to Evolutionary and Passive Light Water Reactors and Their Relationship to Current Regulatory Requirementstt that was issued on February 27, 1992. We provided our recommendations on the open policy issues pertaining to evolutionary plants in our letters which addressed SECY-90-016, SECY-91-078, and the draft SECY paper of February 7, 1992.

Program Summary F-2


The staff has handled the remaining 410 open issues which were identified in the FSER for Volume I1 by classifying them as Vendor or Utility Specific Itemst1 which must be satisfactorily addressed during the staff's review of a vendor- or utility-specific arplication. The staff plans to issue a supplement to the FSER after all evolutionary policy issues have reached final resolution. The staff indicated that they plan to interact with EPRI in an attempt to resolve significant open issues which may be resolved generically, and to include in a supplement any which are resolved.

We recommend generic resolution of as many of these issues as pcssible.

We commend EPRI for developing a comprehensive set of requirements. These will aid in the design of nuclear plants which will be safer, simpler, more robust, and more easily operated and maintained.

We commend the NRC staff for a very thorough review of the EPRI AIlWR Evolutionary URD, and its work with EPRI to identify and r€solve many issues relevant to licensing future LWRs. We recognize the NRC staff's position that its review necessarily is ir complete.

Sincerely,

David A. Ward Chairman

- References : 1. SECY-92-172, dated May 12, 1992, from James M. Taylor,

Executive Director for Operations, for the Commissioners, Subject: Final Safety Evaluation Report for Volume I1 of the Electric Power- Research Institute's Advanced Light Water Reactor Requirements Document, including the following enclosures: 0 Draft Safety Evaluation Report for Volume I, IIProgram

Summary of the NRC Review of the Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document," prepared by the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory' Commission, dated May 1992

0 Safety Evaluation Report for Volume 11, IINRC Review of the Electric Power Research Institute/s Advanced Light Water Reactor Utility Requirements Document for Evolutionary Plant Designs, prepared by the Off ice of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, dated May 1992

Prog,ram Summary F-3


2. Advanced Light Water Reactor Utility Requirements Document, Volume 11, "ALWR Evolutionary Plant," Chapters 1-13, through Revision 4, dated April 1992, Prepared for Electric Power Research Institute

Program Summary F-4

U S . NUCLEAR REGULATORY COMMISSION

NRCM 1102, 3201,3202 BIBLIOGRAPHIC DATA SHEET

(see instructions on the reverse)

NRC Review o f E l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor U t i l i t y Requirements Document

Program Summary

8. PERFORMING C RGANIZATION - NAME AND ADDRESS / i f NRC, provide Division, Officeor Region, U.S. Nuclear Regulatow Cot name and meillng II( dmrcJ

Associate D i r e c t o r a t e f o r Advanced Reactors and License Renewal O f f i c e o f Nucl ear Reactor Regul a t i on U. S. Nucl ear Regul a t o r y Commi s s i on Washington, D.C. 20555

1. REPORT NUMBER IAulonad by NRC. Add Vol., Supp., Rev., and Addendum Number:, If any.)

NUREG-I 242 V O l . 1

3. DATE REPORT PUBLISHED

M O N T H I YEAR 1

August 1992 1. F I N OR G R A N T NUMBER

6. TYPE OF REPORT Safety Eva1 u a t i o n Re pork

7 . PERIOD COVERED /Inclusive Dares1

tision, and mailing address; if contractor, provide

- NAME AND ADDRESS ( I f NRC. type "Same as above"; if contrxtor. provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing addrerc '

Same as above. I 1- RY NOTES

P r o j e c t Number 669 1 1. ABSTRACT 120 I words or bssJ

The s t a f f o f t h e U.S. Nuclear Regulatory Commission has prepared Volume 1 o f a sa fe ty e v a l u a t i o n r e p o r t (SER), "NRC Review o f E l e c t r i c Power Research I n s t i t u t e ' s Advanced L i g h t Water Reactor U t i 1 i t y Requirements Document - Program Summary, I' t o document t h e r e s u l t s o f i t s rev iew o f t h e E l e c t r i c Power Research I n s t i t u t e ' s "Advanced L i g h t Water Reactor U t i 1 i t y Requirements Document. I' This SER p rov ides a d i scussion o f t h e o v e r a l l purpose and scope o f t h e Requirements Document, t h e background o f t h e s t a f f ' s rev iew, t h e rev iew approach used by t h e s t a f f , and a summary o f t h e p o l i c y and t e c h n i c a l issues r a i s e d by t h e s t a f f d u r i n g i t s rev iew.

13. A V A I L A B I L I T Y STATEMENT 12. KEY WORDSJD ISCR!PTORS (List words orphnroer that willassist meamhem in locating the nepon.I

gdvanced 1 i g h t water r e a c t o r (ALWR) po l i c y i ssues Unl i m i t e d

~ f i n a l design approval ( F D A ) 10 CFR P a r t 52 Uncl ass i f i ed

U t i l i t y S tee r ing Committee U n c l a s s i f i e d

U t i 1 i t y Requirements Document (URD) o p t i m i z a t i on sub jec ts 14. SECURITY CLASSIFICATION

1 E l e c t r i c Power Research I n s t i t u t e (EPRI ) 1 i censabi 1 i t y (This Pagel

design c e r t i f i c a t i o n ( D C ) passi ve p l an ts /This Report1

.- combined opera t i ng l i c e n s e (COL)

s tandard i z a t i on

, Advanced Reactor Serv ice Accident Program ( A R S A P )

e v o l u t i o n a r y p l an ts 15. NUMBER OF PAGES

r e g u l a t o r y s t a b i l i z a t i o n 16. PRICE

NRC FORM 335 12-89)