ML20105C984
| ML20105C984 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| PROJECT-669A NUREG-1242, NUREG-1242-V01, NUREG-1242-V1, NUDOCS 9209240187 | |
| Download: ML20105C984 (138) | |
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s 4-NUREG-1242 Vol.1 i
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NRC Review of Electric Power Research Institute's Advanced Light Water Reactor j
Utility Rec uirements Document
-Program Summary
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Project Number 669 i
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U.S. Nuclear Regulatory Commission Office o^f Nuclear Reactor Regulation j
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1 AVAILABILITY NOTICE Availabihty of Reference Matnrfals Cited in NRC Pubhcations 4
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Most c'ocuments cited in NRC pubhcations will be avallsble from one of the following sources:
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The NRC Pubhc Document Room, 2120 L Stree', NW., Lower Level. Wsshington, DC l
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The Superintendent of Documents U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082 3.
T-National Technical inform 9 tion Service Springfield, VA 22161 l
Although thJ listing that follows represents the majority of documents cited ifi NRC pubhca.
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tions, it is not intended to be exhaustive, t
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N URiiG-1242 Vol.1 NRC Review of Electric Power Research Institute's Advanced Ligat Water Reactor Utility Rec uirements Document l'rogram Summary l'roject Number 669 Manuwript Completed: August 1992 llite Published August 1992 Associate Directorate for Advanced Iteactors and License llenewal Omce of Nuclear iteactor Regulation U.S. Nuclear Regulatory Coinmission Washington, DC 20555 p* = %,,
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ABSTRACT i
The Electric Power Research Institute (EPRI) is preparing a compendium of i
technical requirements, referred to as the " Advanced Light Water Reactor i
( ALWR] Utility Requireraents Document," that is applicable to the design of an ALWR power plant.
When completed, this document is intended to be a compro-hensive statement of utility requirements for the design, construction, and j
performance of an ALWR power plant for the 1990s and beyond.
The Requirements Document consists of three volumes. Volume 1, "ALWR Policy and Summary of Top-Tier Requirements," is a managament-level synopsis of the Requirements Document, 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 r
conceptual design state to a completed, functioning power plant.
Volume 11 consists of 13 chapters and contains utility design requiremerts for an evolu-tionary nuclear power plant [approximately 1350 megawatts-electric (MWe)].
Volume 111 contains utility design requirements for nuclear plar.ts for which passive features will be useet in their designs (approximately 600 MWe).
The staf f of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, has prepared Volumes I and 2 (Parts 1 and 2) of its safety evaluation report (SER) to ducument the results of its review of Volume I and 11 of the Requirements Document.
Volume 1, "NRC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summary," provides a discussion of the overall purpose and j
scope of the Requirements Document, the background of the staff's review, the
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review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.
Volume 2, "NRC Review of i
Electric Power Rasearch Institute's Advanced Light Water Reactor Utility Requirements Document - Evolutionary Plant Designs," gives the results of the i
staff's review of the 13 chapters of the Requirements Document for evolution-ary plant designs.
Voluma 3. "NRC Review of Electric Power Research Insti-l tute's Advanced Light Water Reacter Requirements Document - Passive Plant Designs," schedoled to be issued in Septecber 1993, wili give the results of the staff's review of tne 13 chapters of the Requirements Documcat for passive f
plant designs.
Prelimhary drafts of Volumes 1 and 2 were forwarded to the Commission and the l
Advisory Committee on Reactor Safeguards (ACPS' on May '/,1992.
In its letter dated April 24, 1992, the staff issued a draft of Volume 3 on all of l
the chapters of the Requirements Document for passive plant designs.
After l
the staff has completed its review of EFRI's responses to the draft SER (DSER) on passive plant designs in the form of revisions to the Requirements Docu-ment, it will issue a final SER to dir:uss its conclusions regarding its review of the final version of the document.
In staff requirements memoranda (SRH), the Commission instructed the staff to provide an analysis detailing where the staff propo:es departure from current regulations or where the staff is substantially supplementing or revising Program Summary iii y
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interpretive guidance applied to currently licensed LWRs.
The staff considers these to be policy issues.
Appendix B 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 B to the DSER on Chapter 1 of the Requirements Document for passive plant designs gives the regulatory analysis of those issues identified for the passive plant designs.
These issues have been addressed in Commission papers SECY-90-016, " Evolutionary Light Mater 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," at.d " Design Certification and Licensing Policy issues Portaining to Passive and Evolutionary Advanced Light Water Reactt.r Designs," that were issued on February 27 and July 6, 1992, resper-
- tively, in SRM dated June 26, 1990, and April 1, 1991, the Commission provided its decisions on SECY-90-016 and SLCY-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 to consider.
These issues are considered fundamental to agency decisions on the acceptability of the ALWR designs.
The staff will ensure satisfactory implementation of Commission guidance regarding these matters during its review of individual applicatio'is for final design approval and design certification.
There are no open issues pertaining to the Requirements Document fur evolu-tionary plant designs other than policy issues on which the staff has taken a position, but for which +.he Commission 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.
Program Summary iv
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TABLE OF CONTENTS Ease ABSTRACT iii 1
INTRODUCTION......
1-1 J
1.1 Background and Review Status 1-1 1.2 Purpose and Regulatory Status of EPh!'s ALWR Utility i
Requirements Document.....................
1-2 1.3 EPRI's Policy Statements 1-3 1.4 ALWR Design Bases.......................
1-4 1.5 Regulatory Stabilization
.-4 1.6 NRC Review Criteria......................
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1.7 Format and Availability of Documenti'lon i-7 i
2 POLICY ISSUES 2-1 j
3 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS.......
3-1 4 0UTSTANDING ISSUES.........................
- 4. l 4.1 Outstanding Issues Pertaining to the Pequirements Document for I-Evolutionary Plant Designs 4-1 4.2 Outstanding issues Pertaining to the Requirements Document for 4
Passive Plant Designs.....................
4-2 5 VENDOR-OR UTILITY-SPECIFIC ITEMS 5-1 1
l 5.1 Vendor-or Utility-Specific items Pertaining to the Re Document for Evolutionary Plant Designs....... quirements 5-1 5.2 Vendor-or Utility-Specific Items Pertaining to the Requirements Documant for Passive Plant Designs 5-21 I
6 CONCLUSION.............................
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APPENDIX A CHRON0 LOGY OF CORRESPONDENCE A-1 APPENDIX B REFERENCES B-1 APPENDIX C LIST OF ABBREVIATIONS..........,_....._...
. C-1 APPENDIX D PRINCIPAL CONTRIBUTORS D-1
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APPENDIX E COMMISSION PAPERS APPLICABLE TO ADVANCED LIGHT WATER REACTORS E-1 APPENDIX F REPORT BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS F..
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Program Summary' v
1 INTRODUCTION The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the " Advanced Light Water Reactor
[ALWR] Utility Requirements Document " that is applicabIe to the design of an ALWR power plant. When completed, this document is intended to be a compre-hensive statement of utility requirements for the det.ign, construction, and performance of an ALWR power plcat for the 1990s and beyond.
Those partici-pating in the program include utilities with nuclear plant experience, nuclear steam supply system vendors, architect-engineering firms, and con,ultants in related fields.
The Requirements Document consists of three vclumes.
Volume 1, "ALWR Policy and Summary of Top-Tier Requirements," is a management-level synopsis of the Requirements Document, including the decign 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 complettd, functioniag power plant.
Volume II l
consists of 13 chapters and c w tains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-olectric (MWe)]. Volume Ill contains utility design requirements for nuclear plants (approximately 600 MWe) in which passive features will be used for the ultimate safety protection of the plant.
The staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission (NRC), has prepared Volumes 1 and 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volume I and 11 of the Requirements Document.
Volume 1 "NPC Review of Electric Power Research Institute's Advanced Light Water Reactor Utility Requirements Document - Program Summary," provides a discussion of the overall purpose and scope of the Requirements Document, the background of the staff's review, the review approach used by the stuff, and a summary of the policy and technical issues raised by the staff during its review, Volume 2 "NRC Review of Electric Power Research Institute's Advanced Light Water-Reactor Utility Requirements Document - Evolutionary Plant Designs," gives the results of the staff's review of the 13 chapters of the Requirements Document 'for evolution-ary plant designs.
Volume 3, "NRC Review of Electric Power Research Insti-tute's Advanced Light Water Reactor Utility Requirements Document - Passive Plant Designs," scheduled to be issued in September 1993, will give the results of the staff's review of the 13 chapters of the Requirements Document for passive plant designs.
1.1 Backaround and Review Status in 1983, EPRI began its program by working with.the NRC staff to identify and resolve key safety and licensing issues. This joint effort resulted in a process whereby the unresolved and generic safety issues applicable to ALWRs Program Summary 1-1
as of July 1,1986* were identified.
This process was consistent with the procedures described in NUREG-0933, "A Prioritization of Generic Safety issues." Additional infornation about this effort and its results is provided in NUREG-1197," Advanced L ght Water Reactor Program - Program Management and Staff Review Methodology,;' dated December 1986.
In 1985, two new phases were added to the EPRI program:
the development of EPRI's ALWR Utility Require-ments Document for evolutionary plants and the assessment of small-plant options.
This assessment resulted in the development of the Requirements Document for passive plant designs.
Chronoloav of Revigw of Reauirements Document for Evolutionary Plant Desians from June 30, 1986, through October 26, 1989 EPRI submitted Revision 0 of the Requirements Document for evolutionary plant designs.
From September 24, 1987, through November 4, 1991, the staff developed and issued its draft SERs (DSERs) on these submittals.
lable 1.1 gives the dates when these documents were issued.
On September 7, 1990, EPRI submitted Revision 1 of the Requirements Document for evolutionary plant designs, modifying the document in its entirety.
On April 26 and November 25, 1991, and-Apri? 17, 1992, EPRI submitted Revi-sions 2, 3, and 4, respectively.
Volume 2 of this report addresses the Evolutionary Document fnr evolutionary plant designs through Revision 3.
Where possible, the staff's review included consideration of Revision 4 of the Requirements document.
A preliminary draft of Volume 2 was forwarded to the Commission and the Advisory Cohnittee on Reactor Safeguards (ACRS) on May 12, 1992.
The staff discussed the contents of the SER with the Committee and has included the views of the ACRS in Section 3 of this report.
Chronoloav of Review of Reauiren 'nts Document for Passive Plant Desioni 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 plrnt designs.
After the staff has 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.
1.2 Purpose and Reaulatory Status of EPRI's ALWR Utility Reauirements Document EPRI's ALWR Utility Requiremer' Jocument is designed to serve as a vehicle to obtain consisten' resolution of common operating problems, issues generically applicable to designs, severe-accident issues,: and certain unresolved and generic safety issues.
The document is to be used with companion documents, such as utility procurement specifications, tnat cover the remaining technical
- This date has since been changed to January 1990.
Program Summary 1-2
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requirements applicable to new plant projects.
It is also designed to serve i
as a vehicle to identify early in the design process major concerns about j
design concepts for LWRs in which passive safety systems will be used.
9 EPRI's ALWR Utility Requirements Document, because it is an agreement between j
the vendors and the nuclear power utilities, identifies what utilities desire a
in future designs.
The Requirements Document has no legal or regulatory status.
It is not intended to demonstrate complete compliance with the
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Commission's regulations, replatory guidance, or policies, nor is it intended j
to be used as a basis for supporting final design approval and design certifi-cation (FDA/DC) for a specific design.
Commission Guidance In its staff requirements memorandum (SRM) dated December 15, 1989, the Commission assigned the review of the Requirements Document for evolutionary plant designs priority equal to those of General Electric Company's Advanced j
Bolling Water Reactor and Combustion Engineering, Inc.'s System 80+.
In addition, the Commission directe' the staff to compare future designs against the Requirements Document for evolutionary plant designs.
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In the same SRM, the Commission directed the staff to complete its review of l;
the Requirements Document for passive plant designs before it submitted the results of its review of the licensing review basis (LRB) for passive designs
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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.0_gh of Federal Reaulations (10 CFR Part 52) negated the need for such a document, I
the staff interprets this guidance as directing it to complete its review of the Requirements Document for passive plant designs befor9 significant inter-action with the ACRS begins on these designs.
It is the staff's position i
l that, with the issuance of the DSER on Volume 3 in April 1992, this has been accomplished.
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In its-SRM of June 22, 1990, the Commission directed the staff to formally-resolve major technical and policy issues in the context of the review of the j
Requirements Document for passive plant designs.
The staff has identified
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such issues during its review of Volume !!! of the ALUR Utility Requirements Document and conceptual design information on the passive ALWRs.
The staff 1
developed the draft Commission papers dated February 27 and July 6,1992,. to 1
address resolution of these issues.
To ensure timely resolution of these policy issues, the staff will continue to evaluate resolutions-proposed by j
EPRI and the ALWR vendors, and will address them in future Commission papers.
1.3 EPRI's Policy Statements The ALWR Utility Steering Committee established policies to provide guidance for the overall develcoment of the EPRI's ALWR Utility Requirements Document i
and to provide the plant designer with guidance in applying the design criteria. Although nut design criteria themselves, these policies cover i
t fundamental principles that have a broad influence on the design criteria of the Requirements Document.
These policies include consideration of simplifi-j cation, design margin, heman factors, safety, regulatory stabilization, j-4 Program Summary 1-3
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I standardization, use of proven technology, maintainability, constructibility, j
quality assurance, economics, protection against sabotage, and environmental j
effects.
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j 1.4 ALWR Desian Basgi The term "ALWR design bases," as defined by EPRI, refers to the three sets of requirements tnat form the foundation for the ALWR design criteria.
The first set of requirements forms the " licensing design basis," which includes the 4
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requirements necessary to satisfy regulatory criteria. These requirements and j
j associated analytical methods are based on conservative, NRC-approved methods, and equipment is designed to safety-grade standards.
The second set is the l
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" risk-evaluation-basis," which extends the licensing design basis to meet public safety objectives.
Probabilistic risk assessments are used for the 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 l
methodology is used.
EPRI states that the licensing design basis is intended i
to provide an adequate level of safety, whereas the risk evaluation and j
performance design bases provide additional or enhanced protection.
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1.5 Reaulatory Stabilization Consistent with the overall ALWR program approach, as described in NUREG-1197, J
regulatory stabilization for an ALWR design can be achieved through the j
identification and resolution of plant optimization subjects and generic safety and licensing issues.
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Plant optimization subjects are proposals, initiated by EPRI', to deviate from regulatory 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. -These i
issues are identified for both the evolutionary and passive plant design criteria in Section 2 of Appendix B to Chapter 1 of Volumes II and III of the-l l
Requirements Document.
The staff's evaluation is provided in the correspond-ing sections of Volume 2 of this report and the April 1992 draft of Volume 3 j
of this report.
In addition, EPRI specifically addressed those generic safety issues that were
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classified as "high" or " medium" priority as of January 1990 in.Section 3 of Appendix B to Chapter 1 of Volumes 11 and 111 of the Requirements Document.
l The staff's evaluation of the generic safety issues specifically addressed by EPRI r.nd 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.-
i 1.6 NRC Review Criteria r
Criteria-Governina the Review of EPRI's'ALWR Utility Reauirements Document l
The staff's review of the Requirements Document is being conducted as described in NUREG-1197. As noted therein, the staff is using NUREG-0800, L
" Standard Review Plan [SRP) for the Review of Safety Analysis Reports for Program Summary 1-4 i
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f Nuclear Power Plants," for review guidance.
in addition to the cri % ia of i
NUREG-0800, the staff's review reflects the requirements of 10 CFR Psrt 52, "Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Reactors," and the Commission's " Policy Statement on Severe Reactor Accidents Regarding future Designs and Existing Plants" (50 TR 32138, August 8, 1985) and its policy statement on " Safety Goals for the Operations i
of Nuclear Power Plants" (51 FR 28044, August 4,1986).
The Requirements Document places primary emphasis on preventing significant problems that have been experienced in existing plants; however, many details that will be provided in specific design applications are missing. Therefore, the staff is reviewing the proposed requirements at-the level of detail pre-sented by EPRI but is not determining their adequacy to meet all NRC require-ments.
Although the SRP is being used as guidance, the level of detail does not permit a review for 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.
Therefore, the staff's evaluation became one of " review by exception," and the staff is con-ducting its review with the understanding that EPRI inte:.ds that the Require-ments Document contain design criteria that meet all current regulations, except where deviations are ider.tified in the document.
The staff's review of the Requirements Document is focused primarily on determining whether the EPRI criteria do or do not conflict with current regulatory requirements.
If a rer"irement proposed by EPRI is found not to conflict with NRC requirements, the staff judges it to be acceptable.
However, in certain technical areas (e.g., advanced instrumentation and control design, human factors considerations), little or no requirements or guidance exists.
Therefore, design criteria in the Requirements Document associated with these areas are unlikely to conflict with regulatory require-ments.
In those cases, although using the conflid standard would result in a finding of acceptability, the staff deems it appropriate to apply additional standards to-judge the acceptability of a requirement.
In such cases, the staff is using a combination of consideration of those regulations that do apply, good engineering judgment, past practice and experience at operating power plants, and applications in other industries to evaluate the require-ments proposed by EPRI in the Requirements Document.
During its review, the staff also identified potential incompatibilities l
between EPRI-proposed design requirements and current regulatory requirements, and where possible misinterpretations of regulatory requirements exist in the Requirements Document, In the February 15, 1991, SRM pertaining to SECY-90-377, " Requirements for Design Certification Under 10 CFR Part 52," the Commission. directed the staff to review the Requirements Document to ensure that it is1 sufficient to allow
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the' staff to evaluate the severe-accident issues and the incorporation of experience from operating events in current designs. The staff has used both deterministic and.probabilistic methods of evaluation, considering how-the Requirements Document addresses these issues through specific design criteria and through the guidelinc4 it provides for performing a probabilistic risk-assessment.
Program Summary 1-5
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in addition to addressing such matters relative to safety, the staff also has l
provided constructive comments on the document that, while not specifically regulatory in nature, would offer improvements in its requirements.
The staff I
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 l
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.
i Finally, as discussed in Section 1.2, the staff's review of the Requirements l
Document is not intended to substitute for any portion of the staff's review i
of future applications for final design approval and design certification.
l Additional Criteria Governina the Review of tO_.Reauirements Document for i
Passive Plant Desians i
The licensing design-basis analysis of the Requirements Document for passive plant designs relies solely en the passive safety systems to demonstrate i
compliance with the acceptance criteria for various design-basis transients i
and accidents.
Consequently, un rrtainties remain concerning the performance of the unique passive features and overall performance of core and containment heat removal because of lack of a proven operational performance history.
For
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example, uncertainties exist about the performance of check valves in the passive safety systems, which operate at low differential pressures provided
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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 in overcome stuck valves). As 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-cepth protection and the capability
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to prevent. accidents and core damage.
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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 i
well as safety analysis and testing required to demonstrate system capability i
for defense-in-depth considerations. However, the Requirements Document does
- l not provide specific requirements pertaining to the reliability of thera systems.
EPRI has indicated that it is evaluating the specific reliability targets and other measures to ensure that the passive plants will meet i
performance requirements and that it will address these safety concerns for i
both passive safety and active non-safety-related systems.
In addition, technical specification development is a subset of the overall
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regulatory treatment of the. passive designs. The staff is evaluating the need to establish reliability-based technical specification (TS) for the passive plant designs to determine which systemt and components (including certain i
non-safety-related systems) will require the imposition of TS, and the t
Program Summary 1-6
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parameters of the TS (limiting conditions for operation, surveillance, etc.).
j The reliability assurance program is expected to strongly influence TS.
I Since the passive ALWR design philosophy departs from current licensing i
practices, the staff has raised this issue to the Cummission as a policy l
issue.
The staff has not completed its review of this issue and, therefore, j
has not provided a recommendation to the Commission.
This issue is discussed further in the staff's regulatory departure analysis, which is given in Appendix B to the DSER on Chapter 1 of the Requirements Document for pass.ve plant designs.
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1.7 format and Availability of Documentation i
i Volume 1 of this report is a program summary of the NRC revit.w of EPRI's ALWR j
Utility Requirements Document.
It is not intended to provide a discussion of Volume 1 of the Requirements Document (EPRI submitted Volume I for information i
j only and did not request the staff to review it). Sections 1 through 5 provide a discussion of the overall purpose and scope of the Requirements j
Document and the rasults of the staff's review of that document.
Section 6 i
gives the staff's conclusions regarding the review of the Requirements Docu-ment for the evolutionary plant designs and the status of the review of the i
Hequirements Documents for the passive plant designs. Append'x A is a chro-i nology of the correspondence related to the review of the document. Appen-dix B contains the references for Volumes 1-3 of this report.
Appendix C is 4
a list of abbreviations used in Volumes 1-3 of this report.
Appendix D is a I
list of principal contributors.
Appendix E is a list of Commission papers that are applicable to the staff's review of the ALWR applications for FDA/DC.
Appendix f is a copy of the report of the Advisory Committee Reactor Scfe-guards on the ALWR Utility Requirements Document.
i The format of Volume 2 and the April 1992 draft of Volume 3 follows that of i
Volumes 11 and 111 of the Requirements Document as closely as possible, i
Copies of Volumes 1 and 2 of this report and the April 1992 draft of Volume 3 i
are available for inspection at the NRC Public Document Room, 2120 L Street, l
N.W., Washington, DC 20555.
The NRC project managers for the staff's review of EPRI's ALWR Utility l
Requirements Document are J. H. Wilson and T. J. Kenyon. They may be contact-ed by calling (301) 504-1118 or by writing to:
Associate Directorate for Advanced Reactors and License Renewal, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 4
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- NRC documents (e.g., NUREG reports -and regulatory guides) are not-included in Appendix B because they may be retrieved as indicated in the " Availability Notice" on the inside front cover of this report.
Program Summary 1-7 4
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Table 1.1 Chronoloay of the Review of LPRI's Requirements Document for Evolutionary Plant Designs 1
Chap-Date of Subuittal Date DSER ter Title of Ravision 0 Whs issued a
1 Overall Requirements June 30, 1986 September 24, 1987 July 8, 1986 February 18, 1988 1A PRA Key A2sumptions and June 30, 1989 Nov2mbar 4, 1991 Groundruler February 22, 1990 IB Licensing and Regulatory None issued None issued Requirements and Guidance 2
Power Generation Systems October 15, 19516 February 18.-1988 3
Reactor Coolant System and June 18, 1987 May 13, 1988 Reactor Non-Safety Auxiliary December 11, 1987 Systems 4
Reactor Systems June 1S,19f17 June 10, 1988 5
Engineered Safety Systems December 8, 1987 February 28, 1990 Bullcl ng Design and Arrange-November 18, 1988 January 15, 1991 6
i ment 7
Fueling and Refueling Systems February 28, 1989 January 15 01 8
Plant Cooling Water Systems December 30, 1988 January 1.,,
91 9
Site Support Systems January 11, 1989 January 15, 1991 10 Man-Machine Interface Systems October 26, 1989 October 8, 1991 11 Electric Power Systr.ms April 10, 1989 April 3, 1991 12 Radioactive Waste Processing.
December 23, 1988 January 15, 1991 Systems 13 Main Turbine-Generator dystems February 6, 1989 January 15, 1991 L
Pmgram Summary 1-8=
I I
Table 1,2 EPRJ 4roposed Plant Optimization Subjects-I Applicability l
Plant Optimization Subject Evoluti:> nary Passive 3.
l
%eratitig-Basis Earthquake and Dynamic X
.X Analysis Methods i
j Tornado Design X
X l
Boiling Water Reactor (BWR) Main Steamline X
X j
isolation Valves and Leakage Control Simplification of Postaccident Sampling X
X l
System Type C Containment Leakage Rate Testing X
X Interval h
Source Term Y
~X i
l Dedicated Containment Vent Pene.,, on X
X 1
i Mitigation of Anticipated Transients X
X i
Without Scram for the Advanced BWR l
Simplification of Offsite Emergency N/A X
j Planning Safe Shutdown of Passive ALWRs N/A X
l
[
N/A = not applicable i
i l
l i
i l
i Program Summary 1-9 i
(.
i 1
l 2 POLICY ISSUES 1
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 substanti-ally supplementing or revising interpretive guidance applied to currently licensed light water reactors.
The staff considers these to be policy issues.
Appendix B to Chapter 1 of Volume 2 of this report gives the staff's regula-i tory analysis of those issues identified for the evolutionary plant designs.
j Appendix B to the DSER on Chapter 1 of Voiume 3 of the Requirements Document
~;
gives the regulatory analysis of those issues identifi' l for the passive plant designs.
These issues have been addressed.in Commission papers SECY-90-016,-
" Evolutionary Light Water Reactor Certification Issues and Their Relationship to Current Regulatory Requirements," and SECY-91-078, " Chapter 11 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 3
Reactors and Their Relationship to Current Regulatory Requirements," and
" Design Certification and Licensing Policy Issues Pertaining to Passive and 4
Evolutionary Advance Light Water Reactor Designs," that were issued on l
February 27 and July 6, 1992, respectively.
In its SRMs dated June 26, 1990, and August 15, 1991, the Commission provided i
its decisions on SECY-90-016 and SECY-91-078 as they apply to evolutionary designs.
4 The February 27 and July 6, 1992, draft Commission papers have been forwarded to the Advisory Committee on Reactor Safeguards.
The staff will includa its i
views in the final paper and document its final positions before seeking Commission approval.
The approaches to resolving these issues have not been reviewed by the Commission, and, therefore, do not represent agency positions.
l These policy issues are considered fundamental to agency decisions on the acceotability of the ALWR designs. To aid in identifying its positions, the i
staff underlined those for which it requested Commission approval in the i
Commission papers discussed above. Table 2.1 contains a list of these policy j
issues and their applicability to the evolutionary and passive plant designs.
i Table 48.1 of Appendix B to Chapter 1 of Volume 2 of.this report lists the issues 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 5B.1 of Appendix B to Chapter 1 of the April 1992 draft i
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 Commission regarding policy issues that the staff has identified to date for both evolutionary and passive ALWRs. The staff developed these papers as a i
result of its review of the EPRI's ALWR Utility Requirements Document, the 4
l final design approval and design certification applications for the evolution-l ary plants, and the conceptual design information on the passive plants.
Program Summary 2-1
.. ~. -. - _.. _. - -. -._- -.
Table 2.1 Policy-Issues Pertaining to the Evolutionary and Passive Plant Designs Applicability Policy issue Evolutionary Passive Use of physically based source term X
X Anticipated transients without scram X
X Mid-loop operation X.
N/A Station blackout X
X Fire protection X-X Intersystem loss-of-coolant accident X
X Hydrogen control X
X Core-concrete interaction - Capability to X
X cool core debris i
High-pressure core melt ejection X
X Containment performance X
X i
l Dedicated containment vent penetration 1
X l
Equipment survivability X
X Elimination of operating-basis earthquake X
X Inservice testing of pumps and-valves -
X-X i
l_
Industry ccdes and standards X
X i
l Electrical -distribution X
X i
Seismic _ hazard curves
.X-
-X.
X Classification of main steamline-of.
X X
-boiling-water reactor L
Containment bypass X-X N/A = not applicable Program Summary 2...
Table 2.1 -(Continued)
Applicability Policy issue Evolationary Passive Tornado design basis X
X Containment bypass X-X Containment leak rate testing X
X Postaccident sampling system X
X Level of detail X
X Prototyping X
X Inspections, tests, analyses, and X
X acceptance criteria Probabilistic risk assessment-beyond X
X design certification Reliability assurance program X
X Severe-accident mitigation design.
X X
alternatives Generic rulemakina related to design X
X certification Regulatory treatment of non-safety-N/A X
related systems Definition of passive failure N/A-X Thermal-nydraulic stability of the.
N/A X
simplified boiling water reactor Safe shutdown requirements N/A X
Control room habitability N/A X
Radionuclide attenuation
'N/A X
Simplification of offsite emergency N/A X
planning N/A = not applicable Program Summary 2-3
....... -..... -. -.. -. -.. - - - - ~..
l-
}
l Table 2.1 (Continued)-
i
{
Applicability i.
Policy Issue Evolutionary Passive i
j Defense against common-mode failures in X
X j
digital instrumentation and control i
i Analysis of external' events beyond the X-
-X i
design basis Multiple stetm generator tubes ruptures N/A X
d j
Ra' 9 of the passive plant control room N/A X
i operator-i l
Control room annunciator reliability X
X lt l
N/A = not applicable i'.
}
4 i
i i
1 i
i l-4 i
i I
1 L
4 i
i.
}
i-l-
3 i
4 Program. Summary 2-4
k f
j l
3 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS I
In September 1984, the NRC staff, EPRI, and the Advisory Committee on Reactor Safeguards (ACRS) met to discuss early efforts leading to the development of EPRI's ALWR Requirements Document.
Since that time, these parties have met periodically to discuss the programmatic aspects and progress of EPRI's ALWR program.
The ACRS met with the staff and EPRI numerous times during the period Octo-ber 1987 through January 1992 to discuss the contents of each chapter of 3
Volume II of the ALWR Requirements Document and corresponding DSER after 4
issuance of the documents.
Each chapter and DSER was discussed with the ACRS Subcommittee on Improved Light Water Reactors one or more times.
In addition, many of the policy issues identified in Appendix B to Chapter 1 of Volume 2 were discussed with the Committee.
The reports of the-ACRS and the staff's responses to their comments on these policy matters'are in Annexes A-C of Appendix B to Chapter 1 of Volume 2 of this report.
The staff met with the ACRS Subcommittee on Improved Light Water Reactors to discuss the May 1992 draft of the SER on Volume II of the ALWR Requirements Document on June 17 and 18, and July 27, 1992.
The staff met with the ACRS Full Committee to discuss the results of its review of Volume II during the 387th and 388th meetings of the ACRS on July 10 and August 7, 1992, respec-tively. A copy of the ACRS report on Volume II of EPRI's ALWR Utility Requirements Document, dated August 18, 1992, is. attached as Appendix F.
e The staff will continue to interact with the ACRS during its review of policy l
issues and any additional issues on evolutionary ALWRs that may be resolved in i
the context of Volume I! of the ALWR Utility Requirements Document.
In addition, the staff wil: continue to interact with the ACRS during its review i
of Volume III of the ALWR Utility Requirements Document and will report the results of the Committee's review of that document and the staff's SER when it j
is completed.
I 2
Program Summary 3-1 l
N 4
4 OUTSTANDING ISSUES 4
During its review of EPRI's ALWR Utility Requirements Document, the staff.
identified several items for which additional information is required before the staff can reach a final conclusion.
The staff considers these issues to be outstanding.
These outstanding issues fall into one of three categories:-
(1) open policy issues on which the staff has taken a position, but for which j
the Comission has not provided guidance; (2) open issues that must be sati-sfactorily resolved before the staff can complete its review of the Require-4 ments Document; or (3) confirmatory issues for which the staff will ensure that EPRI meets it comitments to revise the Requirements Document. The outstanding issues for the entire Requirements Document, provided by chapter or appendix with references to appropriate sections of the SER chapter or j
appe, dix given in parentheses, are list.ed below.
The designators in front of each issue provide a unique identifier for each issue. The letter "E" or "P" indicates that the issue applies to. evolutionary or passive plant designs, respectively.
The first number designates the chapter in which it is identified. The letter that follows designates the appendix,.if applicable.
The letter "0" or "C" designates whether it is an open or a confirmatory issue, respectively. The final number is the sequen-tial number assigned to it in Section 1.4 of each chapter or appendix.
4.1 Outstandina Issues Pertainino M the Reouirements Document for Evolutionary Plant Desians 1
OPEN ISSUES The following is a list of open policy issues pertaining to the Requirements-Document for evolutionary plant designs (Volume II) on which the staff has taken a position, but #or which the Comission has not had the opportunity to i
provide guidance. There are no other types of open issues resulting from the staff's review of the Requirements Document for evolutionary plant designs.
Chapter 1 - Overall Reauirements E.1.0-1 tornado wind speeds (4.5 2)
E.1.0-2 leak before break (4.5.5)
Apoendix B to Chapter 1 - Licensino and Reaulatory Reauirements and Guidance 3
E lB.0-1 neact of the elimination of the operating-basis earthquake from the design process (2.1.1, Item IV. A of Annex A, and item I.M of Annex C, Item C of Annex D) i E.18.0-2 applicability of industry codes and standards (2.1.1 and Item II.A of Annex C)
E.lB.0-3 tornado design basis (2.1.2 and Item II.F of Annex C)
Program Sumary 1 1
i I
l i
E.lB.0-4 main steamline classification (2.3.1.1 and Item.II.E of Annex C) i E.18.0-5 simplitication of postaccident sampling system (2.3.2 and Item i
II.I of Annex C) 1 E.lB.0-6 containment leak rate testing (2.5.1 and Item II.H of Annex C) a i
E.lB.0-7 source term (2.5.2.1, 2.5.2.2, Item I.B of Annex A, and Item I.A l
of Annex C) i E.lB.0-8 seismic hazard curves (Item II.C of Annex C)
E.18.0-9 leak before break (Item 11.0 of Annex C)-
i l
E.lB.0-10 containment bypass (Item II.G of Annex C)
(
E.lB.0-ll prototyping (Item II.K of Annex C)
I E.lB.0-12 reliability assurance program (Item II.M of Annex C)
E.lB.0-13 defenst against common-mode fail ees in digital instrumentation l
and control systems (Item A of Annex D) i E.lB 0-14 analysis of external events beyond the-design basis (Item B of l
Annex D)
E.1B.0-15 control room annunciator reliability (Item G of Annex D)
Chaoter 5 - Enaineered Safety Systems E.5.0-1 core debris coolability (6.6.2)
CONFIRMATORY ISSVES There are no confirmatory issues pertaining to the Requirements Document for evolutionary plant designs.
4.2 Outstandina Issues Pertainino to the Reouirements Document for Passive Plant Desians OPEN ISSUES The following is a list of open issues that must be resolved before the staff f
can complete its review of the Requirements Document for passive plant designs
-(Volume III).
Chapter 1 - Overall Reouirements-
[
P.1.0-1 scope of mitigation requirements (2.1 and 2.4)
[
P.1.0-2 regulatory treatment of non-safety-related systems (2.3.1, 4.3.1, j
7, 10, 12.2.1, 12.2.3,- and Appendix 3) i t
Program Summary 4-2 1
1
]
J 5-i P.I.0-3 automatic standby liquid control system for passive boiling-water-reactor (BWR) design (2.3.2 and Appendix B) 1 l
P.l.0-4 check valve categorization (2.3.2) l P.I.0-5 tornado wind speeds (4.5.2.5) j P.l.0-6 leak before break (4.5.5)
P.1.0-7 seismic evaluation and design of small-bore piping (4.7.3) 1
{
P.l.0-8 use of Institute of Electrical and Electronics Engineers (IEEE)
Standard 323 (4.8.2) i P.l.0-9 method of environmental qualification of mechanical and electrical i
equipment (4.8.2) l P.l.0-10 limits on nitrites, nitrates, and total halogens as chlorine
[
(5.2.8) j P.l.0-Il pressurized-water-reactor (PWR) water chemistry (5.5.2)
]
P.l.0-12 reliability assurance program. framework (6.2) l P.1.0-13 quantitative reliability and availability goals (6.2) i P.l.0-14 integration of reliability-engineering techniques-(6.2,-6.3, and 6.4) l P.I.0-15 relationship of system requirements to overall plant safety reli-ability and availability goals (6.2).
l-P.l.0-16 difference between reliability assurance program for safety-and
[
non-safety-related systems (6.3) i
~
P.l.0-17 human factors considerations for. operation and maintenance provi-l sions (8.2) i j
P.l.0-18 computer security reference (11.12) f Apoendix / to Chapter 1 - PRA Kev-Assumotions and Groundrules P.lA.0-1 reporting' of core-damage-frequency results as mean values '(1.7)
P.1A.0-2 point-estimate quantification- (1.8)
P.lA.0-3 quantitative treatment of uncertainties (1.9)
P.lA.0-4 guidance on presenting results of probabilistic risk assessment (PRA) (1.10) e P.1A.0-5 guidance on modeling detail required to represent passive system l
behavior-(2.1) i i
1
-l Program Summary 4-3 1
1 1
i
J I
i P.lA.0-6 guidance on modeling interactions between passive and active systems (2.1) j P.lA.0-7 guidance for developing the success criteria for passive systems (2.3) l P.lA.0-8 determination of an appropriate mission time (2.10)
P.lA.0-9 requirements to address the important passive design-specific areas of uncertainty (6.1)
P.lA.0-10 failure rate for the main step-up transformers (8.2)
]
ADoendix B to Chaoter 1 - licensino and Reculatory Reouirements and Guidance l
P.lB.0-1 impact of the elimination of the OBE from the design process i
(2.1.1,3.3.1, and Item I.H of Annex A) 1 P.lB.0-2 applicability of industry codes and standards (2.1.1 and Item II.A j
of Annex A)
P.lB.0-3 tornado design basis (2.1.2 and Item II.F of Annex A) j i
P.lB.0-4 simplification of emergency planning requirements (2.1.3 and Item III.G of Annex A)
P.lB,0-5 need for leakage control system for main steam isolation valves (2.3.1 and Item II.E of Annex A)
P.lB.0-6 main steam isolation valve leakage rate (2.3.1 and Item II.E rf Annex A)
P.lB.0-7 simplification of postaccident sampling system (2.3.2 and' Item il.I of Annex A) i
{
P.lB.0-8 containment leak rate testing (2.5.1 and Item II.H of Annex A)
{
.P.lB.0-9 scurce term _(2.5.2 and Item I.A of Annex A) l P.lB.0-10 hydrogen generation and control (2.5.3, 3.2.26 and Item I.G of An-j nex A)
P.lB.0-11 safe shutdown requirements-(2.5.6 and-Item III.D of Annex A)-
~
P.lB.0-12 revised deficiency reporting requirements (Generic Safety Issue II.J.4.1) (3.2.3) i P.lB.0-13 criteria for safety-related operator actions (Generic Safety Issue B-17)-(3.2.5) i i
P.lB.0-14 diesel generator reliability (Generic Safety Issue B-56) (3.2.6) l P.18.0-15 allowable emergency core cooling system _ equipment outage periods (Generic Safety Issue B-61) (3.2.7)
Program Summary 4-4 2
P.lB.0-16 reactor coolant pump seal failures (Generic Safety Issue 23)
(3.2.10)
P.lB.0-17 bolting degradation or failure (Generic Letter 91-17) (Generic Safety Issue 23) (3.2.12)
P.lB.0-18 use of EPRI NP-5076 on good bolting practices for bolted joints (3.2.12)
P.lB.0-19 bolting degradation or failure (Generic Safety Issue 29) (3.2.12)
P.lB.0-20 effects of fire protection system actuation on safety-related equipment (Generic Safety Issue 57) (3.2.13)
P.lB.0-21 power-operated relief valve and block valve reliability (Generic Safety Issue 70) (3.2.14)
P.lB.0-22 anticipated transients without scram (Generic Safety Issue 75)
(3.2.16 and Item I.B of Annex A)
P.lB.0-23 unanalyzed reactor vessel thermal stress during natural convection cooldown (Generic Safety-Issue 79) (3.2.17)
P.lB.0-24
' control room habitability (3.2.18) i j
P.18.0-25 low-temperature overpressure protection (Generic Saf ty issue 94)
(3.2.21)
P.lB.0-26 piping and use of highly combustible gases in vital areas (Generic i
Safety Issue 106) (3.2.23) l P.lB.0-27 dynamic qualification and testing of large-bore hydraul, bbers (Generic Safety Issue 113) (3.2.24) l P.lB.0-28 reliability, operability, and on-line testability of tl+ final actuation contacts in protection systems (3.2.25) l P.lB.0-29 essential service water pump failures at multiplant sites-(Generic j
Safety Issue 130) (3.2.29)
P.18.0-30 guidelines for upgrading other procedures'(Generic Safety Issue-l HF 4.-4) (3.2.33)-
P.1B.0-31 electronic disp' lay of proc _edures,:ura nf mixed types of' procedures-from one. control station to another,_and use of active simulator to validate procedures (3.2.33) i P,18.0-32 clarification of definition of local control -stations (3.2.34):
l-P.lB.0-33 centralization of safety functions l(3.2.34) j P.18.0-34 development of a human factors verification and validation test
-a j
plan (3.2.35) r Program Summary 4-5
)
I
-. ~
P.lB.0-35
<acumentation of test activities for traceability and assurance l
that all human factors requirements are addressed during test and j
evaluation (3.2.35) t i
P.lB.0-36 development of quantitative measures to as.;ess human-system performance (3.2.35) l P.1B.0-37 uniform damping values (3.3.1)
P.lB.0-38 modal combination of high-frequency modes for vibratory loads (3.3.1)
P.lB.0-39 safety implications of control systems (Generic Safety Issue-A-47)
(3.3.3) t P.lB.0-40 heat exchanger testing (3.3.4)
[
P.lB.0-41 control of biofouling (3.3.4)
P.lB.0-42 zebra cussel fouling (3.3.4) f P.lB.0-43 mid-loop operation (Item I.C of Annex A)
P.18.0-44 station blackout (Item I.D of Annex A)
P.lB.0-45 fire proiection (Item I.D of Annex A)
}
i P.lB.0-46 intersystem loss-of-coolant-accident (Item I.E of Anr.ex A) t P.lB.0-47 core-concrete interaction - capability to cool core debris (Item I.G of Annex A) r P.lB.0-48 high-pressure core melt ejection (Item I.I of Annex A) i P.lB.0-49 containment performance (Item I.J of Annex A)
P.lB.0-50 dedicated containment vent penetration (Item I.K of Annex A) i P.lB.0-51 equipment survivability (Item I.L of Annex A)
P.lB.0-52 inservice testing of pumps and valves (Item I.N of Annex A)
E P.lB.0-53 electrical distribution (Item II.B of Annex A)
P.lB.0-54 seisniic hazard curves (Item II.C 'of Annex A)
P.lB.0-55 leak before break (Item II.D of Annex'A) i P.lB.0-56 classification of main steamline of BWB (Item II.E of Annex A)
P.lB.0-57 containment bypass (Item II.G of Annex A) f P.lB.0-58 level of detail (Item II.J of Annex A)
P.ogram Summary 4-6
+
P.lB.0-59 prototyping (Item II.K of Annex A) f P.lB 0-60 inspections, tests, analyses, and acceptance criteria _(Item 11.1.
I of Annex A) i P.lB.0-61 reliability assurance program (RAP) (Item II.M of Annex A)
P.lB.0-62 site-specific PRAs (Item II.N of Annex A)
P.lB.0-63 severe-accident mitigation design alternatives (Item II.O.of i
Annex A)
P.lB.0-64 generic rulemaking related to design certification (Item II.P of Annex A)
P.lB.0-65 regulatory treatment of non-safety systems (Item III.A of Annex A)
P.lB.0-66 definition of passive failure (Item III.B of Annex A) l P.lB.0-67 thermal-hydraulic stability of the Simplified Boiling Water Reactor (Item III.C of Annex A) l P lB.0-68 control room habitability (Item III.E of Annex A) f P.lB.0-69 radionuclide attenuation (Item III.F of Annex A) f i
[.hapter 2 - Power Generation Systems P.2.0-1 safety valve design (3.4) f Chapter 3 - Reactor Coolant System and Reactor Non-Safety Auxiliary Systems l
P.3.0-1 leak-testing of feec' water system valve that_ performs containment i
isolation function (5.5) l P.3.0-2 postaccident sampling system (7) l Chapter 4 - Reactor Systems i
P.4.0-1 rod insertion capability af_ter an earthquake (2.2.6) 7 P.4.0-2 inservice-inspection of reactor pressure vessel (RPV) internals i
(2.3.2)
P.4.0-3 RPV thermocouples (3.3) l Chapter 5 - Enoineered Safety-Systqtts l
P.5.0-1 justification for 72-hour design-oasis period for. control room-I habitability (2.1.1, 2.2, and 6.5) l P.5.0-2 need for activated charcoal filters (2.1.2)
P.5.0-3 timing of fission product release (2.1.3)
Program Summary 4-7
P.S 0-4 evaluation of aerosol fission product removal (2.1.6)
P.5.0-5 secondary building fission product holdup and plateout (2.1.7 and 6.4)
P.5.0-6 chemical form of iodine in containment (2.19)
P.5.0-7 guidance on vendor-supplied information (2.4.2)
P.5.0-8 identification of vital equipment (2.5)
P.5.0-9 use of carbon austenitic stainless steel for passive decay heat removal (PDHR) heat exchhnger piping material (3.3)
P.S.0-10 diverse reactor protection system (RPS) input to control rods (3.4)
P.S.0-Il definition of the safe shutdown condition (4.3)
P.5."-12 PDHR water pool capacity (4.3)
P.5.0-13 gravity drain tank for standby liquid control system (4.5)
P.5.0-14 separate connecting line for the two trains of the automatic depressurization system final stage'(5.4)
P.S.0-15 elimination of Type C leakage rate testing (5.2)
P.5,0-16 Type B air lock leak test interval requirements (6.3)
P.5.0-17 dose consequence criteria for design-basis cccidents (6.4).
P.S 0-18 maximum interval for Type C leakage rate testing (6.3)
P.5.0.
need for safety-grade containnent spray-system and engineered safety features atmosphere cleanup (2.1.6 and 6.4)
P.S.0-20 thyroid and beta skin radiation dose limits and credit for long-term use of breathing apparatus (6.5)
P.S.0-21 hydrogen concentration for PWR dry containment (6.6)
P.S.0-22 allowable compressive stress consistent with the American Society of Mechanical Engineers Boilec and Pressure Vessel Code (6.6)-
P.5.0-23 safety-grade hydrogen reccmbiners system (6.6)
P.5.0-24 containment steam bypass capability consistent with Standard Review Plan (7.2)
[hapter 6 - Buildina Desian and Arranaement P.6,0-1 classification of non-safety-grade auxiliary systems as ~ vital equipment-(2.3.3)
Program Summary 4-8
~
t i
I i
P.6,0-2 turbine-generator building safe shutdown earthquake loading condi-i tions (4.4)
P.6,0-3 location of the control complex (4.5.4) l P.6,0-4 requirements and acceptance criteria for human factors consider-ations (4.5.4)
Chapter 7 - Fuelina and Refuelina Systems P.7.0-1 nondestructive testing of the spent fuel pool liner (2.3.1) t P 7.0-2 inservice testing requirements for the spent _ fuel pool liner (2.3.1)
P.7.0-3 criticality of new fuel in new fuel storage facility (2.3.3) f P.7,0-4 radiological consequences of a fuel handling accident (2.3.9) i i
l P.7,0-5 source term for a fuel handling ace.ident (2.3.9) t P.7.0-6 incorrect water clarity requirement *eference (3.1.1) f Chapter 8 - Plant Coolina Water Systems f
P.8.0-1 regulatory treatment of non-safety-related active systems (2.2, i
3.1, 4.1, 5, and 7) j i
P.8.0-2 design requirements for the fuel pool cooling systnm (9) l t
Chapter 9 - Site Support Systems i
P.9.0-1 balance-of-plant fire protection program (3) i P.9.0-2 independence of ventilation systems inside the containment to pre-vent migration of smoke and hot gases (3.3.1)
P.9.0-3 requirements for smoke-removal capability _(3.3.1)_
P.9,0-4 safety classification for heating, ventilating, and air condition-ing (HVAC) systems (including pumphouse ventilation system) (8.2) j P.9.0-5 seismic Category II/I criteria for HVAC systems and component; (8.2.1 and 8.4.1)
P.9.0-6 use of post filter instead of high-efficiency particulate-air i
filters (8.2.1)
P.9.0-7 thyroid :.nd. beta rkin radiation dose requitements (8.2.2) f P.9,0-d control room habitability following design-basis accident (8.2.2)
[
P.9.0-9 omission of fans for warehouse areas (8.2.5)-
I l _
Program Summary-
'9 I
~,-
i P.9,0-10 omission of cooling coil supplied by chilled water for reactor cavity system (8.4.1)
P.9.0-ll use of Uniform Building Code, Zone 2A (8.5)
P.9.0-12 power source and performance requirements following a loss of-offsite power (8,5)
P.9.0-13 hVAC design for several areas not required t. meet requirements specified in the General Design Criteria, Standard Review Plan, or regulations (8.5)
Chapter 10 - Man-Machine Interface Systems (M-MIS)
P.10.0-1 independence of the software vetification and validation review-teams (3.1.2, 3.1.4, and 6.1.5; P.10.0-2 vse of commercial compilers for software used in safety systens (3.1.2)
P.10.0-3 dedication of commercial-grade software (3.1.2)
P.10.0-4 clarification of requirements for analysis and validation testing of M-MIS (3.1.3)
P.10.0-5 use of rapid prototyping in the development and validation of functional specifications (3.1.3)
P.10.0-6 operator aids (3.4.5)
P.10.0-7 establishment and use of reliability and availabiliy -N.timates (3.5)
P.10.0-8 definition of " practical" (3.5.3)
P.10.0-9 environmental conditions for equipment design including compati-bility with tasks (3.5.3)
P.10.0-10 component reliabiiity of M-MIS (3.5.4)
P.10.0-Il overall reliability of M-MIS (3.5.4) 2 P.10.0 sneak circuit analysis (3.5.4)
P.10.0-13 software maintenance (3.5.4)
P.10.0-14 minimum tests for continuous on-line testing (3.6.1).
P.10.0-15 vulnerability of power supplies for ale.rm systems (4.3.1)
P.10.0-16 guidance on criteria to establish priorities-(4.3.4)
P.10.0-17 guidance on the maximum number of alarms (4.3.4)
Program Summary 4-10
1 P.10.0-18 alarm sequence recording-(4.3.4)
P.10.0-19 valve position indication (4.4)
P.10.0-20 guidance on frequency allocation plan (4.6)
P.10.0-21 guidLnce on interference between communication-systems and M-HIS equipment (4.6)-
P.10.0-22 environmental conditions for minimally used local control stations (4.9.2)
P.10.0-23 guidance or, inadvertent actuation of controls at local control stations (4.9.2)
P.10.0-24 special physical security measures for the data transmission cable (5.2.1)
P.10.0-25 guidance on data system characteristics (5.2.2)
P.10.0-26 propagation of common-mode failures through the data system
_(5.2.2)
P.10.0-27 expan :;.1 capability of multiplexers (E.2.3)
P.10.0-28 reliability of multiplexing sy3 tem (5.4)
P.lc.0-29 software design aids and tools (6.1.1 and 6.1.5)
P.10.0-30 quality assurance requirements for safety-related software (6.1.2)
P.10.0-31 configuration management requirement for software (6.1.2)
P.10.0-32 software integrity (6.1.2)
P.10.0-33 guidance on software user documentation (6.1.2)
P.10.0-34 acceptance testing of commercially available software (6.1.2)-
P.10.0-35 notification of software errors or modifications of connercially delivered software products (6.1.2)
P.10.0-36 long-term configuration control of software (6.1.2)
P.10.0-37.
clarification of top-down' structured design approach (6.1.3)
P.10.0-38 guidance on convolutior, of sof tware structure (6.1.3)
P.10.0 behavior' of commercial software when assumptions are violated (6.1.3)
P.10.0-40 guidance on memory protection (6.1.3)
P.10.0-41 use of_ information by redundant safety cl.annels (6.1.3)
Program Summary 4-11
0 I.
4 j
P.10.0-42 guidance on tools to improve reliability and quality of software j
(6.1.5) i P.10.0-43 definition of reasonable testing and degree of _ confidence to be l
considered (6.1.5) l P.10.0-44 specification of the level of diversit-; in safety systems (6.1.6, 6.2.3, and 6.2.5)
J 4
P.10.0-45 guidance on performance of reliability evaluation -(6.1.6) 4 j
P.10.0-46 reference to IEEE 1050-1989 (6.2.2 and 6.2.9)
~
P.10.0-47 compatibility between M-MIS equipment and its external power supply systems (6.2.2) 3, P.10.0-48 alarmed, self-diagnostic feature on clock update (6.2.3) l P.10.0-49 guidance on position of sensor isolation valves (6.2.5)
I j
P.10.0-50 capacitance-type pressure sensors (6.2.5)
+
i P.10.0-51 minimal acceptance review criteria for isolation device (6.2.6) f P.10.0-52 voltage design of battery and de system (6.2.8) f P.10.0-53 standards for surge suppression (6.2.8)
P.10.0-54 electromagnetic interference /radiofrequency interference j
(EMI/RFI) considerations for wiring shields (6.2.9)
P.10.0-55 specific grounding standards (6.2.9) i' P.10.0-56 use of qualified isolators for wiring shields (6.2.9)
P.10.0-57 requirements for signal reconstruction (6.3.3)
P.10.0-CB use of interrupts (6.3.3) i l
P.10.0-59 redundancy of safety systems (6.3.4)
P.10.0-60 selection of automatic or manual control (7.2)
P.10.0-61 operation of_ plant oy load dispatcher (7.3.2)
P.10.0-62 alternate means of inventory monitoring (7.13)-
P.10.0-63 clarification of actuation logic self-testing (8.2.3)
P.10.0-64 initiation of standby liquid control system (8.7) i P.10.0-65 inclusion of four elements in human factuc: arganizational stru :-
ture (3.1.2 of Appendix B) 1 i
Program Summary 4-12
I l
P.10.0-66 scheduling of human factors studies before start of control room design process (3.1.3 of Appendix B).
P 10.0-67 guidance to the M-MIS designer to overcome past problems (3.1.4 of Appendix B)
P.10.0-68 guidance to improve interfaces between the operator and the plant (3.1.4 of Appendix B) i P.10.0-69 identification of human factors criteria, guidance, etc., that l
were used as the supporting bases for M-Mis requirements (3.2.1 of Appendix B)
P.10.0-70 traceability of human factors requirements to original source (3.2.2 of Appendix B) 3 P.10.0-71 method for establishing effective human factors rcquirements (3.2.3 of Appendix B) i P.10.0-72 guidance on systems analysis (3.2.4 of Appendix B) l P.10.0-73 organization of plant information (3.2.4 of Appendix B).
i j
P.10.0-74 configuration of operator's work:tation (5.2.5 of' Appendix B) l P.10.0-75 human factors guidelines for new technology (3.3 of Appendix B)
P.10.0-76 illumination levels (3.3-of Appandix B)
P.10.0-77 maintenance procedures (3.4 of Appendix B)'
P.10.0-78 selection and qualification of plant personnel (3.5 of Appendix B) i P.10.0-79 training requirements for top-level personnel (3.6 of Appendix B) l P.10.0-80 human factors verification and validation test plan (3.7.1 of Appendix B) 4 l
P.10.0-81 documentation of human factors test activities (3.7.2 of Ap-
]
pendix B)
P.10.0-82 team performance (3.7.4 of. Appendix B) l Chapter 11 - Electric Power Systems
~
4 P.11.0-1 regulatory treatment of active non-safety systems (2.2.1,- 3.2.2, i
5.2.1, 5.2.4, 5.2.5, and 8.2.1)
}-
P.11.0-2 clarification of terminology describing the role of_ non-safety j
systems-(2.2.1) j P.ll.0-3 loss of non-safety ac power during shutdown conditions (2.2.6)_
j i-1 l
Program Summary 4-13 t
b P.ll.0-4 second offsite power supply circuit for permanent non-safety load buses during power operations (3.2.1)
P.ll.0-5 consistency of Appendix A to Chapter 1 assumptions with Chapter 11 (3.2.3)
P.ll.0-6 data provided in Appendix A to Chapter 1 relative to loss-of-offsite-power events (3.2.5)
P.ll.0-7 non-safety ac electrical power systems during a small-break loss-of-coolant accident (5.2.1)
P.ll.0-8 clarification of revised requirements in Sections 1.5.2 and 5.3.1.1 (5.2.1)
P.ll.0-9 applicability of the 2000-hour rating and its impact on peaking operatica (5.2.1)
P.ll.0-10 peaking operation of non-safety standby power sources (5.2.3)
P.ll.0-Il consistency of failure and unavailability rates in Appendix A to Chapter 1 (5.2.4)
P.ll.0-12 justification for the use of independent self-contained cooling
,ystems (5.2.6)
P.ll.0-13 potential loss of de buses (7.2.1)
P.ll.0-14 lcad shedding for 72-hour battery endurance (7.2.3)
P.11.0-15 backup battery and battery charger (7.2.4)
P.ll.0-16 transfer scheme for the safety dc power supply system (7.2.5)
P.ll.0-17 battery and battery charger instrumentation and alarms (7.2.8)
P.11.0-18 lack of requirements for electric protective assenhlies in RPS power for BWRs (7.2.9)
Chaoter 12 --Radioactive Waste Processina Systems P.12.0-1 source term basis for designing radioactive waste systems and evaluation of offsite effluent radioactive nuclide concentration (2.2.2)
P.12.0-2 basis for 2-minute-delay requirement -for BWR turbine' gland seal system exhaust (3.3.1)
P.12.0-3 production sources for essentially nonradioactive steam (3.3.1)
P.12.0-4 discrepancy between Figure 12.3-1 and requirement in Chapter 13 (3.3.1)
Program Summary 4-14
P.12.0-5 use of ;,ost-filter downstream of charcoal adsorber in ventilation exhaust systems (3.3.3)
P.12.0-6 guidance regarding direct piping from radioactive plant systems to sumps or waste collection tanks (BWR) (4.2)
P.12.0-7 requirements for liquid radioactive waste processing sy.ctams (LRWPS)-filter housing and components (4.2)
P.12.0-8 requirements for LRWPS filters (4.2)
P.12.0-9 requirements fcr LRWPS ion exchangers (4.?.)
Chapter 13 - Main Turbine-Generator Systems P.13.0-1 turbine iss"a 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 ISStiji The following is a list of confirmatory issues for which the staff will ensure that rPRI meets its commitments to revise the Requirements Document for passise plant designs.
Chapter 1 - Overall Reauirements P.I.C-1 tornado wind speeds (4.5.2)
P.I.C-2 internal flooding design criteria (4.5.5)
P.l.C-3 compliance with Regulatory Guides 1.26 and 1.29 (9)
Accendix A to Chapter 1 - Key Assumptions and Groundrules None Accendix B to Chapter 1 - Licensina and Reaulatory Reauirements and Guidance P.1B.C-1 commitments to 10 CFR Part 21 and 50.55(e) (3.2.3)
P.18.C-2 Target Rock safety / relief valves (3.2.5)
Chapter 2 - Power Generation Systems None Chapter 3 - Reactor Coolant System and Reactor Non-Safety Auxiliary Systems P.3.C-1 anticipated-transients-without-scrams events-(5.5)
Chapter 4 - Reactor Systems Program Suamary 4-15
None Chapter 5 - Er.aineered Safety Systemi None
]
Chapter 6 - Buildino Desian and Arranaement None Chapter 7 - Fuelina and Refuelina Systems P 7.C-1 quality group classification of components for the-new and spent fuel storage racks (2.3.1)
Chapter 8 - Plant Coolina Water Systems None Ch&oter 9 - Site Support Systems P.9.C.1 guidance on the need for assessing the area interior to security detection equipment (5.2.7)
Chapter Man-Machine-Interface Systems P.10.C-1 intention of use (1)
P.10.C-2 -impact of support software on installed software (6.1.2)
P.10.C-3 software common-mode failures (6.1.6, 6.2.5, and 8.2.1)
P.10.C-4 BWR depressurization system design change (8.6)
P.10.C-5 mannine of M-MIS that controls security functions (10.2.3)
Chapter 11 - Electric Power Systems P.11.C-1 isolation of safety systetus from their non-safety sources (2.2.5)
P.11.C-2 minimum starting voltages for valve actuator motors (2.2.8)
P.ll.C-3 requirements for main step-up transformers (3.2.6)
P.11.C-4 harmonic distortion effect of adjustable speed drives (4.2.3)
P.ll.C-5 location of non-safety standby ac power sources (5.2.2)
P.11.C-6
. sizing of thermal overload devices (6.2.1)
P.11.C-7 battery sizing criteria specified for 72-hour coping capability (7.2.2)
P.ll.C-8 shysical and electrical separation of the safety de and low-voltage Program Summary 4-16
/I
vital ac power supply system (7.2.5)
P.ll.C-9 non-safety de power supplies for switchyard circuit protection and control equipment (7.2.6)
P ll.C-10 battery capacity margin (7.2.7)
P.ll.C-Il normal and emergency lighting following design-basis events (8.2.1)
P.ll.C-12 intensity of emergency lighting system (8.2.2)
Chapter 12 - Radioactive Waste Processina Systems.
None (h.aoter 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) s Program Summary 4-17
k l
5 VENDOR-OR UTILITY-SPECIFIC ITEMS During its review of EPRI's ALWR Utility Requirements Document, the staff l
identified items that were inadequately addressed in the document or were issues that could not be addressed generically.
These items will have to be satisfactorily resolved during the staff's review of a vendor-or utility-specific application (i.e., an application for final design approval and design certification or a combined construction permit and operating license).
As discussed in Section-i.2 d 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 poli-j cies.
It is not intended to be used as a basis for supporting final design approval and design certification for a specific design,. nor is it to be used j
to substitute for any nortion of the staff's review of future applications for i
final design approval e,J design certification.
Specifically, satisfactory resolution of the items identified in this saction for a vendor-or utility-specific application will not, by itself, support a finding that the applica-tion complies with the Commission's regulatory requirements. The staf' will perform a complete licensing review of these applications using the Standard Review Plan (NUREG-0800) and other appropriate Commission guidance.
Satisfac-tory resolut~ ion of the vendor-or utility-specific items constitutes only or.e 3
j portion of the staff's review.
The vendor-or utility-specific items for the entire Requirements Document, i
provided by chapter or appendix with references to appropriate sections of the SER chapter or appendix given in parentheses, are listed below. The designat-4 ors in front of each issue provide a unique identifier for each issue. The letter "E" or "P" indicates that the issue applies to evolutionary-or passive plant designs, respectively.
The first number designatt s the chapter in which it is identified.
The letter that follows designates the appendix,-if l
applicable. The letter V designates that it is a vendor-or utility-specific j
item.
The final number provides the sequential number assigned to it in-i Section 1.5-of each chapter or appendix.
5.1 Vendor-or Utility-Soecific Items Pertainino to-the Reouirements Document i
for Evolutionary Plant Desions The following is a list of vendor-or utility-specific items that are identi-fied in the SER (Volume 2 of this report) on the Requirements Document for evolutionary-plant designs.
Chapter 1 - Overall Reouirements E.1.V-1 scope of mitigation features (2.1 and 2.4)
E.1.V-2 implementation of design characteristics intended to enhance acci-dent resistance (2.2) 4 i
E.1.V-3 bounding analysis by standard site design parameters (2.3.1)
Program Summary 5-1
i i
E.1.V-4 selection of initiating cvents and their frequency categorization (2.3.2)
E.1.V-5 acceptance criteria for transient and accident analysis (2.3.2)
E.1 V-6 anticipated transient without scram response anal > sis (2.3.2) l E.1.V-7 acceptability of analytical codes and methodologies for safety analysis (2.5) i E.1.V-8 60-year plant life (3.3, 4.8.2, 0.2, and 11.3)
E.1.V-9 operation of PWR with a secured reactor coolant pump (3.5)
E.1.V-10 defense-in-depth analysis (3.5) j E.1.V-Il event response capability (3.5)
I E.1.V-12 fuel burnup requirements (3.6)
[
i E.1.V-13 extended operating life of control blades and control od assem-blies (3.6) l E.1.V-14 safety classification (4.3.1) j E.1.V-15 seismic qualification by experience (4.3.2 and 4.8.1) t E.'.V 16 non-seismic building structures (4.3.2 and 4.7.2)
-[
E.1.v structural design and construction codes (4.4 and 4.4.1)
E.1.V-18 climination of operating-basis earthquake from design (4.4.3, 4.7.3, and Appendix B)
}
E.1.V-19 definition of support group (4.4.3)
E.1.V-20 use of Appendix N of ASME Code,Section III (4.4.3 and 4.7.3)
E.1.V-21 analysis of vibratory loads with significant high-frequency input (4.4.3)-
E.1.V-22 use of nonlinear analysis to account for gaps between pipes and piping supports (4.4.3) t E.1.V-23 probabil'istic approach for-changing existing loads and/or loading I
combinations.(4.5.1)
[
E.1.V-24 recurrence interval for wind loadings (4.5.2)
E.1.V-25 maximum ground water level (4.5.2) i i'
E.1.V-26 precipitation for roof design (4.5.2)
E.1.V-27 snow loading (4.5.2) l Program Summary 5-2 r
a
.,...w.
a..
.-..a-,
l 1
E.1.V-28 detailed quantification of soil parameters (4.5.2)-
3 E.1.V-29 minimum margin against liquefaction (4.5.2) f E.1.V-30 external hazards evaluation (4.5.2) i E.1.V-31 number of full-stress cycles (4.5.2 and 4.8.1) i E.1.V-32 site-specific SSE (4.5.2)
[
E.1.V-33 power spectrum density function of the time history (4.5.2)
E.1.V-34 external impact hazards (4.5.2) i E.1.V-35 design temperature (4.5.2) i E.1 V-38 protection against surface vehicle bombs (4.5.3) i E.1.V-39 BWR safety / relief valve loads (4.5.4) i E,1.V-40 NUREG-1061 methodology and acceptance criteria for leak before
[
break (4.5.5)
E.1.V hydrodynamic loads from safety / relief valves (4.5.5)
E.1.V-42 suppression pool dynamic-loads (4.5.5)
E.1.V-43 design against internal-missile generation (4.5.5)
E.1.V-44 design of concrete containment (4.6.1) 4 1
E.1.V-45 load combinations _for seismic Category I buildings and structures (4.6.1)
E.1.V-46 design of seismic Category I steel structures (4.6.1) f E.1.V-47 combination of pipe re cure loads with seismic' loads for seismic
}
Category I structures (4.5.1 and 4.6.1)
(
E.1.V-48 combination of loss-of-coolant-accident and SSE loads (4.6.1)
[
l E.1.V-49 load combinations for safety-releted portions of the plant (4.6.2) t E.1.V-50 dysmic analysis technique; (4.7.2) j E.1.V-51 methodology for generating design response spectra or time histo-i ries (4.7.2)
?
?
E.1.V-52 structural damping values (4.7.2) l E.1.V-53 masonry walls in Category I buildings (4.7.2)
E.1.V-54 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 stability of shell-type structures under compression (4.7.2)
E.1.V-56 seismir. 2 valuation and design of small-bore piping (4.7.3)
E.1.V-57 use of ASME Code Case N-411 (4.7.3)
E.1.V-58 use of ASME Code Cases N-411 and N-420 in same analysis (4.7.3)
E.1.V-59 construction of core support structures (4.7.3)
E.1.V-60 fatigue design curves (4.7.3)
E.1.V-61 use of IEEE 323 (4.8.2)
E.1.V-62 environmental qualification of mechanical and electrical equipment (4.8.2)
E.1.V-63 use of zinc to reduce radiation fields (5.2.7)
E.1.V-64 limits on nitrites, nitrates, and total halogens as chlorine (5.2.8)
E.1.V-65 grinding controls for PWRs (5.3.1)
E.1.V-66 effect of fabrication processes on i_ntergranular stress corrosion cracking (5.3.1 and 5.3.1)
E.1.V-67 hardness limits for stainless steel (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) l 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 reliability assurance program (0-RAP) (6)
E.1.V-75 organizational description for reliability assurance program (6.1)
E.1.V-76 analyses methods or models used in developing the reliability assurance program (6.2)
E.1.V-77 reliability data bases (6.2)
E.1.V-78 reliability, maintainability,= and testability analyses (6.2)
E.1.V-79 apportionment of contributions _of structures, systems, and compo-nents to core damage frequency (6.3)
Program Summary 5-4
E.1.V-80 priority of safety in accident recovery (6.3) y.
E.1.V-81 relationship between safety and production availability (6.3)
E.1.V-82 eft of limitations on refueling duration on plant safety (6.
E.1.V-83 effect of planned outage duration on plant safety (6.3)
E.1.V-84 effei.t of major outage duration on plant safety (6.3)
E.1.V-85 inspection of construction activities (7 and 11.13)
E.1.V-86 quality assurance for non-safety-related facilities and systems (7)
E.1.V-87 installea operating-phase security system (7)
E.1.V-88 reliability of modular construction (7)
E.1.V-89 use of IEEE P1025 P1023-1988/D5 and EPRI-2360 for guidance regard-ing human factors engineering (8.2)
E.1.V 90 inspection and verification of security locks robotically (8.3)
E.1.V-91 quality assurance requirements for all equipment, structures, systems, facilit hs or software that have some safety importance or has one importance (9)
E.1.V-92 compliance of FDA/DC applications oth Commusion's regulations and guidhnce (10)
E.1.V-93 issue resolution for FDA/DC reviews (10)
E.1.V-94 inspections, tests, analyses, and acceptance criteria (10)
E.1.V-95 implementation of simplification objective (11.4)
E.1.V-96 implementation of standardization objective (11.5)
E.1.V-97 check valve testing methods (12.2.2)
E.1.V-98 full-flow testing of check valves (12.2.2)
E.1.V-99 qualification testing of active and nonactive motor-operated valves (MOVs) (12.2.2)
E.1.V-100 technical concerns regarding MOVs (12.2.2)
E.1.V-101 leak rate testing for individual containment isol:. tion valves (12.2.2)
E.1.V-102 instrumentation to determine net positive tuction head during all modes of operation (12.2.3)
E.1.V-103
+" sting of pump flow rate (12.2.3)
Program Summary 5-5
E.1.V-104 frequency and extent of disassembly and inspection of safety-related pumps (12.2.3) appendix A to Chapter 1 DDA Key Assumptions and Groundrules E.lA.V-1 use of PRA in design E.1A.V-2 modeling of a PRA (1.6)
E.1A.V-3 shutdown and low-power events (1.6)
E.lA.V-4 external events (1.6, 3.3, and 6.1)
E.1A.V-5 core damage frequency (1.7)
E.1A.V-6 uncertainty treatment (1.9 and 6.1)
E.lA.V 7 documentation of method of truncation of accident sequences (1.10 and 2.5)
E.lA.V-8 low-frequency accident initiators leading to core damage (2.2)
E.lA Y-9 mission time (2.10)
E.lA.V-10 f allare rate for components (2.11)
E.lA.V-11 tornadoes and extreme winds (3.2)
E.lA.V-12 external river flooding (3.2)
E.lA.V-13 hurricanes and storm surges (3.2)
E.lA.V-14 tsunami (3.2)
E.lA.V-15 internal fires (3.2)
E.lA.V 16 site-specific extern 11 events E.lA.V-17 internal flooding (3.2)
E.lA.V-18 seismic hazards analysis (3.3)
E.lA.V-19 corr damage-sequence binning (4.1)
E.lA.V-20 plant damage state definition (4.2)
E.lA.V-21 containment isolation assumptions and' criteria (4.3)
E.lA.V-22 in-plant sequence assessment (4.5)
E.lA.V-23 containment event analysis (4.6) e E.lA.V-24 details of uncertainty analysis (4.6)
Program Summary 5-6
-r I
t I
E.lA.V-25 source term definition (4.7)
I 1
E.lA.V-26 event tree binnia.; (1.8) i I
E.lA.V-27 risk measures related to containment performance (4.8) l E.lA.V-28 use of mean values for characterization of risk results (5.1) l E.lA.V-29 assessment of risk measures (5.2) i l
1 E.lA.V-30 calculation of offsite consequences (5,2)
{
E.lA.V-31 importance analysis for input to reliability ar';urance program I
(6.1) s E.lA.V-32 assessment of containment response (6.2) i I
E.lA.V-33 source term (6.3) i E.lA.V-34 scope and objective of human reliability analysis (HRA)(7.1) i E.1A.V-35 process and criteria to confirm adequacy of human reliabilh i
analysis (HRA) (7.2)
E.1 A. V.16 impact of advanced technologies on HRA (7.3) l L
E.1A.V-37 function, task, timeline, and link analy.as (7.3) i E.lA.V-38 generic data sources (7.3)-
i E.lA.V-39 performance shaping factors and their evaluation tools (7.3) j E lA.V-40 quantification methods for HRA (7.3) t E.1A.V-41 loss of offsite power frequency (Annex A) l l
E.lA.V-42 site data (Annex B)
Appendix B to Chapter 1 - licensina and Reaulatory Reauirements and Guid?I;,.g j
E.lB.V-1 compliance of fDA/DC applications with Commission's regulations and
]
guidance (1.3) l E.lB.V-2 issue resolution for FDA/DC reviews (1,3)-
E.lB.V-3 elimination _of missile provisions (2.1.2)-
F E.lB.V-4 dynamic seismic analysis of main steam piping and-condenser (2.3.1.1 and Item II.E of Annex C) j E.lB.V-5 main steamline classification (2.3.1.1 and item II.E of. Annex C) j E.lB.V-6 seismic analysis and plant walkdown of turbine building (2.3.1.1 and item II.E of Annex C) t i
Program Summary 5-7
k i
j E.lB.V-7 plateout considerations for main steam piping and valves (2.3.1.2 l
and Item Ill.F of Annex C) j i
i E.lB.V-8 reactor pre:sure vessel level instrumentation system (2.4.1) t
}
E.18.V-9 source term (2.5.2.1, 2..i 2.2, item 1.B of Annex A, and item I. A of i
Annex C) 1 l
E.lB.V-10 compliance with Branch Technical Position MTEB 6.1 (2.5.2.2) l E.lB.V-11 fission product cleanup analysis (2.5.2.2)
I l-E.lB.V-12 deletion of charcoal arisorbers (2.5.2.2) t i
E.lB.V-13 dedicated containment. vont penetration (2.5.3 and item 1.K of j
Annex C) 1 i
E.lB.V-14 decoupling of operating-basis earthquake (0BE) from safe shutdown earthquake (SSE) in seismic design of structures (Generic Safety i-Issue A-40) (3.2.7)
{
E.lB.V-15 deletion of OBE damping values in seismic design of_ structures j
-(Generic Safety Issue A-40) (3.2.7) i j
E.lB.V-16 use of algebraic sum method for modal combination of high-frequency modes-for vibratory loads (Generic Safety Issue A-40) (3.2.7)
}
E lB.V-17 use of spectral peak shifting techniques in lieu of spectral i
broadening (Generic Safety issue A-40) (3.2.7).
E.18.V-18 plant-specific design and arrangement of control systems (Generic Safety Issue A-47) (3.2.9) i E.lB.V-19 conformance to 10 CFR 50.34(f) hydrogen control requirements (Generic Safety issues A-48 and 121)-(3.2.10 and 3.2.46)
E.lB.V-20 reliability of emergency diesel generators (Generic Safety Is-l sue B-56) (3.2.14) i E.lB.V-21 resolution of Generic Safety issues 2 and 110 (3.2.18 and 3.2.42) l E.lB.V-22 -resolution o' Generic Safety issue 15 (3.2.19)
' ndependent reactor coolant-pump seal cooling during station i
E.lB.V-23 blackout (Generic Safety Issue 23) (3.2.20) i E.lB.V-24 resolution of Generic Rfety issue 24 (3.2.21)
E lB.V-25' design details on threaded fasteners (Generic Safety Issue 29)
(3.2.22)
ErlB.V-26 reduction of biofouling in open-cycle service water and component cooling water systems (Generic Safety Issue 51)-(3.2.23)
Program Summary 5-8 4
i i
E.lB.V-27 resolution of Generic Safety issue 57 (3.2.24) l 1'
E.lB.V-20 resolution of Generic Safety issue 73 (3.2.26)
E.lB.V-29 equipment classification and vendor interface for reactor trip system components (Generic Safety Issue 75) (3.2.27)
E.lB.V-30 2-week requirement for corrective maintenance (Generic Safety I
Issue 75) (3.2.27) 5 E.lB.V-31 preventive maintenance and surveillance program for reactor trip breakers (Generic Safety Issue 75) (3.2.27)
E.18.V-32 resolution of Generic Safety Issue 76 (3.2.28)
)
r E.lB.V-33 cooldown rate in natural convection cooldown analysis (Generic Safety issue 79) (3.2.29)
E.18.V-34 low-density storage racks in spent fuel pool for most recently di.;-
charged fuel (Generic Safety Issue 82) (3.2.30)
E.lB.V-35 plant-specific design and arrangement for control room heating, l
ventilating, and air conditioning (HVAC) system (Generic Safety i
Issue 83) (3.2.31) i E.lB.V-36 design of emergency filter units (Generic Safety Issue 83) (3.2.31)
E.lB.V-37 design details for control room capacity following a design-basis accident (Generic Safety Issue 83) (3.2.31)
[
E.lB.V-38 design details for control room HVAC systems in the smcie removal mode (Generic Safety issue 83) (3.2.31) l E.lB.V-39 resolution of Generic Safety Issue 87 (3.2.33)
I l
E.lB.V-40 adequacy-of low-tempera.ure overpressure protection design (Generic l
l Safety Issue 94) (3.2.34) l l
E.lB.V-41 adequacy of BWR water level redundancy (Generic Safety issue 101) j (3.2.37)
{
i i
E.lB.V-42 interfacing system design details (Generic Safety Issue 105) l (3.2.39) l
..t E.lB.V-43 inservice testing programs and technical specifications for appro-f l
priate pressure. isolation valves (Generic Safety. Issue 105) l l
(3.2.39) l E.lB.V resolution of Generic Safety Issue 100 (3.2.40)
{
E.lB.V-45 environmental qualification and inservice inspection and testing of
.l large-bore hydraulic snubbers (Generic Safety Issue 113) (3.2.43)
I f
Program Summary.
5-9 i
,. _ -. _ ~. _. _ _. _.. _. _ _... _ _
(
}
i J
r j
t.lB.V-46 use of prestressed concrete containments (Generic Salety Issue 118)
(3.2.44) j j
E.lB.V-47 reliability, operability, and on-line testability of protection system final actuation contacts (Generic Safety Issue 120) (3.2.45) i 4
E.lB.V-4B operator training program ar.d emergency operating procedures related to initiating feed-and-bleed cooling (Generic Safety 4
Issue 122.2) (3.2.50) i i
i 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 I
Issue 12B) (3.2.56)
E.lB.V-51 resolution of Generic Safety issue 130 (3.2.57)
)
E.lB.V-52 resolution of Generic Safety 1stue 132 (3.2.58) i E.lB.V-53 resolution of. Generic Safety Issue 135 (3.2.59)
)
E.lB.V-54 resolution at Generic Safety Issue 142 (3.2.60) i E.lB.V-55 resolution of Generic Safety issue 143 (3.2.61)
E.18.V-56 resolution of Generic Safety Issue 151 (3.2.62)
E.lB.V-57 assessment of safety service water system failure modes and_contri-i butions to core damage frequency and identification of dominant I
accident sequences (Generic Safety Issue 153 (3.2.63)
~
l E.lB.V-58 resolution of Generic Safety issue HF 4.4 (3.2.64)
[
E lB.V-51 resolution of Generic Safety issue HF 5.1-(3.2.65) i E.lB.V-60 resolution of Generic-Safety Issue HF 5.2-(3.2.66)
Chapter 2 - Power Generation Systems l
1 E.2.V-1 safety valve design (3.4)
I E.2 V-2 attachment loads for safety and relief valves (3.4) j E.2.V-3. side stream condensate polisher (4.3)
E.2.V-4 condensate makeup system raw water pretreatment_(6.4)
I
[h30ter 3 - Regr.a.9r Coolant System and Reactor Non-Safety Sv dems h
E.3.V-1 power supplies for power-operated relief valves (3.3)
E.3.V-2 pressurizer heater power source control design.(3.4)-
E.3.V-3 chemica', and-volume contro1' system design (6.2)
Program Summary 5-10
Chapter 4 - Reactor Systems E.4.V-1 reactor p:' essure vessel fatigue design criteria (2.3.2)
E.4.V-2 BWR thermal-hydraulic stability performance during an anticipated transient without scram (4.2)
E.4.V-3 BWR nuclear and thermal-hydraulic design for extended cycle operation (4.2)
E.4.V-4 effect of electric protective assemblies on reactor protection system power supply requirements (5.3)
E.4.V-5 PWR thermal-hydraulic stability and xenon stability characteristics (7.2)
E.4.V-6 PWR fuel design for load-following capability (7.2)
E.4.V-7 60-year service life for control-rod drive mechanisms (8.2)
Chapter 5 - Enaineered Safety Systems E.5.V-1 containment performance criteria for severe accidents (2.1)
E.5 V-2 metal-water reaction and hydrogen generation and control during a severe accident (2.3 and 6.5.1)
E.5.V-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 s, pray system (4.4 and 7.2)
E.5.V-7 suppression pool bypass leakage (4.5 and 7.2)
E.5.V-8 suppression pool temperature-monitoring system (4.6)
E.5.V-9 intersystem LOCA (5.2)
E.5.V-10 oneration of RHR system with reduced reactor coolant' system inven-tory (Generic Letter 87-12) (5.2)
E.5.V-11 shutdown risk (5.2) l E.5.V-12 feed-and-bleed capability (5.4)
E.5.V-13 safety depressurization and vent system (5.4, 5.5, and 6.6.5)
E.5 V-14 use of remote manual valves on essential lines that are riot part of the engineered safety systems (6.2)
E.5.V-15 Type C leak testing (6.2)
L
_ Program Summary-5-11
A E.5.V-16 containment integrated leak rate testing (6.3.1)
E.5.V-17 Type A lt'k testing 16.3.1)
E.5.V-18 Type B t.t 39 of air locks (6.3.2)
E.5.V-19 use of water 'n Type C containment leak rate testing (6.3.3)
E.5.V-20 Type C contaiment valve leak rate testing interval (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 cencrete containment analysis (0.6.1)
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 pt control (8.2 and Appendix B to Chapter 1)
Chapter 6 - Buildino Desian and Arranaement 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 Protectic., 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)
\\
Program Summary 5-12
4 i
t E.6.V-6 compliance with the requirements of Three Mile Island (TMI) Action l
Plan Item II.B.2 (2.3)
E.6.V-7 details of shielding design and shielding computer codes (2.3, 2.4,
{
E.6.V-8 e fe site-specific topography on standard overall site arrange-ment (3.1) l E.6.V-9 flooding protection design requirements (3.3.1)
E.6.V-10 alternative seismic restraint devices (4.2.3)
[
t E.6.V-ll piping and instrument line support design (4.2.4) i E.6.V-12 description of airborne radioactive material sources (4.2.5) h E.6.V-13 potential high-radiation areas, shielding, and measures for minimiz-ing exposure (4.2.8 and 4.2.9) l l
E.6.V-14 review of coatings against SRP Section 6.1.2 (4.2.10 and 4.3.2) i E.6.V-15 use of epoxy-coated reinforcing bars at intake structures (4.2.11) i E.6.V-16 features to ensure H2 concentrt. ions do not exceed detenation
'evels (4.3.2) l i
E.6-V-17 elimination of diagonal rebar in reinforced-concrete containment l
4 (4.3.2)
E.6-V-18 floor size for reactor vessel cavity /drywell (4.3.2)
[
t E.6-V-19 design features that preclude potentially lethal radiation levels (4.3.3)
't E.6.V-20 containment access control (4.3.3 and 4.3.4) f i
E.6.V-21 details of design of BWR reactor building (4.4.2)
E.6.V-22 details of design of PWR auxiliary building (4.4.3) l l
l E.6.V-23 turbine-generator building seismic design loading (4.5.2) i l
E.6.V.24 details of design of BWR turbine-generator building _(4.5.4)
- i i
E.6.V-25 details 'of design of radwaste facility (4.6.3) 5 E.6,V-26 details of emergency onsite power supply facility (4.6.4) i E.6.V-27 details of HVAC design for centrol complex (4.6.5)
E.6.V-28 details of design of technical support center (4.6.6)
I Program Summary 5-13
1 i
t i
l i
j Chaoter 7 - Fuelina and Refuelina Sysign i
i E.7.V-1 quality group classification of components for new and spent fuel i
storage racks (3.2.1) l E.7.V-2 radioiogical consequences of fuel handling accident (3.2.2) i E.7.V-3 protection against tampering during refueling activities (3.2.4) i E.7.V-4 design of the overhead bridge crane (6.1.2) l E.7.V-5 radiological consequinces of fuel cask drop accident (6.5)
E.7.V-6 design of the fuel handling system (7.1.2) i j
E.7.V-7 reactor disassembly and servicing equipment for BWRs (7.5) j Chapter 8 - Plant Coolina Water Systems E.8.V-1 pump nanimum flow line or recirculation line design (3.2)
E.8.V-2 reduction of surveillance testing (3.2) l E.8.V-3 availability of emergency power supply for the fuel pool cooling and cleanup system following a design-basis accident (9) l
[_h_goter 9 - Site Suonort Systems E.9.V-1 fire protection review (3) f E.9.V-2 fire hazard analysis (3.2.2) l E.9.V-3 smoke removal capability (3.3.1)
E.9.V-4 security hardware on fire doors (3.3.1) i E.9 V-5 separation of redundant shutdown equipnent in the containment j
(3.3.1)-
E.9.V-6 control room cable fires (3.4.9) j 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-Il HVAC design for miscellaneous areas (8.2.6) i E.9.V-12 charcoal filters in containment purge _ system (Branch Technical Position CSB 6-4, NUREG-0800) (8.4.2)
Program Summary _
5-14
f i
r i
E.9.V-13 design, equipment, and instrumentation for laboratories (9)
{
E.9.V-14 determination of airborne iodine concentration during an accident (Item III.D.3.3 of NUREG-0737) (9)
Chaoter 10 - Man-Machine Interface System 1 E.10.V-1 software protection (2.3) f i
E.10.V-2 level of automation (2.3)
E.10.V-3 review cf equipment used for displays to the operator (2.3)
E.10.V-4 methods to ensure operator alertness (2.3)
E.10.V-5 additional criteria for developing technology (2.3) i E.10.V-6 independence of verification and validation review teams (3.1.2) l f
E.10.V-7 use of commercial compilers far software used in safety systems t
(3.1.2)
E.10 V-8 dedication of commercial-grade software (3.1.2 and 6.1.2) i i
E.10.V-9 use of commercial-grade equipment (3.1.2)
{
E.10.V-10 complexity of M-MIS (3.1.3) f E.10.V-Il clarification of requirements for analysis and validation testing I
of M-MIS (3.1.3) i E.10.V-12 use of unproven technology (3.2.2)
(
E.10 V-13 operator aids (3.4.5) t E.10.V-14 quantitative reliability criteria (3.5) i i
E.10.V-15 establishment and use of reliability and availability estimates (3.5)
I E.10.V-16 selection of equipment failure modes (3.5.1 and 6.2.7)
E.10.V-17 maintenance frequency (3.5.2)
E.10.V-18 reliability analysis (3.5.4)
E.10.V-19 component reliability of M-MIS (3.5.4) l 5
E.10.V-20 overall reliability of M-MIS (3.5.4) t E.10.V-21 minimum tests for continuous on-line testing (3.6.1)
E.10.V-22 automatic reconfiguration af ter failure detection (3.6.4) f Program. Summary 5-15
. ~..
E.10.V-23 surveillance period of automatic testing features (3.6,0)
E.10. V-:'4 automatic bypass initiation (3.6.10, 3.6.13, and 3.6.14)
E.10.V-25 module software concerns (3.7.4)
E.10.V-26 bypass and test lockouts during on-line repairs (3.7.6) l E.10.V-27 guidance on use of si.aulators and mockups (4.1.3)
E.10.V-28 vulnerability of power supplies for alarm systems (4.3.1)
E.10.V-29 alarn suppression techniques (4.3.3)
E.10.V-30 guidance on criteria to establish priorities (4.3.4)
E.10.V-31 guidance on the maximum number of alarms (4.3.4)
E.10.V-32 guidance on frequency allocation plan (4.6)
E.10.V-33 guidance on interference between communication systems and M-MIS equipment (4.6)
E.10.V-34 unauthorized acces: to equipment in remote shutdown stations (4.9.1)
E.10.V-35 guidance on inadvertent actuation of controls at local control statior.s (4.9.2) 4 E.10.V-36 design of emergency operations facility (4.9.4)
E.lt.v-37 modification of security boundaries during an emergency (4.9.4) o E.10.V-38 data storage methods (4.9.4)
E.10.V-39 compliance of perimeter intrusion alarm system with 10 CFR 73.55(h)
(5.2.1)
E 10.V-40 guidance en data system characteristics (5.2.2)
E.10.V-41 signal transport delay (5.2.5)
E.10-V-42 acceptability of digital-to-analog and analog-to-digital. convertors (5.7)
E.10.V-43 software requirement specification (6.1.2)
E.10.V-44 verification of software (6.1.2)
E.10.V-45 documentation of testing and verification of commercially available software (6.1.2)
E.lG.V-46 acceptance testing of commercially availabic sof tware (6.1.2)
Program Summary 5-16
j I
i E.10.V-47 configuration control of software purchased through software clear-l inghouses (6.1.2)
I j
E.10.V-48 gt.idance on convolution of software structure (6.1.3) j E.10.V-49 behavior of commercial software when assumptions are violated (6.1.3)
E.10.V-50 guidance on memory protection (6.1.3) l l
E.10.v-51 separation of databases for redundant safety-related devices j
(6.1.3) j E.10.V-52 definition of reasonable testing and sufficient degree of confi-i dence (6.1.5) l E.10.V-53 specification of the level of diversity in safety systems (6.1.6, i
6.2.3)
E.10.V-54 specific methcds used to meet the requirement for d!versity (6.1.6)
E.10.V-55 elimination of-EMI (6.2.2) i i
E.10.V-56 compatibility between M-MIS equipment and its external power supply l
systems (6.2.2) j
(
E.10.V-57 signal validation methodology (6.2.2) i E.10.V-58 capacitance-type pressure sensors (6.2.5)
E.10.V-59 minimal acceptance review criteria for isolation device (6.2.6,'
f
?
E.10.V-60 EMI/RFI consider ins for wiring shields (6.2.9) l E.10.V-61 restoration state of control system components after loss of power (6.3.2) i E.10.V-62 setting resolution for control parameters (6.3.3) i E.10.V-63 requirements for signal rt.onstruction (6.3.3) i E.10.V-64 use of interrupts (6.3.3)
[
t E'10.V-65 continuous self-testing of actuation logic (8.3.2) t E.10.V-66 radiation monitor placement, calibration frequency, and emergency
[
power provisions-(10.2.1)--
i E.10.V-67 compliance with Item II.F.1.3 of NUREG-0737 (10.2.1)'
j E.10.Y criterh for airborne radioactivity monitors '(10.2.1) f r
j h
Prograta Summary 5 [
...D
l i
E.10.V-69 14-day maintenance criteria for M-MIS for reactor protection i
system, plant control system, and plant information and monitoring systems (GSI 75) (Appendix B)
E.10.V-70 procedures to assess unscheduled reactor shutdcwns (GSI 75)
(Appendix B)
E.10.V-71 safety implication of instrumentation and control systems (USI A-47, GSI 76) (Appendix B)
E.10.V-72 inclusion of computer specialist on design and review teams (3.1 of f
Appendix D) f E.10.V-73 cstablishment of Q-list and associated equipment list (GSI 75)
(Appendix B)
{
t E.10.V-74 handling of vendor interface (GSI 75) (Appendix B)
E,10.V-75 evaluate neutron monitoring system M-MIS (7.4) f 1
E.10.V reliable operation of reactor trip breaker (GSI 75) (Appendix B)
E.10.V-77 design of BWR water level instrumentation (GSI 101) (Appendix B)
E.10.V-78 operator training and emergency operating procedares concccning feed-and-bleed operations (GSI 122.2) (Appendix B)
[
E.10.V-79 human factors organization (3.1 of Appendix D)
E.10.V-80 acoustical environnients in operating control areas (3.7.6 of j
Appendix D) l E.10.V-81 design reference documents include IEEE P10?3/05 (3.7.6 of Appendix D)
Chapter 11 - Electric Power Systems E.ll.V-1 environmental qualification test criteria for electrical power system (2.2) s E.li.Y-2 safety classification of loads (2.2.1) l E.ll V-3 minimization of Class IE components (2.2.4) l f
E.ll.V-4 instrumentation and controls for electric motors (2.2.5) l l
E.ll.V-5 compliance with NFPA Codes and Standards (2.2.6).
1 E.11.V-6 integrity of electrical cabh penetration seals during a tire I
(2.2.6) l E.11.V-7 integrity of bus duct' penetrations during a fire (2.2.6) r E.11.V-8 review of IEEE-standards not endorsed by regulatory guides (2.2.7) j
- Program Summary 5-18
. -... -.... -,,. ~. _. -. -.
. ~..
\\
E.ll.V-9 review of the actual setpoint criteria used for sizing thermal overloads (2.2.9) i E.ll.Y-10 limitation of total voltage distortion to 3 percent (4.2.4)
E.11.V-Il affects of electrical faults on the coastdown capability of the reactor coolant pumps and reactor internal puinps (4.2.5) 4 E.ll.V-12 use of combustion turbine generator as alternate power source i
during shutdown (5.2.1)
I E.ll.V 13 continuous rating "ersus short-term rating for sizing the combus-l tion turbine generator (5.2.3)
E.ll.V-14 inclusion of the pressurizer heaters in the diesel generator power analysis (5.2.4)-
E.ll.V-15 continuous rating of the diesel generators to include emergency lighting (5.2.4)
E.ll.V-16 capability of the dinsel generators to power safety buses in a protected bus configurit'on (5.2.5)
E.11.V-17 emergency diesel engine startit.g system (5.2.6) l E.11.V 18 emergency diesel engine fuel oil 3torage and' transfer system l
(5.2.6) l l
E.ll.V-19 allowed outage time for load center (6.2) l E.ll.Y-20 impact of loss of ac or de bus on single-failure protection in l
safety-related systems (7.2.1)--
t I
E.ll.V-21 outage time fc* dc safety buses in a BWR plant design (7.2.2) j i
E.ll.V-22 common backup ac power sources for' safety-related uninterruptible power supplies (7.2.4)
E ll.V-23 design of-the continuous ac lightir.g in safety-related areas and access routes outside the main control room (8.2.1 and 8.2.2)
E.ll.V-24 method of integrating the emergency lighting system with' the normal I-lighting in the main control room (8.2.3) j
\\
E.)1.V-25 acceptability of lighting system for closed-circuit television system (8.2.4)
\\
--- QLaoter 12 - Radioactive Waste Processina Systems-1 t
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)
{
r I
Program Summary 5 I
.i
.n c
. _. -,,... - - - - -. -, _ -.., - _ _ _. - ~, - -. - - -,...
E.12.V-3 offsite dose calculation manual (2.2.1)
E.12.V-4 fuel source term parameters for design of radioactive waste pro-cessing systems (2.2.2)
E.12.V-5 estimate of personnel radiation exposure (2.2.4)
E.12.V-6 control, monitoring and sampling of liquid and radioactive waste processing and effluent streams (2.2.9)
E.12.V-7 interface between BWR HVAC systems and GRWP systems (3.3.2)
E.12.V-8 use of HEPA filters downstream of charcoal adsorbers (3.3.3)
E.10.V-9 potentially explosive mixtures of hydrogen and oxygen (3.3.4)
E.12.V-10 piping layout and design and cperating procedures for_ filters and ion excnangers in liquid radioactive waste processing systems (4.2)
E.12.V-Il shipping container _ design (5.5)
Chapter 13 - Main Turbine-Generator System E.13.V-1 60-year design life for major components of the main turbine-generator (2.2)
E.13.V-2 ut.e of seismic experience data base for seismic qualification (3.1.1)
E.13.V-3 performance and safety requirements for mein turbine (3.1.3)
E.13.V-4 turbine maintenance program E.13.V-5 effect of other duty cycles on long-term integrity of the turbine (3.1.4)
E.13.V-6 facture toughness properties of turbine casing material (3.1.4)
F.13.V-7 part-machining inspection of one-piece rotor (3.1.5)
E.13.V-8 need for prototype testing of new or significantly changed designs (3.1.6 and 4.1.1)
E.13.V-9 adequacy of turbine control system (3.3)
E.13.V-10 inservice inspection intervals for main stop and control valves and reheat stop and intercept valves (3.3)
E.13.V-11 seal clearances of gland seal system (3.4)
Program Sunr ary 5-20 i
i i
i j
5.2 Vendor-or Utilitv-Specific Items Pertainina to the Reauirements Document for Passive Plant Desians i
i l
The following is a list of vendor-or utility-specific items i.nat were j
identified in the DSEo on the Requirements Document for passive plant designs.
)
Chapter 1 - Overall Rtquirements j-P.l.V-1 implementation of design characteristics intended to enhance acci-dent resistance (2.2) l P.l.V-2 bounding analysis by standard site design parameters (2.3.1)
P.l.V-3 selection of initiating events-and their frequency categorization l
(2.3.2)
P.l.V-4 acceptance criteria for trarsient-and accident analysis (2.3.2) l P.I.V-5 passive plant anticipated transient without scram response analysis l-(2.3.2)-
i i
P.l.V-6 operator actions 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after acc.ident (2.3.2) i P.l.V-7 use of 72-hour design basis (2.3.2)-
P...V-d technical basis for scvere-accident management program and emergency j
operating procedures guidelines (2.3.4)
}
j P l.V-9 acceptability of analytical codes and methodologies for safety i
l analysis (2,5)
L l
P.l V-10 defe ce-in-depth analysis (2.5 and 3.5) a P.I.V-11 60-year plant life (3.3, 4.8.2, 8.2 and 11.3) l P.l.V-12 operation of PWR with a secured reactor coolant pump (3-5) j P.1.V-13 _ fuel burnup_ requirements (3.6) j P.l.V-14 extended operating-life of control blades and control rod assemblies (3.6) i i
P.l.V-15 safety classification (4.3.1) i j
- P.l V-16 seismic qualification by experience (4.3.2 and 4.8.1)
J P.].V-17 non-seismic building structures-(4.3.2.3 and 4.7.2.10) l P.I.V-18 structural design and construction codes (4.4 and 4.4.1) t-P.1,V-19 elimination of operating-basis earthquake from design (4.4.3, 4.1.3,
[
and Appendix B) e
.l.
1 Program Summary 5-21 i.
4
+ - _.,
....e-e
-y-
,_..- _, ~, -. ~, -
_..-,--s
P.l.V-20 definition of support group (4.4.3)
P l.V-21 use of Appendix N of ASME Code, Section 111 (4.4.3 and 4.7.3)
P.l V-22 analysis of vibratory loads with significant high-frequency input (4.4.3)
P.1.V-23 use of nonlinear analysis to account for gaps between pipes and piping supports (4.4.3)
P.l.V-24 probabilistic approach for changing existing loads and/or loading combinations (4.5.1)
P.I.V-25 recurrence interval for wind loadings (4.5.2.1) 0 P.1.V-26 maxi. mum groend water level (4.5.2.2)
P l.V-27 precipitation for roof design (4.5.2 2)
P.1.V-28 snow loading (4.5.2.2)
P.l.V-29 detailed quantification of soil parameters (4.5.2.3)
P.l.V-30 minimum margin against liquefaction (4.5.2.3)
P.1.V-31 external hazards evaluation (4.5.2.3)
P.l.V-32 number of full-stress cycles (4.5.2.4 and 4.8.1)
P.l.V-33 site-specific 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.l.V-35 design temperature-(4.5,2.7)
P.l.V-36 protection against surface vehicle bombs-(4.5.3)
P.1.V-37 design Against internal-missile genera. ion (4.5.5)
P.l.V-38 design of concrete containment (4.6.1.1)
P.l.V-39 load combinations for Cr'egory I buildings and structures (4.6.1.2)
P.I.V-40 design of Category. I steel structures -(4.6.1.2) t P.1.V-41 combination of pipe rupture loads with seismic loads for seismic Category I structures (4.6.1.3 and 4.6.1.4)
P.l.V-42 combination of LOCA and SSE loads (4.6.1.7)
P.l.V-43 load combinations for safety-related portions of the plant (4.6.2)
P l.V-44' dynamic analysis techniques (4.7.2.3) i Program Summary 5-22
__.____J
1 l
1 I
l P.I.V-45 methodology for generating design spectra or time histories j
(4.7.2.5) l P.l.V-46 structural damping values (4.7.2.6) i P l.V-47 masonry walls in Category I buildings (4.7.2.7)
P l.V-48 use of expansion anchor bolts - compliance with Office of Inspection i
and Enforcement Bulletin 79-02 (4.7.2.8 and 4.7.3) l P.l.V-49 stability of shell-type structures under compression (4.7.2.9)
P.l.V-50 de of ASME Code Cases N-411 and N-420 in sar.e analysis (4.7.3)
P.l.V-51 use of ASME Code Case N-411 (4.7.3)
P.l.V-52 construction of cc:e support structures (4.7.3)
P.l.V-53 design fatigue curves (4.7.3)
P.I.V-54 use of zine to reduce radiation fields (5.2.7)
P.l.V-55 grinding controls for PWRs (5.3.1.1)
P.I.V-56 use of Alloy 600 (5.3.1.3)
P.l.V-57 effect of fabrication processes on intergranular stress corrosion cracking (5.3.1.8) i P.l.V-58 selection of seal, gaskets, and protective coatings (5.3.5)
P.I.V-59 aging of cable insulations and other electrical materials (5.3.6)
P.l.V-60 use of hydrogen water chemistry for the advanced BWR_ design (5.5.2)
P.!.V-61 -plant-specific reliability assurance program (6.5)
P.I.V-62 inspection of construction activities (7 and 11.13)
P.l.Y-63 installed operating-phase security system (7)
P.1,V-64 reliability of modular construction (7)
P.l.V-65 inspection and verification of-security locks robotica11y (8.3) _
P.I.V-66 compliance of FDA/DC applications with_ Commission's regulations.and-guidance (10)
P l.V-67 issue resolution for FDA/DC reviews (10)
P.l.V-68 inspections, tests, analyses, and accej::ance criteria (10)
P.l.V-69 implementation of simplification objective (11.4)
Prograt.1 Summary _
5-23
}
l
)
)
J
}
P.I.V-70 implementation of standardization objective (11.5) j 1
h 1
P.I.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.I.V-73 check valve testing methods (12.2.2) i
}
P.l.V-74 full-flow testing of check valves (12.2.2) l 1
P.l.V-75 provisions to test hydraulically and pneumatically operated valves under design-basis differential pressure and flow (12.2.2)
}
P.l.V-76 qualification testing of active and non-active motor-operated valves (MOVs) (12.2.2) l P.l.V-77 technical concerns regarding H0Vs (12.2.2) r P.l.V-78 leak rate testing for individual containment isolation valve (12.2.2) j 1
P.I.V-79 frequency and extent of disassembly and inspection of safety-related pumps (12.2.3) hpoendix A to Chapler 1 - PRA Key Assumntions and Groundrules -
l t
P lA.V-1 long-term decay heat removal in the PRA (1.6)
?
P.lA.V-2 justificativo of mission items and success criteria (2.10) l P.lA.V-3 reliability data (2.11)
P.lA.V-4 review of core-damage-sequence binning (4'.1) f P.lA.V-5 review of actual groupings of the accident sequences into plant i
damage states (4.2)
P.lA.V-6 review of the evaluation of containment leakage paths (4.3)
P.lA.V-7 computer codes for in-plant sequence assessment (4.4)
P.lA.V-3
-verification that the referenc9 site parameters identified in Annex l
B are consi: tant with revised 10 CFR Part 100 (5.2) i P.lA.V-9 differences in computer codes used for calculating offsite conse-
-l quences (5.2)
[
P lA.V-10 source terms for representative accident sequences bounded by the i
physically based sour.ce term'(6.3) e i
l-Program Summary 5-24 j
i
__a._._._._,--_._..___.
-. -. ~.
Apper. dix B to Chapter 1 - LiegnMng_and Reaulat. pry Reauirements asi Guidance P.lB.V-1 r$mpliance of FDA/DC applications with Commission's i 31ations and gt ince (1.2)
P.lB.V-2 issue resolution for FDA/DC reviews (1.2)
P.lB.V-3 dynamic seismic analysis of main steam piping and condenser (2.3.1 and Item II.E of Annex A)
P.lB.V-4 main steamline piping classification (2.3.1 and item II.E of Annex A)
P.lB.V-5 seismic analysis of turbine building (2.3.1 and Item II.E of Annex A)
P.lB.V-6 plateout con:.. aerations for main stean piping and valves (2.3.1 and Itt e III.F of Annex A)
P.lB.V-7 reactor vessel level instrumentation system (2.4.1)-
P.lB.V-8 dedicated containment vent penetration (2.5.4 and Item I.K of Annex A)
P.lB.V-9 fission product leakage control (2.5.2 and item Ill.F of Annex A) l P.lB.V-10 fission product cleanup system (2.5.2 and Item III.F of Annex A)
P lB.V-Il automatic emergency core cooling system switch to recirculation (Generic Safety Issue 24) (3.2.11)
P.lB.V-12 bolting degradation or failure (Generic Safety issue 29) (3.2.12)
P.lB.V-13 detached therm *1 sleeves (Generic Safety issue 73) (3.2.15)
P.lB.V-14 failure of high-pressure coolant injection steamline without isola-tion (Generic Safety Issue 87)(3.2.20)
P.lB.V-15 adequate leak test requirements for pressure isolation valves in inservice testing programs and technical specifications-(3.2.22)
P.lB.V-16 identification and design of interfacing systems (3.2.22)
P.13.V-17 implementation of procedures for electrical power reliability (3.2.27)
P.lB.V-18 leakane through electrical isolators in instrumentation circuits t
(Generic _ Safety Issue 142) (3.2.29)
P.lB.V availability of chilled water' systems and room cooling (Generic Safety Issue 143) (3.2.30)
P.lB.V-20 reliability of recirculation pump trip during at anticipated transient without seram (Generic Safety Issue.151) (3.2.31).
Program Summarf 5-25
1
.I 4
P lB.V-21 loss of emergency service water system (Generic Safety Issue IS3) (3.2.32)
J P.lB.V-22 spectral peak shifting (3.3.1) i j
(h mter 2 - Power Generation Systems i
1
]
P.2.V-1 turbine bypass system flow capacity (3.2)
P.2.V-2 attachment loads for PWR safety and relief valves (3.4) l P.2.V-3 design adequacy of side stream polisher (4.3) i Chapter 3 - Reactor Coolant System and Reactor Non-Safety Auxiliary Systems i
)
P.3.V-1 vent and drain design (2.1.2) i P.3.V-2 reactor coolant interface systems design (2.1.7, 3.1, and 5.1) i P.3.V-3 specific performance requirements and acceptance criteria for active non-safety auxiliary systems (2.2, 6, and 8) j P.3.V-4 safety analyses of the abnormal conditions associated witt loss of i
a feed pump or load rejection (3.2)
P 3.V-5 acceptability of operation with reactor coolant pump (s) out of 3
service (3.2) i P.3.V-6 low-temperature overpressure protection system design details (3.3)
P.3.V-7 design analyses to confirm the capability and reliability of the
}
passive decay heat removal system (3.4) i 1
P.3.V-6 specific design requirements for reactor vessel level instrumenta-
)
i tion system (3.5)
P.3.V-9 design details of the man-machine interface system for steam generator water level control (4.5)
}
P.3 V-10 design details to mitigate excessive leakage of main steam isola-l tion valves (5.4) 1 P.3.V-11 design details for automatic reactor vessel overfill protection (5.5) j i
l P.3.V-12 design details for PWR auxiliary systems, ii.cluding chemical and volume control system, against criterion in Standard Review Plan Section 9.3.4 (6) i I
P.3.V-13 design details for reactor shutdown cooling-pump seals (9)
P.3.V-14 adequate vendor assessment of-shutdown and low-power-operation risk (9)
Program Summary 5-26
]
'f
. - - -. - - - -. ~ _ _ -. _ _ -. - -.. - - -. - - -. -. - - - - - _
1 I
i Chapter 4 - Reactor Systems P.4.V-1 boration of water for fuel handling and storage (2.2.8)
P.4.V-2 fuel assembly reconstitution (2.3.3) i P.4.V-3 condensation carryunder limitation (3.2) t P.4.V-4 experimental data for divided-chimney design (3.2 and 4.2.1)
P.4.V-5 BWR stability (2.2.4 and 4.2.1)
P.4.V-6 decay ratio limits and analysis methods and procedures, (4.2.1)
{
i P.4.V-7 LOCA analysis methodology (4.2.1 and 7.2.1) i P.4.V-8 load-following and maneuvering capability (4.2.1 and 7.2.1)
P.4.V-9 methodology to achieve BWR stability (4.2.1)
P.4.V-10 fuel burnup requirements (4.2.2 and 7.2.2) i P.4.V-Il two-cycle fuel channel lifetime (4.2.4) i P.4.V-12 control rod assembly lifetime (4.2.6 and 7.2.3)
[
t P.4.V-13 control rod scram time (5.2)
P.4.V-14 control rod assembly malfunctions in BWR accident analyses (5.2)
-I P.4.V-15 protection of scram pilot solenoid valves (5.3) t P 4.V-16 thermal shield removal (6.3)
P.4.V-17 reactor pressure vessel level instrumentation (6.3) l P.4.V-18 negative moderator temperature coefficient limit (7.3) f i
P.4.V-19 fuel rod bow penalties (7.3)
P.4.V-20 control rod drive mechanism lifetime (8.2) i Chapter 5 - Enaineered Safety Systems f
P.5.V-1 decontamination factor for containment system (2.1.7)
I P.5.V-2 challenge from inadvertent opening of the DPS (2.2)
[
P.5.V-3 safety-grade provisions for the fire protection system (2.3) l P.5.V-4 LOCA calculations justifying renaval of core spray system (4.1) j P.S.V-5 manual standby liquid control system iritiation (4.5)
L Program Summary 5-27 l
}
e P.5.V-6 system design to minimize condensation water hammer (4.2)
P.5.V-7 in-containment refueling water storage tank boiling suppression (5.3)
P.5.V-8 justification frr use of remote manual valve for containment isola-tion (6.2)
P.S.V-9 evaluation of the ignition system for combustible gas control (6.6)
P.5.V-10 reliability of power supplies for severe-accident equipment (6.7)
P.S.V-Il detailed discussions regarding design-basis-accident events (9)
Chapter 6 - Buildir.a Desion and Arranaement P.6.V-1 method for inspectine structural degradation (2.1.1)
P.6.V-2 evaluation of the engineering backfill (2.1.1) 1 P.6.V-3 use of American National Standards Institute (ANSI) 10.4-1987 (2.1.1)
P.6.V-4 structural modules to be used ir. construction (2.2)
P.6.V-5 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 contro111r] and storing toxic materials (2.3.7)
P.6.V-8 shielding design requirements, shielding computer codes, and radio-active 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-ll 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)
F.6.V-15 design features that preclude potentially lethal radiation levels (4.2.2}-
P.6.V-16 containment design details for aerosol.and radioactive qases (4.2.3)
Program Summary 3-28
l P.6.V-17 use of ANSI /ASME N0G-1, 1983 (4.2.3)
P.6.V-18 design of the control couplex (4.5.4)
Chapter 7 - fuelina and Refuelina Sysis;31 l
P.7.V-1 design of the overhead bridge crane (2.3.2) i P.7.V-2 high-radiation areas (2.3.7) l P.7.V-3 reactor disassembly and servicing equipment for BWRs (3.1.2)
Chapter 8 - P1 nt Coolina Water Systems 0
P.8.V-1 design requirements for the chilled water system (8)
P 8.V-2 time-delay allowance for fuel pool cooling capability (9)
Chapter 9 - Site Suogort Systems P.9.V-1 separation of redundant shutdown equipment in i.he containment (3.3.1)
P 9.V-2 underfloor or ceiling control room cable fires (3.4)
P.9.V-3 security area devitalized during unit shutdown (5.1)
P 9.V-4 security for components that manipulate vital isolation valves (5.?.1)
P.9.V-5 non-safety-related auxiliary systems within the protected area (5.2.1)
-P.9.V-6 sabotage vulnerability analysis (5.2.2)
P.9.V-7 charcoal filters in air filtration systems (8.2.4, 8.2.5, 8.3.3, 8.4.1, and 8.4.2)
P.9.V-8 safety classification of fuel facility ventilation supply subsystem-(8.2.4)
P.9.V-9 safety classification of PWR auxiliary building ventilation supply subsystem (8.4.2)
.Ghaoter 10 - Man-Machine Interface Systems (M-MIS)
P.10.V-1 acceptable' interpretations of requirements (1)
P.10.V-2 software protection (2.3)
P.10.V-3
'evel of automation (k.1)
.P.10.V-4 defense-in-depth and eiurn 'ty_ analysis (2.3 and 4.5)
Program Summary 5-29
I f
P.10.V-5 review of equipment used for displays to the operator (2.3)
P.10.V-6 methods to ensure operator alertness (2.3)
P.10.V-7 additional criteria for developing technology (2.3)
P.10.V-8 independence of verification and validation reviw teams (3.1.2)
P.10.V-9 use of commercial-grade equipment (3.1.2) l P.10.V-10 complexity of M-MIS (3.1.3)
P.10.V-Il use of unproven technology (3.2.2).
f P.10.V-12 quantitative reliability criteria-(3.5)
P.10.V-13 selection of equipment failure modes (3.5.1 and 6.2.7)
P.10.V-14 maintenance frequency (3.5.2)
P.10 V-15 reliability analysis (3.5.4) 4 P.10.V-16 automatic reconfiguration after failure detection (3.6.4)
P.10.V-17 surveillance period of automatic testing features (3.6.8)
P.10.V-18 automatic bypass initiation (3.6.10, 3.6.13, and 3.6.14) i i
i P.10.V-19 module sof tware concerns (3.7.4)
P.10.V-20 bypass and test lockouts during on-line repairs (3.7.6)
[
?
P.10.V-21 main control room staffing (4.2)
[
P.10.V-22 alarm suppression techniques (4.3.3)
P.10.V-23 hse of " dial-up" telephone-type portable radios for security purposes (4.6.3) i I
P.10.V-24 unauthorized access to equipment in remote-suutdown stations i
(4.9.1) i P.10.V-25 computer room within the main control room security boundary (4.9.1)
P.10.V-26 design of emergency operations facilityL(4.9.4)_
P.10.V-27 modification of-security boundaries during an emergency (4.9.4) t P.10.V-28 data storage methods (4.9.4)
P.10.V-29L compliance of perimeter intrusion alarm system with 10 CFR 73..,o(h)
'(5.2.1 and 5.2.5)_.
I.
j i
-Program Summary 5-301 j
i m
i i
=
P.10.V-30 signal transport delay (5.2.5) l P.10.V-31 analog-to-digital and digital-to-analog converters (5.7)
P.10.V-32 software requirement specification (6.1.2 and 6.1.6) l t
P.10.V-33 verification of software (6.1.2) t P.10.V-34 documentation of testing and verification of commercially available software (6.1.2)
P.10.V-35 configuration control of software purchased through software
[
clearinghouses (6.1.2)
P.10.V-36 specific methods used to meet the requirement for diversity (6.1.6) i P.10.V-37 elimination of electromagnetic interference (6.2.2) f P.10.V-38 signal validation methodology (6.2.2) f l
P.10.V-39 restoration state of control system components efter loss of power (6.3.2) i P.10.V-40 setting resolution for control parameters (5.3.3) i P.10.V-41 neutron monitoring M-HIS (7.4)
{
P.10.V-42 selection of variables for automatic actuation (8.2.3)
P.10.V-43 radiation monitor placement, calibration freouency, and emergency
[
power provisions (10.2.1)
[
l P.10.V-44 compiiance-with Item II.F.1.3 of NUREG-0737 (10.2.1)_
j P.10.V-45 criteria for airborne reactivity monitors (10.2.1) l l
P.10.V-46 operating philosophy (2 of Appendix B) i P.10.V-47 use of mockups, prototypes, and simulators (2 of Appendix B) i Charter 11 - Electric Power Systems P.11.V-1 reliance on the non-safety electrical systems beyond 72-hour
+
period (2.2.2) j P.ll V-2 reliance on the ncn-safety electrical systems to achievt cold.
}
shutdown (2.2.2)
P.ll.V-3 applicable regulations and regulatory guidances that are not I
addressed in Chapter 11. (2.2.3)
{
P.llV-4 minimization of Class lE components (2.2.4) j I
i I
Program Summary 5-31 t
. ~,.
P.ll.V-5 use of revisions of IEEE standards not endorsed by the staff (2.2.7)
P.ll.V-6 operating conditions of all plant loads for all re' levant grid conditions and the design of the bus voltage protection schemes (3.2.4)
P.ll.V-7 design of the standby power source starting system (5.2.6)
P.ll.V-8 design of the standby power source futl oil storage and transfer system (5.2.6)
P.11.V-9 design of the electrical separation of de and vital ac power supply systems (7.2.5)
P.ll.V-10 integration of the exterior lighting system with the closed-circuit television system (8.2.4)
Chapter 12 - Radioactive Waste Processina Systems P.12.V-1 requirements for radioactive waste processing systems and effluent paths (2.2.8)
P.12.V-2 gaseous radioactive waste processing system hydrogen control design (3.3.4)
P.12.V-3 decign of dry waste shipping containers (5.5)
Chapter 13 - Main Turbine-Generator Systems 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)
Program Summary 5 3z
6 i
6 CONCLHSION Reouirenents Document for Evolutionary Plan _L0esians Subiect M the resolution of the identified outstanding r.olicy issues an('
vendo-and utility-specific items discussed in the SER (Volume 2 of this the staff concludes saat tha requirements established in the Require-repoc o, ments Document for evoluti,cary plant designs (Volume 11) do not conflict with r
- p..
urant regulatory guiddines and are acceptable.
However, by themselves they do nat provide sufficient information for the sta to determine :f the plant design will be adequate.
Therefore, applicante el rencing the 6:quirements Decument will be required to demonstrate compi ce with the additional d.;
- u 'ance nrovided 5
- the Standard Review plan (NLREG-0800), or provide justi-
~~
.ior for altermive means of implementing the associated regulatory "ts.
r
,f sf' rquirements memarandc:n (SRM) of August 24, 1989, the Commission Af Y
u, an
<d
.ie staff to provide an analysis detailing where the staff proposes
~
de u. rom current regulations or where the staff is substantially sup,.le-j or revising iniecpretive guidance applied to currently 1icensed light t.
'4ctors (LWRs), lhe staff considers these to be policy issues.
/wa -
Ei to Chapter 1 of Volme 2 of this report gives the staff's regula-7 3
tcry R lys:s of those issaes identified for these designs.
These issues have
/
4 5etn addressed !n SECY-90-016 and SECY-91-078, and in draf t Commission papers, ys
" Issues Pertaining to Cvolutionary and Passive Light Water Reactors and Their Relationship to Current %1ulatory Requirements," and " Design Certification g
e and Licer/ing Issues Pt aing to Passive and Evolutionary Advanced Light j
Water Reactor Designs,"
t were issued ca February 27 and July 6, 1992, e
respectively.
In its SRMs dated June 26, 1990, and April 15, 1991, tE Commission provided b
its decisions on SECY-90-016 and SECY-91-078 as they
" j to wolutionary designs.
The 'ecruary 27 a..d July 6, 1992, draft Commission papers have been forwarded to tn Advisory Committee on Reactor Safeguards.
The staff will include its views in the fi al papers and document its final positions before seeking Commission approval. When the staff finalizes these Commission papers, the Commission will complete its review of the basis for the approach that Cie
'inff is proposing for those issues and, acco-dingly, may at some future point f
i t. the review determine that such issues involve policy questions that the Commission may wish to consider.
The apprcaches to resolving these issues have not been reviewed by the Commission, and, therefore, do not represent-agency positions, A
Tharafore, the staff concludes that EPRI's ALWR Utility Requirement-Document
$j for evolutionary plant designs (Volume 11) specifies requirements t' at, d,.
subject to the resolution of the identified outstanding policy issues and vendor-and utility-specFic items, if properly translated into a design and g
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 Program Summary 6-1
f l
will have all the attributes required to ensure that there is no undue rir.k to the health and safety of the public or to the environment.
In addition to g
complying with existing regulations, such a facility would also be consistent with the Commission's policies on severe-accident protection.
Reauirements Docume,it for Passive Plant Desians Subject to the resolutiot, of the identified outstanding issues and vendor-and utility-specific items listed in Sactions 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 Requirerents Document for passive plant designs (Volume Ill) do not conflict with current regulatory guidelines and are acceptable.
However, by 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 yidance 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 deta"ing where the staff proposes departure from current regulations or where the stafi is substantially supplementing or revising interpretive guidance applied to currently licensed LWRs.
The staff considers these to be policy issues.
Appendix B to the DSER on Chapter 1 of the Requirements Document for passive plant designs gives that analysis.
The staff fon.rded these issues to the Commission in draft Commission papers dated February U and July 6, 1992. When the staff finalizes these Commission oapers, the Commission will complete its review of the basis for the approach that the j$
staff is proposing for those issues and, accordingly, rc.ay at some future point in the review cetermine 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.
In addition, certain technical issues still have to be resolved befera the staff can complete its review.
The following conclusions are based on the staff's review as documented in the April 1992 uraft SER (Volume ? 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 deshns.
Therefore, on the basis of its review to date, the staff concludes that EP.a's ALWR Utility Requirements Document for passive plant designs (Volume 111) specifies requirements that, subject to the resolution of the identified outstanding issues and vendor-and utility-specific items, if properly translated ato a design and constructed and operated in accordance with the NRC regulations in fcrce 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 misting regu tions, such a facility would also be consistent with the Commission's policies on severe-accident protection.
Program Summary 6-2
APPENDIX A LHRON0 LOGY OF CORRESPONDENCE This appendix contains a chronological listing of routine licensing correspon-dence 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 Letter fro:n J. C. Mark, Advisory Committee on React + r Safeguards (ACRS), to NRC submitting-suggestions regarding potential safety improvements for incorporation into new designs for nuclear povar plants.
NRC should give appropriate priority and re rces to developing safety tequirements ior future LWu Septernber 15, 1982 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, 1932, meeting.
Participation required in policy committee and program group.
October 20, 1982 Letter from H R. Denton, NRC, to EPRI, responding to September 15, 1982, letter requesting NRC particination in EPRI's Standardized LWR Design Program.
NRC involvement in program to prove beneficial to both NRC and EPRI. Meeting requested.
October 27, 1982 Letter from C. O. Thomas, NRC, to EPRI, summarizing October 21, 1982, meeting witn EPRI regarding LWR standardization progam. Development of LWR plant b.:,eline-designs and approach th s. could be ':ed by vendors in obtaining certification of plant design emphasi:.ed in EPRI program.
December 20, 1982 latter 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.
Related information enclosed.
January 21, 1983 Letter from D. H. Moran, NRC, to EPRI, summarizing January 12, 1983, meeting with EPRI in Bethesda, Maryland, regarding review of 1icensing' issues to be con-sidered in LWR standardization program.
January 27, 1983 Letter from R. E. Nickell, EPRI, to NRC, confirming high-priority safety -and licensing issues selected at -meeting Program Summary
.. A-1
on January 12, 1983, with emphasis on revision of Appendix R to 10 CFR Part 50, decay heat removal (DHR),
and high-strength bolting.
Possible items for discussion at February 9, 1983, meeting suggested.
February 21, 1983 Letter from EPRI to NRC, requesting meeting between NRC Policy Committee and Utility Steering Committee for EPRI's Standardized LWR Design Program.
February 28, 1983 Letter from D. H. Moran, NRC, to EPRI, summarizing February 14, 1983, meeting with EPRI in Bethesda, Maryland, regarding current status and progress te date of LWR standardization program.
March 10, 1983 Letter from H. Denton, NRC, to EPRI, responding to February 21, 1983, request for meeting between NRC i011cy Committee and Utility Steering Committee for EPRI's Standardizc1 LWR Design Program. Meeting to be held on April 6,1983, ir Bethesda, Maryland.
March 10, 1983 Letter from D. H. Moran, NRC, to EPRI, summarizing February 22, 1983, meeting with EPRI and Sol Levy Associates in Pale Alto, California, regarding ' nal priority sequence anided when complete.
Information would 1
support ACRS Oc tober 6,1987, meeting.
l_
September 24, 1987 Letter from L. S. Rubenstein, NRC, to 'EPRI, forwarding DSER on Chapter 1 of ALWR Utility Requirements Document.
i Findings generally favorable to ALWR program.
1 October 6, 1987 Transcript of ACRS Standardization of Nuclear Facilities Subcommittee meeting on October 6, 1987, in Washington, j
D.C.
Pp 1-205.
Supporting documentation-enclosed.
October 12, 1987 Letter fron J. C. Devine, EPRI, to NRC, commenting on proposed new SRP Section 3.6.3 regarding leak-before-break evaluation _ procedures.
Proposed approach would reverely limit application of leak-brfore-bre.k technology for future plants and not i
achit '.e intent of General Design Criterion 4 of Appendix A to-10 CFR Part 50.
l November 13, 1987 Letter from P. H. Leech, NRC, to EPRI, requesting additional information on ALWR Utility Requirements a
Program Summary A-9
i i
2 Oc;ument, Chapters 3 and 4.
Schedule fer review of Chapters 3 and 4 based on receipt of-response by i
December 31, 1987.
l November 24, 1987 letter from A. E. Scherer, Combustion En;;ineering, Inc.,
to NRC, forwarding proposed Advanced Reactor Severe Accident.Progra.n resolutions for four remaining NRC/ Industry Degraded. Core Rulemaking (IDCOR) severe-accident issues.
Six valid DOE /IDCOR resolutions to be adopted in developing System 80+ design.
Concurrence requested.
i December 1, 1987 Letter from L. S..Rubenstein,.NRC, to EPRI, forwarding i
list of significant technical issues that might arise during revia. of evolutionary standard ALWR plants.
i l
Comments requested.
~
December 8, 1987 Letter from E. E. Kintner, EPRI, to NRC, forwarding Revision 0 of ALWR Utility Requiremente Document, Chap-3 j
ter 5 for review.
i December 11, 1987 Letter from P. H. Leech, NRC, to EPRI, requesting additional information en Chapters a and 4 of ALWR Utility Requirements Document.
Notification requested in case responce delayed more than 2 to 3 weeks beyond December 31, 1987.
January 19, 1988 Letter from L. S. Rubenstein, NRC, to EPRI, forwarding i
status report on technical reviews of standardized plant I
designs for January 1988.
Problem neeting listed input dates requested within I week of receipt of letter.
I January 25, 1988 Letter from D. Crutchfield, NRC, to EPRI, forwarding information on ALWR performance goals.
Information assembled after discussion with knowledgeable NRC staff members and during January 19,1988..aeeting with EPRI representatives.
I January 25, 1988 Letter f.om E. E. Kintner, EPRI, to NRC, forwarding l
l 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 fron L. S. Rubenstein, NRC, to EPRI, requesting cdditional information regarding design goals addressing severe-accident releases.
Response requested withi., 30 days of letter date.'
February 5, 1988
_etter from Director, Office of Nuclear Reactor i
4 Regulation, to EPRI' forwarding revised DSER on Chapter 1 j
of ALWR Utility Requirements Document.
February 18, 1988 Letter from L. S. Rcbenstein, NRC, to EPRI, forwarding DSER cn Chapter 2 ALWR Utility Requirements Document.
v 1
Progra Summary A-10 l
-}
I Utility requirements in Chapter 2 are in general l
agreement with NRC guidelines and regulatory requirements i
for power generation system involved, j
February 18, 1988 Letter from L. S. Rubenstain, NRC, to Committee to Review l
Generic Requirements (CRGR), forwarding DSER on Chapter 2 i
of ALWR Utility Requirements Document and informing CRGR members and staff of progress in reviewing Requirements i
Document.
l t
March 18, 1988 Letter from P. H. Leech, ND.C, to EPRI, requesting additional information on Cnapter 5 of ALWR Utility Pequirements Decument.
Current schedule of revicw based on receipt of response by April 29, 1988.
Noti?ication expected it' delay anticipated.
March 25, 1988 Letter from L. S. c.ubenstein, NRC, to EPRI, forwarding i
safety avaluation of recommended modifications of Regulatory Guide 1.76, ' Design Basis Tornado for-Nuclear e
Power Plants." NRC interim position const itutes con-i servative redu> tion of design-basis winds for use by EPRI until revision available.
March 28, 1988
'. er from E. E. Kintner, EPRI, to NRC, forwarding responses to Questions 36 and-40 to resolve NRC comments i
regarding ALWR low temperature overpressure protection-requirements.
April 4, 1988 Letter from P. H. Leech, NRC, to EPRI,-requesting additional informatinn m, Chapter 5 of ALWR Utility i
Requirements Documeat.
Response to questions requested by April 29, 1988.
l i
April 6, 1988 Letter from E. E. Kintner, EPRI,- to NRC, providing i
initial response to January 27, 1988, questions an ALWR r
Utility Requiremer.ts Document regarding impleaentatien of
.i public safety criteria.
Probabilistic risk-assessment-(PRA) key assumptions and groundrules-to be submitted in l
September 1988 to provida more detail, i
May 13, 1985
-Letter from_L. S.'Rubenstein, NRC, to EPRI, forwarding.
DSER on Chapter 3 of ALWR Utility Requirements Document.
Docuruent in general agreement with NRC guidelines based-i
.on January 25 and March 25,-193d,l letters regarding
}
reactor coolant system and nonsafety auxiliary systems.
I
- May'31, 1988 Lettbr from Director, Office of Nuclear Reactor-1 l
Regulatlon,--forwarding' 05ER on Chapter 3 of ALWR -Utility i
j Requirements Document.
l June 8, 1988 Letter from Director, Office of Nuclear Reactor Regulation, Director, forwarding 05ER on Chapter 4' of ALWR Utility Requirements Document.
.l i
Program Summary-A-11' l
l
~ - - ----
lcL-
.n
_. ~
l l
(
June 10, 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, l
1988, letters.
June 20, 1988 Letter from L. S. Rubenstein, NRC, to ACRS, forwarding i
DSERs on Chapters 3 and 4 of ALWR Utility Requirements Document.
Effort to continue until final SER issued in l
1991.
1 j
June 30, 1988 Report by Fauske and Associates Inc., " Technical Support for Hydrogen Control Requirement for EPRI Advanced LWR Requirements Document-Task 8.3.5.A," Advanced Recctor j
Severe Accident Program.
1 July 31, 1988 Letter from R. Stiger, International Technology Corp., to NRC, " Technical-Basis for EPRI Advanced LWR Requirements Documert Assumption on Delayed Fission Preduct Release l
(Task 8.3.5.2)," Advanced Reactor Severe Accident Pro-
[
gram.
t l
August 16, 1988 Lette from C. E. kintner, EPRI, to NRC, forwarding responses to March 18 and April 4,1988, requests for additional information on Chapter 5 of ALWR Utility l
Requirements Document on station blackout.
l September 15, 1988 Letter from E. f. Kintner, EPRI, to NPC, forwtrding j
response to NRC Comments 430.1, 450.2, 430.3, and 430.4, regarding ALWR Utility Requirements Document and two j
reports referenced in responses to NRC Comments 480.5 and 450.1.
j September 23, 1988 Letter from W. O. Long, NRC, to EPRI, regarding low--
i temperature overpressure protection for ALWRs.
i September 23, 1588 Letter from C. Y. Cheng, NRC, to=EPRI, '.larifying j
position on reacto vassel surveillance program and low-tempe.rature'overpres w e protection in Chapters 3 and 4 4
I of ALWR Utility Requirements Document.
j_
Octob;r 26, 1988 Letter fron W. - O. Long'.- NRC, to EPRI,' clarifying-two j
statements iniMay 13,.1988, DSER on Chapter 3 of ALWR L
Utility Requirements Document Noncerning chemical and volume control system.
l November 1, 1988 Letter from W.'0. Long,-NRC, to:EPRI, summarizing:0ctober
{
-26, 1988, meeting with EPRI in Rockville,= Maryland, regarding NRC recomendations Lforaiesign of ALWR5 electrical'systemc --Meeting hande.at enclosed.
3 L
I November 17, 1988 Letter from W. O. Long, NRC, to EPRI, advising that open Issue re5arding suppression-pool loads in Section 3.0.D.2 of Chapter 1 of ALWR'Requirecents Document needed further i
(
' Program Summary
' A-12'
clarification.
Section not clear on whether proposed leak-before-break proposal consistent with General Design Criterion 4 of Appendix A to 10 CFR Part 50.
November 18, 1988 Letter from E. E. Kintner, EPRI, to NRC, forwarding Revision 0 of ALWR Uti'.ity Requirements Documer.t Chapter 6, for review.
November 22, 1988 Letter from D. Crutchfield, NRC, to EPRI, forwarding information regarding scope of safety analysis report for future standardized design applications and NRC revieu of applications.
December 23, 1988 Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Utility Requirements Document, Chapter 12.
December 30, 1988 Letter from E.E. Kintner, EPRI, to NRC, forwarding f.evision 0 of ALWR Utility Requirement Document, Chapter 8.
January 9,-1989 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-terproductive.
January li, 1989 Letter from E. E. Kit.tner, EPRI, to NRC, forwarding ALWR 7
Requiremen+s Document, Chapter 9, for use in chapter eview.
January 31, 1989 L4 tter front O. G. Harrison, Idaho National Engineering Laberatory, to NRC, forwarding
- Interim External Events
' integration for EPRI Advanced LWR Requirements Document WBS 4.3.3."
February 2, 1989 Letter from 9. 0. Long, NRC, to EPRI, discussing proposed resohtions of Generic Safety Issues I.F.1,-" Expand QA List," and IIJ.5 " Classification of Instrumentation and Electrical-Equipment," ar.d suggesting that EPRI reaffirm
- commitment in its December 3,1985, letter in order to
,J resolve issues.
l
(
February'6, 1989 Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Utility Requirementt Document: Chapter 13.
Chapter i
incorporated lessons learned from tvalnation of LWR technology as applied to generation of electricity.
February 23, 1989 Letter from H O.~Long, NRC, to EPRI, requesting additional informatica on AiWR Ottitty Requiccments -
Occument, Chapter 6.
Infcmation should be cravided on level of detail consisttnt erith Regulator:y Guide 1.70 for safety analysis report.
Program Sunmry A-13 l
A f
)
i February 28, 1989 Letter from E. E. Kintner, EPRI,-to NRC, forwarding ALWR i
Utility Rcouirements Document, Chapter 7.
March 21, 1989 Letter from W. O. Long, NRC, to EPRI, summarizing meeting on March 15, 1989, with EPRI and contractors in Palo 4
Alto, California, regarding ALWR issues.
List of attendees and handouts enclosed.
I q
March 22, 1989 Letter from W. O. Long, NRC, to EPRI, requesting additional information on ALWR Utility Requirements Document, Chapters 8, 9, 12, and 13.
Response requested 1
within 60 days.
2 March 22, 1989 Letter from W. O. Long, NRC, to EPRI, requesting review l
of Chapter 7 of ALWR Utility Requirements Document.
Schedule for review given.
1 March 30, 1989 Letter from E..E. Kidner, 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, l
1988.
March 30, 1989 Letter from E. E. Kin *ner, EPRI, to NRC, submitting l
additional inforntie
~.'icient to resolve Generic
}
Safety Issue (GSI)
- .t.6.1, "In Situ Testing of Valves,"
ALWR Utility Require.ments Document addressed subissues of i
GSI II.E.6.1..Cor.currence on resolution of issue j
requested.
I March 30, 1989 Letter f,om L, E. Kintner, EPRI,-to NRC, responding to i
March 25, 19Ed, letter regarding optimization issue on i
tornado design in Chapter 1 of ALWR Utility Requirements i
Document. Agreed to use tornado design criteria in Table j
2 and Figure 2 of safety evaluation for ALWR.
i April 3, 1989 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-meat.
April 5, 1989 Letter from E. E. Kintner, EPRI, to NRC, regarding GS!s I.F.1 and II.F.5.
i j
April'10, 1989-Letter from E. E. Kintner, EPRI, to NRC, forwardir g ALWR Requirements Document, Chapter 11.
EPRI believed major j
objectives had been achieved and concerns about station l
blackout would be virtually eliminated.
i j
April 11, 1989 Letter from EPRI, to NRC, for.arding Revision 0 of ALWR Utility Requirements Document, Chapter 11.
1 Program Summary A-14 i
April 28, 1989 Letter from W. O. 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.
May 16, 1989 Letter from W. O. Long, NRC, to EPRI, discussing ALWP, 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.
Ma,v 17, 1989 Letter from W. O. 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 snd ignition, and design-buis-aucident source term assumptions.
May 24, 1989 Letter from W. Long to EDRI requesting additional information on Chapter 7, 8,11, and 12 of the ALWR Requirements Document.
June 8, 1989 Letter from W. O. Long, NRC, to EPRI, requesting additional information on GLs-81-12 and 8'-?u.
June 1989 Letter from W. O. Long, NRC, to EPRI, regarding GSIs s,
II.G.1 and II.G.23 pertaining to pressurizer e<1uipment electric power and threat of reactor coolant pump seal n
loss of-coolant-accident.
hly 3, 1989 Letter fror E. E. Kintner, EPRI, to NRC, forwarding Revision 0 af ALWR 'Jtility Requirements Documei,
Appendix A to Chapter 1.
Appendix to provide guidance to be used in performing PRA.
July 3, 1989 Letter from E. E. Kintner, EPRI, to NRC, forwarding Chapter 6 of ALWR Utility Requirements Document.
July 14, 1989 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.
July 19, 1989 Letter from T. Kenyon, NRC, to-EPRI, summarizing June 22, 1939, meeting with EPRI representatives of nuclear power-industry and source term and related issues for ALWR program.
List of attenlees and handouts enclosed.
August 10, 1989 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.
Program Summary A-15
September 15, 1989 letter from E. E. Kintner, EPRI, to NRC, forwarding response to M'" 24, 1989: request for additional information on Chapters 7, 8, 11, and 12 of ALWR Utility Requirements Document.
Power supplies of emergency response facility system to be designed in accordance with NUREG-0696.
October 19, 1989 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.
October 26, 1989 Letter from E. E. Kintner, EPRI, to NRC, forwarding ALWR Utility Requirements Document, Chapter 10, and Topic Paper, " Reactor Pressure Vessel Levd Instrumentation for PWRs."
November 6, 1989 Letter from J. L. Blaha, NRC, to SECY, forwarding correspondence between EPRI and NRC on Chapter 5 of ALWR Utility Requirements Document.
November 28, 1989 Letter from T. J. Kenyon, NRC, to EPRI, requesting information on reactor safeguards within 60 days of date of letter.
December 22, 19E3 Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to requests for additional information on Chapters 1, 6, 7, 8, 9, 11, and 13 cf ALWR.
December 26, 1989 Letter from E. E. Kiilter, EPRI, to NRC, regarding review priorities and process.
January 18, 1990 Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to <; quests for additional information on Chapters 6, 7, 8, 9, 12, and 13 of ALWR Utility Require.m ts Document regarding physical security, inside. e d outsider sabotage threats, and controlled access to the containment.
January 31, 1990 Report by Fauske and-Associates, Inc., " Technical Support for Hydrogen Control Requirement for EPRI Advanced LWR Requirements Document."
February 3, 1990 Letter from E. E. Kintner, EPRI, to NkC, forwarding
" Technical Support for Hydrogen Control Requirement for EPRI Advanced LWR Requirements Document."
February 22, 1990 Letter from E..E. Kintner, EPRI, to NRC, forwarding updated Revision 0 of ALWR Utility Requirements Document, Appendix A to Chapter 1.
Document reissued to-submit-Section 3.2.2, " Earthquake," for NRC review.
Program Summary A-16
Fd 'uary 27, 1990 Letter fr3m K. M. Carr: NRC, to EPRI, responding to December 26, 1989, letter expressing conce'ns regarding progress of NRC review of ALWR Utility Requitements Document.
February 28, 1990 Letter from C. L. Miller, NRC, to EPRI, forwarding DSER Chapter 5 on ALWR Utility Requirements Document, 4
including severe-accident prevention and mitigation, hydrogen generation, control source term issues, and station blackout.
March 16, 1990 Letter from E. E. Kintner, FPRI, 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.
March 29, 1990 Letter from E. E, Kintner, EPRI, to NRC, regarding treatment of generic safety issues for ALWR-requirements.
March 31, 1990 Letter from EPRI to NRC, forwarding Volume I, "ALWR Policy and Summary of Top-Tier Requirements." of ALWR Utility Requirements Jocument.
April 10, 1990 Letter from T. J. Kenyon, NRC, to EPP.1, requesting addi-tional information on Volunie II' or AlWR Utility Require-ments Document on the basis of staff's review of Appendix A to Chapters 10 and 11.
Response requested within 60 days of letter date.
June 20, 1990 Litter from T. J. Kenyon, NRC, to EPRI, summarizing December 6, 1989, meeting-with EPRI regarding Chapter 10 of Volume II of ALWR Utility Requirements Document.
List of attendees and viewgraphs enclosed.
June '.,J, 1990 Report by Fauske and Associates, Inc., " Technical Support for Debris Coolability Requirements for Advanced LWRs in Utility-EPRI LWR Requirements Document."
July 3, 1990 Letter from S.~H. Smith, Nuclear Power Oversight Commit-tee, to NRC, advising NRC that U.S. utility ndustry vitally interested-in timely certification of both evclutionary and cassive ALWR designs currently under-review by NRC.
Maintaining current certification review schedules important.
July 3, 199C Letter from E. E. Kintner, EPR!, to NRC, forwarding g
" Technical Support for Debris Coolability Requirements for Advanced LWRs in_ Utility-EPRI LWR Requirements Document."
July 13, 1990 Letter from T. J. Kenyon, NRC, to EPRI, reque: ting information needed to complete review of design criteria.
Program Summary A-17
w July 23, 1990 Letter from E. E. Vintner,-EPRI, to NRC, forwarding response to April 10, 1990, request for adait.onal information on Chapters 1, 10, and 11 of ALWR Utility Requirements Document.
July 23, 1990 Letter from EPRI, to NRL, forvrarding "Presention of Early Containment Failure Due t, High Pressure Melt Ejection and Direct Containment Heating for Advanced LWRs."
August 2, 1990 Letter from T. J. Kenyon NRC, to EPRI,,"? " sting additional information on Chapter 10 c. U ' me 11 of ALWR-Utility Requirements Document regardir ectr, al system.
August 2, 1990 Letter from T. J. Kenyon, NRC, to EPRI, summarizing May 31, 1990, meeting with EPRI. NIHARC and standardized plant vendors regarding Volume 11 of ALWR Utilit/
Requirements Document.
List of attendees enclosed.
O August 8, 1990 Letter fr;m T. J. Kenyon, NRC, to EPRI, summarizing May 31, 1990, meeting.with EPRI regarding source term to be used for future LWR.
List of attendees and viewgraphs enclosed.
August 13, 1990 Letter from J. M. Taylor, NRC,:to S. H. Smith, Nuclear Power C-tersight' Committee, responding to July 3,1990, letter to Chairman K. Carr regarding timely certification of two evolutionary passive designs.
Sufficient time needed ny NRC staff to identif; ar.d evaluate issues to ensure that proposed ALWR designs provide adequate protection.
t August 15, 1990 Letter from 1. J. Kenyon, NRC, to EPRI, forwarding comments regarding unresolved and generic safety issues cddressed in Appendix B to Chapter 10 of Volume 11lof
[
ALWR Utility Requirements Document.
August 22, 1993 Letter from T. J. Kenyon, NRC, to EPRI, requesting g
- additional information on Volume 11 of ALWR Utility Re-quirements Document _regarding design criteria. ' Response requested within 60 days of_ letter-.
August 23, 1990_
Letter from T. J. Kenyon, NRC, to EFRI, forwarding-
-summary of July-16, 1990,- meeting-with EPRI regarding hydrogen generation and containmeat-performance.
List of-attendees, NRC slides, and EPRI presentation enclosed.
August 27, 1990 cetter from T. J. Kenyon, NRC,Lto EPRI, ferwarding revised request for~ additional information on Chapter-1 of Volume -11 of ALWR UtilityL Requirements-Document.
. Program _ Summary A-18 8
e r
August 30, 1990 Letter from T. J. Kenyon,.iRC, to EPRI, requesting additional information on ALWR Utility Requirements Document, Volume I, and Chapters 1, 6, and 10 of Volume i
11 to complete review of design criteria.
i August 31, 1990 Letter from EPRI, to NRC, submitting annotated Revision 1 of Volume II to ALWR Utility Requirements Document.
1 September 7, 1990 Letter from E. E. Kintner, EPRI, to NRC, forwarding proprietary Revision 1 of Volume II (ALWR evolutionary i
plant) and Revision 0 of Volume III (ALWR passive plant) of ALWR Utility Requirements Document for safety evaluation review.
Documents no longer withheld per March 29, 1991, letter from EPRI.
1 September 30, 1990
'leport by D. E. Leaver and L. P. Tenera, " Licensing Design Basis Source Term Update for Evolutionary Advanced LWR."
2 October 12, 1990 Letter from E. E. Kii.tner, 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.
~
October 18, 1990 Letter from E. E. Kintner, EPRI, to NRC, forwarding i
" Licensing Design Basis Source Term Update for i
Evolutionary Advanced LWR."
October 29, 1990 Letter from W. R. Sugnat, EPRI, to NRC, forwarding tables listing NRC rules and regulatory guidance not in Appendix B to Chapter 1 of Volumes II and III of Advanced LWR Utility Requirements Document.
November 7, 1990 Letter from E. E. Kintner, EPRI, to NRC, forwarding-additional information on Appendix A to Chapter 1 and Chapter 12 of ALWR Utility Requirement Document.
2 November 29, 1990 Transcript of_ closed-session meeting with ALWR Utility Steering Committee on November 29, 1990, with NRC.
Pp 64-223.
Supporting information enclosed. Transcript r.o longer proprietary per March 29, 1991 letter from EPRI.
l December 6, 1990 Letter from E. E. Kintner, EPRI, to NRC, forwarding proprietary response to Autust 30, 1990, request for additional information on himan factors considerations.
i Response addressed questions a ALWR Utility Requirements Document, Volume I, and Chapters 1, 6, and 10 of Volume II.
Enclosure no longer withheld per March 29,.1991, letter from EPRI.
December 21, 1990 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on unresolved and generic safety issues and addressing I
i Program Summary A
questions on Volume II of ALWR Utility Requirements Document.
Response no longer withheld per March 29, 1991, letter form EPRI.
December 21, 1990 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-Utility Requirements Document.
January 9, 1991 Letter from W. R. Sugnet, EPRI, to NRC, forwarding proprietary information on ALWR seismic design evaluation program.
Enclcsures no longer withheld per March 29, 1991, letter from EPRI.
January 15, 1991 Letter from C. L. Miller, NRC, to EPRI, forwarding DSER on Chapters 6, 7, 8, 9,12, and 13 of Volume II of ALWR Utility Requirements Document.
January 25, 1991 Letter from T. J. Kenyon, NRC, to EPRI, requesting additional information on Volumes II and III of ALWR Utility Requirement; Document regarding quality assurance.
Document no longer withheld per March 29, 1991 letter from EPRI.
January 25, 1991 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 II of ALWR Utility Requirements Document.
Response no longer withheld per March 29, 1991 letter from EPRI.
January 28, 1991 Letter from W. R. Sugnet, EPRI, to NRC, forwarding comparison between Volumes II, and III of ALWR Utility Requirements Document.
Enclosures on longer withheld per March 29, 1991 letter from EPRI.
January 30, 1991 Letter from W. R. Sugnet, EPRI, to NRC, forwarding data base for tracking open issues in DSER for Chapters 1-5 of Volume II of ALWR Utility Requirements Document.
Enclo-sure no longer withheld per March 29, 1991 letter from EPRI.
January 31, 1991 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.
February 1, 1991 Letter from W. R. Sugnet,:EPRi, to NRC, forwarding Revision 1 of " Development of Seismic Hazard Input fer Advanced LWR Seismic PRA," in response to NRC request at meeting on January 14, 1991.
L February 4, 1991' Letter from D. Cratchfield, NRC, to EPRI, providing preliminary views on Revision 0 of Volume III of ALWR Program Summary A-20
~
i Utility Requirements Document regarding criteria to be used in design for combustible gas control. Document no i
longer withheld per march 29, 1991 letter from EPRI.
February 7, 1991 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 l
Requirements Document.
Summary no longer withheld per i
March 29, 1991 letter from EPRI.
February 7, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding responses to questions raised at meeting on November 29 and 30, 1990.
Numbering scheme fo responses also enclosed.
Responses no longer withheld per March 29, 1991 letter from EPRI.
f February 22, 1991 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, " Guidelines for Piping System Reconciliation (NCIG-05, Revision 1)."
February 28, 1091 Letter from a R. Sugnet, EPRI, to NRC, forwarding I
Sections 1-5 of Chapter 1, Volume II of ALWR Utility Requirements Document to propose level of information to be witheld from public disclosure. Document no longer withheld per March 29, 1991 letter from EPRI.
February 28, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting
[
. additional information on Volume III of ALWR Utility Requirements Document.
I February 28, 1991 Report by M. W. McCann, Jack R. Benjamin and Associates, Inc., P.evision l'of " Development of Seismic Hazard Input for Advanced LWR Seismic PRA."
March 3, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting i
additional information Volume III of ALWR Utility Requirements Document regarding quality assuran:e.
Document no longer withheld per March 29, 1991 letter from EPRI.
[
March 1, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume III of ALWR Utility Re-quirements Document regarding physical security-and safeguards requirements.
Document no longer proprietary per March 29, 1991 letter from EPRI.
M rch 1, 1991 Letter from J. H. Wilson,_NRC, to EPRI, requesting additional information on Volume III cf ALWR Utility j
Requirements Document regarding plant systems. Document no longer withheld per March 29, 1991 letter from EPRI.
I f
Program Summary A-21
=
March 1, 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.
March 1, 1991 Letter from J. H. Wil.,on, NRC, to EPRI, providing summary of N bruary 11 and 12, 1991, meetings with EPRI in Rock-vilie, Maryland, to discuss NRC staff review of Volume 11 of ALWR Utility Requirements Document.
List of attendees
- enclosed, March 1, 1991 Letter from J. H. Wilson, NRC, to EPRI, correcting summa-ry of February 11 and 12,1991, meetings in Rockville, Maryland, to discuss NRC staff review of Volume 11 of ALWR Utility Requirements Document.
March 1, 1991 Letter from W. R. Sugnet, EPRI, to NRC, forwarding industry list of technical issues central to design of ALWR passive plants.
March 8, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume III of ALWR Utility Re-quirements Document regarding reactor systems. Document no longer withheld per March 29, 1991 letter from EPRI.
March 14, 1991 Letter from E. E. Kintner, EPRI, to i4RC, providing perspective of progress on work on ALWR Utility Require-ments Document.
March 15, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to January 25, 1991, request for additional information on quality assurance.
Response no longer withheld per March 29, 1991 letter from EPRI.
March 15,-1991 Letter from E. E. Kintner, EPRI, to NRC,-forwarding-response to requests for additional information on quality assurance.
Enclosure no longer withheld per March 29, 1991 letter from EPRI.
March 19, 1991 Letter from T. H. Boyce NRC. to EPRI, providing summary of March 5 and 6, 1991, meetugs with Combustion Engineering and EPRI in Rockville, Maryland, regarding NRC staff review of Volume II of ALWR Utility Re-quirements Document.
List of attendees and agenda en-closed.
March 28, 1991 Letter from T. H. Boyce, NRC, to EPRI, providing summary of March 28, 1991, meeting with EPRI in Rockville, Maryland, to discuss hRC staff review of Volume II of ALWR Utility Requirements Document.
Program Summary A-22
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=_
April:3, 1991 Letter from C. L. Miller,-NRC, to EPRI, forwa-ding DSER on Chapter ll of Volume II of ALWR Utility Requirements Document.
April;3, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information-on Volt.me. III of ALWR Utility Requirements Document. :Information needed to complete review of design criteria addressed during review of
- Chapter 9, " Site Support System."
April. 5,1991 Letter frcm-J. H. Wilson, NRC, to EPRI, concluding that regulations and guidance not applicable to design requirements should be added tc " 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.
'L tter from J. H. Wilson, NRC, to EPRI, requesting EPRI's April 8, 1991 e
position on changes to criteria in Volume III of ALWR-Utility Requirements Document that apply to evolutionary designs.
April 17, 1991 Letter froir J. H. Wilson,- NRC, to EPRI, requesting information on Volume III of ALWR Utility Requirements Document regarding human factors.
April 17, 1991
- Letter from J. H. Wilson,-NRC, to EPRI, requesting additional information on Volume III of ALWR Utility Requirements Document regarding safeguards.
April 22, 1993 t.etter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume -III of ALWR Utility Requiremen#s Document regarding reactor systems.
April 24, 1991 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 tnat review of-advanced passive reactors presented different kind of challenge.
April 24,~1991 Letter from J. H. Wilson, NRC, to EPRI, requesting
- additior.al-.information on Volume III of ALWR Utility Requirements Docement'regarding structural engineering-Concerns.
April-25, 1991-Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume III of ALWR Utility Requirements Document regarding materials and chamcial engineer 4 ig concerns.
April-25,:1991 Letter from E.-E. Kintner, EFRI, to-NRC, forwarding response to requests =for additional information on Program Summary LA-23 0
Chapters 4 and'7 of Volume 111 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 11 and Revision 1 of Volume III of-ALWR Utility Requirements Document-in-response to NRC concerns regarding quality assurances, human factors, generic and unresolved safety issues, and open issues in DSERs.
Changes resulting from elimination of operating-basis earthquake also included.
April 30, 1991 Letter from EPRI, to NRC, submitting copies of Revision 1 of Volume 11 of ALWR Utility Requirements Document; and Chapter 1 and Appendix B to Chapter 1 updated through Revision 2.
April 30, 1991 Letter from EPRI, to NRC, transmitting revised pages issued as Revision 2 of Volume 11 and Revision 1 of Volume 111 of ALWR Utility Requirements Document.
April 30,-1991 Letter _from EPR1,'to NRC, transmitting " Position Paper on-Standardization."
May 6, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information_on Volume III of ALWR Utility Requirements Document regarding materials engineering concerns.
May 6, 1991 Letter from D. Crutchfield, NaC, to EPRI, clarifying issues discussed in SECY-90-016.
May 8, 1991 L9tter from E. E. Kiatner, EPRI, to NRC, forwarding response to April 8,1991, letter-regarding consistent changes to Volumes 11 and_111 of ALWR Utility Requirements Document.
May 8, 1991 Letter from E. E. Kintner, EPRI, te NRC, forwarding response to March 8, 1991, recuest for additional information on Chapter 4 of 6 lume 111 of ALWR Utility Requirements Document.
May 8, 1991 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, May 9, 1991 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 11 of ALWR Utility Requirements Document, Program Summary A-24 J
l May 9, 1991 Letter from E. E, Kintner, EPRI, to NRC, forwarding j
matrix approach-to containment performance criteria fc j
evolutionary plants and rationale for their selection.
i i
May 13, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to March 1,_1991, _ request for additional information on-Volume III of-ALWR Utility Requirements Document regarding measures to be taken to prevent high operational 'and post-shutdown radiation levels in reactor
]_
coolant system piping, i-1 May 13, 1991-Letter from E. E. Kintner, EPRI, to NRC, forwarding i
response to February 28,-1991, request for additional 1
j information on Volume III of ALWR Utility Requirements l
Document regarding system roliability, quality assurance, and seismic qualification or capability of non-safety-j grade active systems.
l-May 13, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding l
response to March 1,1991, request for additional 1
information on-Volume III of ALWR Utility Requirements l
Document regarding resolution of GSI A-29 for advanced reactors.
l May 13, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding partial response to March 1, 1991, request for additional f
information on Volume II of ALWR Utility Requirements regarding fire protection.
1 May_13, 1991 Letter from EPRI, to NRC forwarding response to March 1,
[
1991, request for additional information on Volume III of l
ALWR Utility Requircments Document, i
May 17, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting i
additional information on. Volume.III of ALWR Requirements l
Document regarding mechanical engineering concerns.
l-May 17, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on _ Volume ~III of ALWR Utility j--
Requirements Document regarding instrumentation and j
controls.
May 17, 1991 Letter from J. H. Wilson, NRC, to EPRI, forwarding 1-page 16 omitted from April _ 21,-1991, request for addi-l-
tional information regarding review of Chapter 5 of Volume III of ALWR Utility Requirements Document, i
May 17, 1991-Letter from J. H. Wilson, NRC, to EPRI, requesting i
additional information on Volume III of ALWR Utility
{
Requirements Document regarding reactor systems.
i
}
)
i j
Program Summary A-25 i -
I
~-
. _ = -
_=_ _
May 17, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting addi-tional information on Volume 111 of _ ALWR Utility Require-ments Document submitted on September 7, 1990 regarding safeguards considerations.
May 22, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on Volume 111 of ALWR-Utility Requirements Document regarding plant systems.
May 22, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to request for additional information on Volume III of ALWR Requirements Document regarding radiation protection.
May 22, 1991 Letter from G. Bockhold, EPRI, to NRC, forwarding updated open issues tracking system for Chapters 1-9, 11, and 12 of Volume 11 of ALWR Utility Requirements Document.
June 5, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting addi-tional information on reactor systems to complete review of Volume II of ALWR Utility Requirements Document.
June 13, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to April 17, 1991, request for additional information on Volume 111 of ALWR Utility Requirements Document regarding human factors.
June 24, 1991 Letter from E. E. Kintner, EPRI, to NRC,-forwarding responses to April 3 and 17, 1991,. requests for addi-tional information on ALWR Utility Requirements Document regarding safeguards concerns.
July 1, 1991 Letter from E.-E. Kintner, EPRI, to.NRC, forwarding response to requests for additional information on_ Volume ill of ALWR Utility Requirements Document regarding compliance of leak-before-break methodology with accep-tance criteria in statement of considerations for final rule,-General Design Criterion 4 of Appendix A to 10 CFR Part 50.
July 1, 1991 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 I of Volume III of ALWR Requirements Document regarding reliability requirements for non-safety critical systems.
July 2, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to April 24, 1991, request for additional information on Volume III of ALWR Utility Requirements Document regarding-safety-margin basis requirements.
Program Summary A-26
July 8, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991,. request for additional infor-mation on Volume III of ALWR Utility Requirements Docu-ment regarding emergency planning.
July 22, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume III of ALWR Utility Requirements Document regarding methodology for dedicating commercial software for designing safety-related systems.
July 22, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume III of ALWR Utility Requirements Document regarding site security interfaces in plant security systems and key-locked controls.
July 22, 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 Document regarding safes, single-action valves, security barriers, and intrusion detection systems.
August 1, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to April 22, 1991, request for additional information on Volume III of ALWR Utility Requirements Document regarding reactor systems.
August 1, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume III of ALWR Utility Requirements Document regarding mecnanical engineering regarding material and chemical engineering concerns.
August 1, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding final response to May 17, 1991, request for additional infor-mation Volume III of ALWR Utility Requirements Document.
August 12, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to June 5,1991, request for additional infor-mation on Volume Il of ALWR Utility Requirements Document regarding reactor systems.
August 16, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting addi-tional information on Volume III of ALWR Utility Requirements Document.
August 19, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume III of ALWR Utility Requirements Document regarding human factors.
August 19, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting addi-tional information on Volume III of ALWR Utility Requirements Document regarding safeguards.
Program Summary A-27
L August 20, 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.
August 29, 1991 Letter from D. Crutchfield, dRC, to EPRI, requesting additional information on Volume Ill-of ALWR Utility-Requirements Document.regarding Question 210.40.
September 5, 1991 Letter from J.- H. Wilson, NRC, to EPRI, requesting addi-tional information on Volume II of ALWR Utility Require-ments Document regarding shutdown risks.
September 5, 1991 Letter from J H. Wilson, NRC, to EPRI, requesting addi-tional information on Volume lil-of ALWR Utility Requirements Document regarding shutdown risks.
September 6, 1991 Letter from J. H._ Wilson, NRC, to EPRI, transmitting open issues from staff review of Appendix A to Chapter 1 of Volume 11 of ALWR Utility Requirements Document.
September ll, 1991 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.
September 11, 1991 Letter frnm J. H. Wilson, NRC, to EPRI, requesting additionai information on_ Volume ill of ALWR Utility-Requirements Document regarding unresolved and generic safety issues.
September 23, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Appendix A to Chapter 1 of Volume 111 of ALWR Utility Requirements Document.
October 1, 1991 Letter from G. Bockhold, EPRI, to NRC, forwarding Volumes 11 and III of ALWR_ Utility Requirements Document consisting of-hand-marked page-changes in accordance with NRC_ request.
October _2, 1991
' Letter _from J. H. Wilson, NRC, to EPRI, summarizing
- August-27, 1991, meeting with utilities in Rockville, Maryland, regarding development.of updated source term for LWRs.
October 2, 1991 Letter from J. H. Wilson, NRC, to-EPRI, su'ni.u.rizing
-August 14 and 15, 1991, meetings with EPRI'in-Rockville,-
Maryland, to discuss. issues associated with staff review of Volume III ALWR Utility Requirements Document.
October 3, 1991 Letter from R. Chambers, Idaho National l Engineering Labo-ratory, to NRC, forwarding two copies of Advanced Reactor Severe Accident Program report " Interim External-Events Integrati_on for EPRI Advanced LWR Requirements Document Program Summary A-28
WBS 4.3.3,"
DOE-ID-10227, by D. G. Harrison in response to request by J.
D._ Trotter of EPRI'to support review of Project 669.
October 8, 1991 Letter from R. C. Pierson, NRC, to EPRI, forwarding DSER on Chapter 10 of Volume 11 of ALWR Utility Requirements Document.
October 9, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to August 19, 1991, request far additional information on Volume III of ALWR Utility Requirements Document.
Battery rooms to be locked and alarmed, and isolation zone lighting to be designed to permit observation.
October 10, 1991 Letter from J. H. Wilson, NRC, to EPRI, requesting additional information on Volume Ill of ALWR Utility Requirements Document submitted on September 7, 1990,-
regarding the reliability assurance program.
October 10, 1991 Letter from J. H Wilson, NRC, to EPRI, forwarding-open issues from review of Section 6 of Chapter 1 of Volume -11 of ALWR Utility Requirements Document.
October 17, 1991 Letter from G. Bor N d, EPRI, to EPRI, forwarding response to questr.' raised during teleconference.
regarding ALWR seise.c hazard curve.
October 23, 1991 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 11 of_ ALWR Utility Requirements -Document transmitted by letter dated October 8, 1991.
October 30, 1991 Letter from J. H. Wilson, NRC, to EPRI, summarizing September 26, 1991, meeting with EPRI in Rockville, Maryland, to present results of research by NRC and contractors concerning development of updated sourco term.
List of attendees provided.
November 4, 1991 Letter-from R. C. Pierson,-NRC, to EPRI, forwarding-DSER on Appendix A to Chapter 1 of Vo)ume II.of ALWR Utility Requirements Document.
Open issues required resolution.
November 4, 1991 Letter from G. Bockhold, EPRI, to'NRC, forwarding information regarding errors and resoiutions in response to request for additional information on ALWR Utility Requirements Document.
November 6, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to August 29, 1991, request for additional information on. Volume 111 of ALWR Utility Requirements Document regarding mechanical-engineering concerns.
Program Summary A-29
_~ - -.. -
1
)
i, h
l November 15, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to August 19, 1991, request for additional j
information on Volume-111 of ALWR Util_ity Requirements Document regarding human factors.
i November 20, 1991 Letter from E. E. K1ntner, EPRI, to NRC, forwarding response to September 11, 1991, request for additional information on Volume III of ALWR Utility Requirements Document regarding unresolved and generic safety _ issues.
November 25, 1991 Letter from E. E. Kintner, EPRI, to NRC, forwarding Revision 3 of Volume II of ALWR Utility Requirements j
- Document, j
t November 27, 1991 Letter from E. E. Kintner and J. J. Taylor, EPRI, to NRC, proposing agenda of meetings _at senior levels of NRC i
staff and Utility Steering Committee to discuss remaining l
generic safety issues in ALWR Utility Requirements Docu-ment as discussed at workshop on November 4 and 5, 1991.
}
December 2, 1991 Letter from E. E. Kintner, EPRI, to_NRC, forwarding response to September 5, 1991, request for additional l
information on Volume 11 of ALWR Utility Requirements i
Document regarding snutdown risk.
December 6, 1991 Letter from EPRI, to NRC, forwarding ALWR positions on central issuer pertaining to evolutionary plant
_i identified during July 1991 meeting with NRC.
Issues included containment performance, core debris i
coolability, ar.d seismic hazard.
December 10, 1991 Letter from J. D. Trotter, EPRI, to NRC, forwarding pen-and-ink changes to Chapters 5, 6, 9, 10, and 11 of Volume II of ALWR Utility Requirements Document regarding security and suggesting working-level meeting during _ week of January 6,1992, to resolve any outstanding concerns.
December 16, 1991 Letter from E. E.--Kintner, EPRI, to NRC, forwarding 1
response to September 5_,1991, request for -additional i
information-on Chapters 4 and 5 of Volume II of ALWR-l Utility Requirements Document'regarding scope of PRA.for operating' conditions when plant is at power and general initiating events.
December 18, 1991 Letter from G. Bockhold, EPRI, to NRC, advising that modification of Figure 12.3-1 in Chapter 12 of Volume 11 of ALWR Utility _ Requirements _ Document was inadvertently-i omitted from. list of changes submitted to NRC in-June and i
October 1991.
December 20, 1991 Letter from D. Crutchfield, NRC, to EPRI, (Jentifying issues pertaining to evolutionary and passive' plant de-i signs.
Information provided to initiate discussion of l
approaches to resolving issues.
l 4
Program Summary A-30 t
i- - - -
December 20, 1991 Letter from J. D. Trotter, U RI, to NRC, forwarding summary of technical rationale by C. Negin regarding high-efficiency particulate air filters in radioactive waste off-gas systen.
December 21, 1991 Letter from J. D. Tretter, EPRI to NRC, forwarding-proposed n ior revisions of ALWR Utility Requirements Document audressing open issues pertaining to reactor pressure vessel materials.
December 26, 1991 Letter from D. Crutchfield, NRC, to iPRI, discussing preparation of final SER on Volume 11 of ALWR Utility Requirements Documents and requesting that EPRI submit all responses and positions by January 31, 1992, so that staff could complete its technical evaluations.
January 9, 1992 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 11 of ALWR Utility-Requirements Document.
January 9, 1992 Letter from E. E. Kintner, EPRI, to NRC, forwarding response to-0ctober 10, 1991, request for additional information on Volume Ill of ALWR Requirements Document.
January 10, 1992 Letter from E. E. Kintner, EPR'., to NRC, forwarding response to August 16, 1991, request-for. additional information on Volume 111 of ALWR Utility Requirements Document regarding electrical-systems.
January 10, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding response to May 17, 1991, request for additional information on Volume 111 of ALWR Utility Requirements Document regarding core support structures because initial response dat2d August 1,1991, was inappropriate.
January 22, 1992 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.
January 24, 1992 Letter from J. H. Wilson, NRC, to EPRI, summarizing January 8, 1992, meeting with EPRI in Rockville, Maryland, to discuss-EPRI-proposed changes to security requirements.
-January 24, 1992 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-Ill of ALWR Utility Requirements Document.
Program Summary A-31
January 24, 1992
-Letter from G. Bockhold, EPRI, to NRC, forwarding responses to open and confirmatory issues in DSER on-Appendix A to Chapter 1 of Volume 11 of ALWR Utility Requirements Document.
Enclosure included both data base summaries for each issue and markups of pages in Volume 11 pages.
4 l
January 24, 1992-Letter from E. E. Kintner, EPRI, to NRC, submitting j
changes to Chapters 5, 6, 9, 10, and 11 of Volume 11 of i
j ALWR Utility Requirements Document that dealt with
}
l safeguards-related requirements.
Changes should be
[
j considered in final SER on Volume 11.
y i
r i
January 24, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding 1
respor.ses to three issues in DSER on Chapter 11 of Volume 11 of ALWR Utility Requirements Document.
Enclosure l
1
' :cludes b*h data base summaries for each issue and i
markups of pages in Volume II.
l January 28, 1992 Letter from G. bcckhold,. EPRI. to NRC, forwarding first i
part of responses to oper and confirmatory issues (99 of i
i 126 issues) in DSER on Chapter 10 of Volume 11 of ALWR i'
Utility Requirements Document. Minor proposed changes to be reflected in Revision 4 of Volume II.
i February 3, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding second I
part of responses to open and confirmatory issues in-DSER t
on ' hapter 10 of Volume 11 of ALWR Utility Requirements 4
Document and stating that responses had been given for
. i
+
all but four of the issues.
Responses to remaining four l
issues to be provided later this montn.
j February 3, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding changes to Volume II of ALWR Utility aequirements Document that addressed DSER open.and confirmatory issues or other
-concerns.
Changes to be incorporated into Revision 4 of l
Volume II.
l February 4, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding 4
prvosal for maintenance feed to top-tier electrical bus
- action, j
February 10, 1992 Letter from T. U. Marston, EPRI, to NRC, suggesting staff-level meetings be held in late February and early.
- f March-of 1992 -to expedite quality closure' of DSER issues t
certaining to chapters on man-machine interface systems and PRA in Volume II of Utility Requirements Document.
j
?
February 11,.1992 Letter from J.-D. Trotter, EPRI, to NRC, discussing
[
miscellaneous. items for Project 669 including resolution i
of steam generator tube ruptures by increasing design i
-pressure.
Program Summary A-32 i
.m..
..i _,
m...
, ~.. -......
Y February 18, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding i
responses to four open issues in DSER on Chapter 10 of i
Volume II of ALWR Utility Requirements Document in-i response to October 10, 1991, request for additional i
information and containing both data base and road-m'ap j
summaries for each issue and markups of pages in j
Volume II.
l February 2/, 1992 Letter from D. Crutchfield, NRC, to EFRI, forwarding l
draft Commission paper describing major technical and policy issues on evolutionary and passive plant designs.
Positions supersede those in-February 20, 1991 letter.
l March 2, 1992 Letter from T. J. Kenyon, NRC, to EPRI, summarizing i
January 30, 1992 senior management meeting with EPRI and l
nuclear industry representatives on technical _ issues _for j
evolutionary and passive-ALWRs.
List of attendees and l
presentations enclosed.
j March 3, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding 1
February 1992 status report.
Significant progress j
apparent in area of policy issues.
1 March 3, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding proposed changes to Chapter 1, Appendix B to Chapter 1, l
and Chapters 10 and 11 of Volume II of ALWR Utility j
Requirements Document.
1 i
March 9, 1992 Letter from T. J. Kenyon, NRC, to EPRI, issuing errata to l
summary of senior management meeting on January 30, 1992 j.
on technical issues for evolutionary and passive ALWRs.
March 10, 1992 Letter from J. D. Trotter, EPRI, to NRC, forwarding final draft of position paper.for passive plant system l
classification and requirements.
t March 19, 1992 Letter from E. E. Kin,1er,-EPRI, to NRC,-forwarding position paper on passive plant system classification and requirements.
March 19, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding changes i
to Volume II of ALWR Utility Requirements Document'per
^
continuing discussions with NRC.
March 30, 1992 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 i-distribution policy issue.
March-31, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding changes to Volume III of.ALWR Utility Requirements Document j.
addressing open issues on performance requirement for turbine exhaust boot and design of radial and thrust
}.
bearings.
. Program Summary A =
7.__.____..______
t i
}
April 3, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding comments on draft NUREG/CR-5747z" Estimate of Radionuclide Release Characteristics into: Containment Under' Severe t
l Accident Conditions."
April 7, 1992 Letter from G. Bockhold, EPRI, to NRC, submitting response to NRC request for additional information-on 1
j second source of power to_non-safety loads required l
exclusively for unit operation as discussed in Sections 3.3 and 4.2 of Chapter 11 of DSER.
l April 9, 1992 Letter from G, Bockhold, EPRI, to NRC, on diversity-against common-mode software failures. Listed elements provided in Chapter 10 of ALWR Utility Requirements l
- Document, g
l April 9, 1992 Letter from W. Borchardt, NRC, to EPRI,. summarizing i
March 27, 1992 meeting with EPRI in Denver, Colorado: on-i major issues resulting from NRC review of ALWR Utility Requirements Document for evolutionary and passive I
designs.
List of attendees, meeting agenda handouts,'and.
slides enclosed.
I April 9, 1992 Letter from G. Bockhold, EPRI, to NRC, submitting summary of methods & assumptions used in development of ALWR 80th percentile meteorological dat' abase.
i
~
l April 17, 1992 Letter from G. Bockhold, EPRI, to NRC, submitting changes to Chapter 3 of Volume II of ALWR Utility Requirements l
Document concerning testability of third feedwater l
isolation valve in bwrs.
Similar change will be made to i
Volume 111.
i l
April 17, 1992 Letter from E. E. Kintner, EPRI,.to NRC, forwarding i
i Revision 4 to Volume II of ALWR Utility Requirements
{
l-
- Document, i
April 24, 1992 Letter from D. Crutchfield,-NRC, to EPRI, forwarding DSER
{
on Volume III of ALWR Utility Requirements Document, i
i:
April 30, 1992 Letter from D. Crutchfield, NRC, to EPRI, discussing i
topics under consideration-for inclusion in Commission 1
paper to discuss additional issues on future reactor designs.
May 1,-1992.
Letter from G. Bockhold, EPRI, to NRC, forwarding i_
respnnse-to open issue on the use of physically-based source term on Volume II of the ALWR Utility Requirements Document.
4 May 5, 1992 LetterfromG.Bockhold,ErRI,toNRC; forwarding-
_ modifications to ALWR Utility Requirements Document to address general concern of NRC on vulnerability of_ ALWRs during shutdown & low power operation.
-Program Summary A-34 i
l t
l May 5, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding draft _
~
i of PWR and BWR passive plant system classification.
Enclosure submitted to facilitate preparations for
-t May 14, 1992 meeting on regulatory treatment of nonsafety.
systems.
May 8, 1992 Letter from G. Bockhold, EPRI, to NRC, forwarding EPRI's j
draft positions on additional technical and policy-issues j
j on passive and evolutionary plant designs in parallel l
with Utility Steering Committee review.
May 13, 1992-Letter from E. E. Kintner, EPRI, to NRC, forwarding I
Revision 3 to Chapters 1-13 of Volume III of the ALWR
[
Utility Requirements Document.
)
May 15, 1992 Letter from D. Crutchfield, NRC, to EPRI, forwarding draft of SER on Volume II of' ALWR Utility Requirements Document.
i May 28, 1992 Letter from T. J. Kenyon, NRC, to EPRI,_ summarizing.
l meeting with EPRI on May 6, 1992_on in-containment i
fission product removal mechanisms-for evolutionary and i
passive plant designs.
I May 26, 1992 Letter from T. J. Kenyon, NRC, to EPRI,-summarizing 1
meeting with EPRI.on March 20, 1992 on EPRI's initial work on reliability-based-technical specifications for passive ALWRs.
List of attendees and EPRI presentation included.
l i
June 4, 1992 Letter from E. E. Kintner,-EPRI, to NRC, responding to i
February 27, 1992 letter on major technical-and policy
}
issues for evolutionary and passive plant designs June 15, 1992 Letter from G. Bockhold, EPRI, to-NRC, forwarding Apil and May 1992 status report June 23, 1992 Letter from D. Crutchfield, NRC, to EPRI, discussing i
EPRI-proposed optimization subject on simplification of off-site emergency planning for ALWRs using passive safety systems.
July 2, 1992 Letter for G. Bockhold, EPRI, to NRC, forwarding data-
{
base for issues identified on Chapters 2-10 of-Volume'III of the ALWR Utility Requirements. Document.
July 6, 1992 Letter from G. Bockhold, EPRI, to NRC,-~ forwarding-discussion of term ' decommissioning' as term relates-to
-ALWR Utility Requirements Document.
l f,
July 6, 1992 Letter from D. Crutchfield, NRC, to EPRI, forwarding draft commission paper, " Design Certification and.
i Licensing Policy Issues Pertaining to_ Passive and j
Evolutionary Advanced Light Water Reactor designs."
}
Program Summary
'A-35 T
f
)
i l
I July 17, 1992 Letter from G. Bockhold; EPRI, to NRC, forwarding revised j
data base for' issues on Chapter 9 of Volume-III of.the l
ALWR Utility Requirements Document.
Issues include l-balance-of-plant. fire protection program, independence of -
I ventilation system inside containment, and requirements i
for smoke removal capability.
July 7, 1992 Letter frem T.G. Hiltz, NRC, to EPRI, summarizing meeting with EPRI on June 11, 1992 on generic system j
classification, reliability-based technical a
specifications, and shutdown risk considerations for
(-
ALWRs.
l August 3, 1992 Letter from G. Bockhold, EPRI, to NRC,_ forwarding i
supporting information on charcoal filters per j
teleconference with NRC.
f I
4 i
3 1
4 r
}
i i
i i
i i
?
i 3
L 1
4
?
I
- r i
f Program Summary.
A-36 i
t s
a-
,--e a
r-e v
I 4
1 l
l APPENDIX B REFERENCES g
i Atomic Industrial Forum (AIF), AIF/NESP-020, " Compendium of Cesign Features to Reduce Occupational Radiation Exposure at Nuclear Power Plants."
--, AIF Study, 6.4-12.
f Committee on the Biological Effects of Ionizing Radiation, "The Effect on i
Population of Exposure to Low level of Ionizing Radiation," July 1980 Electric Power Research Institute (EPRI), EPRI NP-309, " Human Factors Review of Nuclear Power Plant Control Room Design."
--, EPRI NP-1081, " Refueling Outage Water Clarity Improvement Study."
j l
--, EPRI NP-1982, " Evaluation of Proposed Control Room Improvements Through Analysis of Critical Operator Actions."
--, EPRI NP-2294, " Guide to Design of Secondary Systems and Their Components To Minimize Oxygen-Induced Corrosion."
--, EPRI NP-2360, " Human Factors Methods for Assessing and Enhancing Power Plant Maintainability."
1
--, EPRI NP-2411, " Human Engineering Guide for Enhancing Nuclear Control l
Room."
--, EPRI NP-2777.
--, EPRI NP-3448, "A Procedure for Reviewing and Improving Power Plant Alarm Systems."
i 4
--, EPRI NP-3659, " Human f actors Guide for Nuclear Power Plant Control Room e
[
Development."
t EPRI NP-3701, " Computer-Generated Display Guidelines" (Volumes 1 and 2),
f l.
--, EPRI NP-3784, "A Survey of the Literature on low-Alloy Steel Fastener f
Corrosion in PWR Power Plants," J. S. Hall, December.1984.
--, EPRI NP-4350, " Human Engineering Design Guidelines for Maintainability." '
l
--, EPRI NP-4762-SR.
--, EPRI NP-4947-SR, "BWR Hydrogen Water Chemistry Guidelines," 1987.
[
r i
j' i
j Program Summary B-1 t
--, EPRI NP-5067, " Good Bolting Practices, A Reference Manual for Nuclear i
Power Plant Maintenance Personnel," Volume 1:
"Large Bolt Manual," 1987 and l
Volume 2:
"Small Bolts and Threaded Fasteners," 1990.
[
--, EPRI NP-5159, " Guidelines for Specifying Integrated Computer-Aided i
Engineered Applications for Electric Power Plants."
--, EPRI NP-5283-SR-A, " Guidelines for Permanent BWR Hydrogen Water Chemistry Installations," September 1987.
--, EPRI NP-5479, " Application Guidelines for Check Valves in Nuclear Power 5
Pl an t s. "
--, EPRI NP-5639, " Guidelines for Piping System Reconciliation (NCIG-05, Revision 1)."
---, EPRI NP-5652, " Guidance for the Utilization of Commercial Grade Items in
[
Nuclear Safety Related Applications (NCIG-07)."
l
--, EPRI NP-5693.
i
--, EPRI NP-5769, " Degradation and Failure of Bolting in Nuclear Power Plants," R. E. Nickell, Principal Investigator, Volumes 1 and 2, April 1988.
--, EPRI NP-5960, "PWR Primary Water Chemistry Guidelines," Revision 1.
i
--, EPRI NP-5989, " Effects of Control-Room Lighting on Operator Performance, A Pilot Empirical Study."
l
--, EPRI NP-6202, " Material Specification for Alloy X-750 in LWR Internal-Components."
l
--, EPRI NP-6209, " Effective Plant labeling and Coding."
[
--, EPRI NP-0316, " Guidelines for Threaded-Fastener Application in Nuclear Power Plants," Looram Engineering, Inc., July 1989.
--, EPRI NP-6433.
--, EPRI NP-6539.
1
--, EPRI NP-6628.
I 5
--, EPRI NP-6748.
--, EPRI NP-7077, Revision 2.
--, EPRI NP-7183-SL, " SHARP 1, A Revised Systematic Human Action Reliability i
Procedure."
--, EPRI RP-2184-7.
--, EPRI RP-2705-7.
j Program Summary B-2
i r
I t
i i
--, EPRI TR-100259, "An Approach to thel Analysis of Operator Actions in i
Probabilistic Risk Assessment," G. W. Parry, et al., Draft, November 1991.
]
Fauske and Associates, Inc., " Process for Evaluating Accident Management Capabilities."
--, " Technical Support for the Hydrogen Control Requirement for the EPRI Advanced Light Water Reactor Requirements Document," June 1988.
Federal Guidelines on Dam Safety.
General Electric Company, NED0-22155, " Generation and Mitigation of I
Combustible Mixtures in Inerted Mark I Containments."
--, NEDO-31643 P (proprietary).
1
--, NED0 31858.
l Illumination Engineering Society, "IES Lighting Handbook."
-l National Research Council, " Estimated Probabilities of Extreme Winds," 1988.
Nuclear Construction Issues Group (NCIG)-07, " Guidance for the Utilization of l
Commercial Grade Items in Nuclear Safety Related Applications."
f
--, NCIG-14 (EPRI NP-6628), " Procedure for Seismic Evaluation and Design of t
i Small Bore Piping," April 1990.
l Nuclear Safety Analysis Center (NSAC), NSAC-39, " Verification and Validation l
for Safety Parameter Display Systems."
.j r
--, NSAC-147, " Losses of Off-Site Power at U.S. Nuclear Power Plants; Through j
1989."
q Nuclear Utilities Management and Resource Council (NUMARC), NUMARC-87-00.
I
--, NUMARC Containment Integrity Working Group Report, February 1988.
- -, " Process for Evaluating Accident Management Capabilities."
INDUSTRY CODES AND STANDARDS l
American Concrete Institute (ACI), 318, " Building Code Requirements for:
l
[
Reinforced Concrete."
_t l
l.
American National Standards Institute (ANSI),10.4, " Guidelines for the i
L Verification and Validation of Scientific _ and Engineering Computer _ Programs.
for the Nuclear Industry," 1987.
i
--, 35.1.
l
--, A13.1-1981 (Reaffirmed 1985), " Scheme _ for the Identification of Piping-I Systems."
i Program Summary B-3
[
_._ _ _.. _ ~_ _
i i
f 1
i i,
l A58.1-1982, " Minimum Design Loadings for Buildings and Other Structures."
4 C96.1.
f 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 Nuclear Power Plants."
--, N45.2.2, " Packaging, Shipping, Receiving, Storage, and Handling of Items for Nuclear Power Plants."
I
--, N45.2.3, " Housekeeping During the Construction Phase of Nuclear Power i
Pl ant s. "
--, N45.4-1972, " Leakage Rate Testing of Containment Structures for Nuclear j
Reactors."
)
--, N101.2-1980, " Protective Coatings (Points) for Light Water Nuclear Reactor Containment Facilities," 1980, i
j
--, N101.4-1972, " Quality Assurar.e for Protective Coatings Applied to j
Nuclear Facilities."
i
--, c35.1-1972, " Accident Prevention Signs, Specification for" 1
l
--, Z86.1.
i f
" Leak Rate Testing of the Containmeat Structures for Nuclear Reactors."
American National Standards Institute /Amt ican Concrete Institute (ANJI/ACl),
j 349, " Code Requirements for Nuclear Safety-Related Structures."
American National Standards Institute /American Institute of Steel Construction (ANSI /AISC), N-690, " Specification for the Design, fabrication, and Erection i
of Steel Safety-Related Structures for Nuclear facilities," Chicago, Illinois.
i j
American National Standards Institute /American Nuclear Society (ANSI /ANS),
j 2.3, " Standard for Estimating Tornado and Extreme Wind Characteristics at
[
Nuclear Power Sites," 1980.
--, 2.5-1984, " Standard for Determining Meteorological Information at Nuclear i
Power Sites."
I i
--, 2.8-1981, " Standard for Determining Design Basis Flooding at Power Reactor Sites."
1
--, 2.12, "American Nuclear Society Guidelines for Combining Natural and Man-l Made Hazards at Power Reactor-Sites."
i
--, 3.1, " Selection, Qualification and Training for Nuclear Power Plants."
l
--, 3.3-1988.
L 1-53-4
. Program Summary _
i
--,18.1, " American National Standard Radiation Source Term for Normal Operation of Light Water Reactors."
--, 51.1, " Nuclear Safety Criteria for the Design of Stationary PWR Plants."
---, 52.1, " Nuclear Safety Criteria for the Design of Stationary BWR Plants."
--, 55.1, " Solid Radioactive Waste Processing Systems for Light Water Cooled Reactor Plants."
--, 55.4, " Gaseous Radioactive Waste Processing Systems for Light Water Reactor Plants."
--, 55.6, " Liquid Radioactive Waste Processing Systems for Light Water Reactor Plants."
--, 56.2, "Containmer.t Isolation Provisions for Fluid Systems After a LOCA,"
1976 and 1984.
-, 56.7.
--, 56.8, " Containment System Leakage Testing Requirements," 1987 3
--, 57.1, " Design Requirements for Light Water Reactor Fuel Handling Systems."
--, 57.2,
- Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants."
--, 57.3, " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants."
--, 58.1, " Plant Design Against Missiles."
--, 58.2, " Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Pipe Rupture."
--, 58.8-1984, " Time Response Design Criteria for Safety Related Operator Actions at Nuclear Power Plants."
--, 58.9, " Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems."
--, 59.2-1985, "Safcty Criteria fer HVAC Systems Located Outside Primary Containment."
American National Standards Institute /American Society of Mechanical Engineers (ANSI /ASME), 3.3-1988.
--, AG-1, " Code on Nuclear Air and Gas Treatment," 1089.
--, B31.1, " Power Piping."
Program Summary B-5
1
-. ~ _
i 4f-1 I
g-1
--, B.31.1, Appendix 2, "Non-Mandatory Rules for the Design of Safety Valve L
Installations."
--, N509, " Nuclear Power Plant Air Cleaning Units and Components,"' 1989.
]
--, N510, " Testing of Nuclear Air-Cleaning System," 1989.
1 i-
---, N0G-1-1983, " Rules for Construction of Overhead and Gantry Cranes (Top j
Running Bridge, Multiple Girder)."
s i
--, NQA-1.
i
, NQA-2.
l
--, OM-6, " Inservice Testing of Pumns."
4 l'
--, OM-10, " Inservice Testing of Valves."
I' American National Standards Institute / Institute of Electrical and Electronics -
s l
Engineers'(ANSI /IEEE), 387, "IEEE Standard Criteria for Diesel Generator Units l
Applied as Standby Power Supplies for Nuclear Power Generating Stations."
--,.30, "Sof tware Quality Assurance Plans."
1 I
--, 828.
l
--, 829, " Software Test Documentation."
i
--, 982.1-1988, "IEEE Standard Dictionary of Measures to Produce Reliable Software."
i' l
--, 982.2-1988, "!EEE Guide for the Use of IEEE Standard Dictionary of l
Measures To Produce Reliable Software."
i j
--, ;012-1986, "IEEE Standard for Software V&V Plan,"
(_
--, 1042.
i p
--,1063-1987, " Standard for' Sof tware Users Documentation."
}l
--, ANS-7-4.3.2-1982, " Application Criteria for Programmable: Digital Computer-l Systems in Safety Systems of Nuclear Power Generating Stations."
i American-Nuclear Society (ANS),-5.1, " Decay Heat Power in Light Water Reactors," La Grange Park, Illinois, October 1975 and October 1979.
I
--,-18.2-1973, " Nuclear Safety Criteria for the Design.of Stationary i.
Pressurized Water Reactor Plants."
American Society for Testing and Materials (ASTM), A262 Practice E,' Modified l
Strauss Test.
l
--, A 708 Strauss Test.
(
(
Program Summary B-6 l-
1 i
)
A 800.
l
--, D 3803, " Standard Test Methods for Radiological Testing of Nuclear-Grade j
Gas-Phase Absorbeni.s."
l i
--, D 3842.80.
--, E-185-82, " Standard Recommended Practices for Surveillance Tests for Nuclear Reactor Vessels," 1982.
--, E-813.
American Society of Civil Engineers (ASCE), 4-1986, " Seismic Analysis of i
Safety-Related Nuclear Structures and Co nentary on Standard for Seismic Analysis of Safety-Related Nuclear Structures," September 1986.
--, 7-1988 (formerly ANSI A58.1).
l i
American Society of Mechanical Engineers,-80iler and Pressure Vessel-Code I
(ASME Code),Section III, " Nuclear Power Plant Components."
l
--,Section III, Appendix II, Paragraph 11-1430.
{
t
--,Section III, Appendix N.
I
--,Section III, Division 1, " Nuclear Power P_lant Components, with l
Appendices."
l e
--,Section III, Division 2, " Code for Concrete Reactor Vessels and Containments."
--,Section III, Subsection CC 3720.
--,Section III, Subsection NB/NC/ND-1100(a).
i
--,Section III, Subsection NCA-ll40.
l
--,Section III, Subsection NF, "Coinponent Supports."
7 l
--,Section III, Subsection NG, " Core Support Structures."
f l
--,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Comoonents."
L
--,Section XI, Subsection IWV 3421-3427(a).
I
.i Code Case N-397.
_--, Code Case N-411, "Al.ternative Damping Values for Seismic Analysis of
-r Classes 1, 2,-and 3 Piping Sections."
-3
--, Code Case N-420, " Linear Energy' Absorbing Supports for Subsection NF, l
Classes 1, 2, and 3 Construction,Section III, Division _l."
f l
Program Summary B-7 i
a
--, Code Case N-451.
--, Code Case N-462.
--, NCA-ll40.
--, NQA-2A, Part 2.7, " Quality Assurance Requirements cf Computer Software for Nuclear Facility Applications."
--, NQA 2.7.
--, PTC 6, " Steam Turbines Performance Test Code,"
--, PTC 6.1, " Alternative Procedure for Testing Steam Turbine."
--, TDP-2.
Institute of Electrical and Electronics Engineers (IEEE), 279, " Criteria for Protection Systems for_ Nuclear Power Generating Stations."
--, 308-1980, " Criteria for Class IE Power Systems-for-Nuclear Power Generating Stations."
--, 323-1974, "lEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations."
--, 334, " Standard for. Type _ Tests of Continuous Duty Class-lE Motors for Nuclear Power Generating Stations."
--, 338-1977, " Criteria for the Periodic Testing of Nuclear Power Generating Station Safety Systems."
--, 344-1987, " Recommended Practice., for Seismic Qualification of Class -lE-l Equipment for Nuclear Power Generating Stations,"
!~
--, 352, " Guide for General Principles of Reliability Analysis of Nuclear Power Generating Station Protection Systems."
i.
l.
--,.383, " Type Test of Class IE Electric Cables, Field-Splices, -and Connections for Nuclear Power Generating Stations."
--, 384, "IEEE Trial-Use Standard criteria _ for Separation of Class IE
(
Equipment and Circuits," 1974.
, 472-1974.
---, 484.
l'
--, 519.
i l-
--, 577, " Requirements for Reliability Analysis in the Design and Operation of Safety: Systems for Nuclear Power Generating. Stations."
i i
Program Summary B-8 I
i
--, 603, " Trial-Use Standard Criteria for Safety Systems for Nuclear Power i
Generating Stations."
l
)
-, 741, "lEEE Standard Criteria for the Protection of Class IE Power Systems and Equipment in Nuclear Power Generating Stations."
-j
--, 765-1983, "IEEE Standard for Preferred Power Supply for Nuclear Power Generating Stations."
=
i
--, 828-1983, "IEEE Standard for Software Configuration Management Plans."
--, 981.1.
--, 981.2.
t I
--, 982.2-1988, "IEEE Guide for the Us's of IEEE Standard Dictionary of Measures To Produce Reliable Software."
--, 1008-1987, "IEEE Standard for Software Unit Testing."
l
--, 1012.
--, 1042.
l t
--, 1050-1989, "IEEE Guide for Instrumentation and Control Grounding-in Generating Stations."
--, C37-90.1-1989.
--, C62.41-1980, "!EEE Guide for Surge Voltages in Low-Voltage AC Power
[
Circuits."
i
--, P1023/05, " Guide for the-Application of Human f actors Engineering to l
Systems, Equipment and Facilities of Nuclear Power Generating Stations."
j Instrument Society of America (ISA), 67.15, Draft-RP 67.04, Part II,
[
" Methodology for the Determination of Setpoints for Nuclear-Safety-Related i
Instrumentation."
Insulated Power Cable Engineers Association (IPCEA), 5-61-402.
International Electrotechnical Commission (IEC), 880-1986, " Software for Computers in the Safety Systems of Nuclear Power _ Stations."
l l
National P ectrical Code.
l t
National Fi.e Protection Assnciation-(NFPA),13, " Standard for-the Instal-1 lation of Sprinkler Systems."
l
--, 70.
--, 90A-1989, " installation of Air Conditioning and Ventilating Systems."
--, 101, " Life Safety Code."
{
l Program Summary B-9 1.
--, 803, " Fire Protection for Light Water Nuclear Power Plants."
Tubular Heat Exchangers Manufacturers Association (THEMA), T-2.41.
Uniform Building Code, i
l U.S. Department of Defense (D0D), D0D-MIL-HDBK-217E, " Reliability Prediction of Electronic Equipment."
j
--, D0D-Mll-HDBK-263.
--, D00-MIL-HDBK-338,'" Electronic Reliability Design Handbook."
--, D00-Mll-HDBK-472, " Maintainability Prediction."
--, 000-STD-781, " Reliability Test-Methods, Plans, and Environment of Engineering Development, Qualification, and Production."
--, D00-STD-1399.
--, 000-STD-1629A, " Procedures for Performing a Failure Modes Effects and Criticality Analysis."
Program Summary B-10
f f
1 I
i I
i APPENDIX C LIST OF ACBREVIATIONS The following is a list of abbreviations-used throughout this report and the DSER.
r ac alternating current AAC alternate ac ABWR advanced boiling water reactor i
ACI American Concrete Institute ACRS Advisory Committee on Reactor Safeguards A/D analog to digital 1
ADS automatic depressurization system
[
AEOD Office for Analysis and Evaluation of Operational Data l
AHU air handling unit AIF Atomic Industrial Forum, Inc.
i AISC American Institute of Steel Construction-ALARA as low as is reasonably achievable
.i ALWR advanced light water reactor ANS American Nuclear Society i
ANSI American National Standards Institute A00 anticipated operational occurrence i
ARSAP Advanced Reactor Severe Accident Program i
ASCE Avierican Society of Civil Engineers ASD adjustable speed drive ASME American Society of Mechanical Engineers i
ASTM American Society for Testing and Materials ATWS anticipated transient (s) without-scram AWS American Welding Society BE best estimate BEIR Biological Effects.of Ionizing Radiation, Committee on the B0P balance of plant BTP branch technical position d&W Babcock and Wil " x BWG British Wire Gauge BWR boiling water reactor BWROG Boiling Water Reactor Owners Group l
C-1
' Category I C-Il Category 11
?
CAE computer-aided engineering L
CAGS compressed air and gas system CAS central alarm system CCDF tomplementary cumulative distribution function CCFP conditional containment failure. probability 1
CCIC core coolant inver. tory control CCTV closed-circuit television l
-Program Summary.
C-1
CCW component cooling water-CCWS component cooling water system CDF-core damage' frequency CDWS chilled water system CE Combustion Engineering, Inc.
CET comtainment event tret CFR Code of Federal Reaulations ClV containment isolation valve CMEB former NRC Office of Nuclear Reactor Regulation Chemical Engineering Branch CMT core makeup tank COL combined construction and operating license CPU control processing unit CRA centrol rod assembly CRD control rod drive CRDM control rod drive mechanism CRGR Committee to Review Generic Requirements CS containment spray CSS containment st 'y system CT combustion turvine CTSI condt sate treatment systems influent CVC chemical and volumn control CVCS chemical and volume control system CWS circulating water system D/A digital to analog DBA design-basis accident DBT design-basis tornado DC design certification DC direct current DCH direct containment heating DDACS digital data acquisition and control system DEGB double-ended guillotine break DHR decay heat removal DNBR departure from nucleate boiling ratio D00 Department of Defense-D-RAP design reliability assurance program DS drywell spray DSER
_ draft cafety_ evaluation-report DWST domineralized water storage tank ECCS emergency core cooli'tig system ECWS essential chilled water system EDG emergency diesel generator EFU.
_emer_ ency filter unit q
EFW emergency feedwater EFWS emergency feedwater system EFWST emergency feedwater storage tank EMI.
electromagnetic interference-EMS
. environmental monitoring system E0F emergency operations-facility E0P emergency operatir.g proce6ure EPA electric protective assembly-EPA Environmental Protection Agency-Program Summary
-C-2
1 i-EPG emergency procedures guideline EPRI Electric Power Research Institute ERF emergency response facility ESF engineered safety feature (s) l ESFAS engineered safety feature actuation system
.j ESI Energy Systems Group i
EbW essential service water i
ESWS essential service water system FDA final design approval FIVE fire vunerability evaluation FN ferrite number i
FPCCS fuel pool cooling and cleanup system j
FPLC fission product leakage control FPLCS
' fission product leakage control system FPS fire protection system FSER final safety evaluation report j
GDC general design criterion (a)
GE General Electric Company j
GI generic-issue GIMCS generic issues management control system GIP Generic Implementation Procedure
{
GL generic letter i
GPM gallon (s) per minute r
GRWPS gaseous radioactive waste processing system GSI generic safety issue j
HCLPF high confidence / low probability of failure
'i HEPA high-efficiency particulate air HF human factors i
HPCI high-pressure coolant injection l
HPI high-pressure injection
~l HRA human reliability analysis HVAC heating, ventilating, and air conditioning i
HWC hydrogen water chemistry l
IASCC irradiation-assisted stress corrosion cracking l
I&C instrumentation and controls ICC inadequate core cooling l
ICS integrated control system ICST influcrt to condensate storage tank i
IDCOR Industry Degraded Core Rulemaking IE Office of Inspection and Enforcement-l I&E Office of Inspection and Enforcement f!
IEB Office of Inspection and Enforcement bulletin IEC International Electrotechnical Commission IEEE Institute of Electrical and Electronics Engineers 3
IES Illumination Engineering Society i
IGA intergranular attack IGSCC intergranular stress corrosion cracking i
IIT int.ident investigation team l
ILRT integrated leak rate test IMS information management system j
Program Summary C-3
-[
L
.. ~. _ _. -.. - -. - - - _ -
1 i
3 i
l IN NRC information notice it INPO Institute of Nuclear Power Operations IPCEA Insulated Power Cable 'i.ngineers Association i
IPE individual plant evaluation 1
IRWST in-containment refueling water storage tank j
ISI inservice inspection ISLOCA intersystem loss-of-coolant accident IST inservice testing _
l ITAAC inspections, tests, analyses, and accaptance criteria 5
LBHS large-bore hydraulic snubber i
LC0 limiting co.idition(s) for operation l-LCS leakage contrel system l.
LCS local cr'rol station LDD licen*.ng design basis LDR load aefinition report i
LLNL Lawrence Livermore National' Laboratory j
LOCA loss-of-coolant accident i
i L0cT loss-of-fluid test j
LOOP loss of offsite power LRB licensing review basis i
LRFD load and resistance factor design i
LRWPS liquid radioactive waste processing system LTOP low-temperature overpressure protecticn LWR
_ light water reactor l
- MCC motor control center MAAP modular accident analysis program MCPR minimum critical power ratio l
MCR main control room l
MF moderate frequency MFW main feedwater
{
d-MI man-machine interface M-MIS man-machine interface system (s)
MOV motor-operator valve i-MPA multiplant action MSIV main steam isolation valve MSIVLCS main stemn isolation valve leakage control system 1
MTC moderator temperature _ coefficient i
MWD /MTU megawatt-day (s) per metric ton of uranium
_MWSE makeup water systems effluent MWSG makeup water to steam generator j
MWST makeup water storage tank-i-
j NCC natural convection cooldown 4
NCIG Nuclear Construction issues Group i
-- NDE nondestructive examination:
NOT.
nil ductility temperature e
NECWS nonessential chilled water system j
NESWS noneseential service water system j-NFPA National Fire Protection Association
{
NNS n Enuclear safety
{
Program Summary C-4 1
5.
l NPHS normal power heat sink l
NPRDS nuclear power plant reliability data system l
NPSH net positive suction head i
NRC V.S. Nuclear Regulatory Commission NS non-seismic NSAC Nuclear Safety Analysis Center
+
I NSSS nuclear steam supply system NUMARC Nuclear Utility Management and Resources Council N/VT neutron (s)/ square meter t
OBE operating-basis eartnquake i
DDCM offsite dose calculation manual 0-RAP operations reliability assurance program OSHA Occupational Safety and Health Administration j
PAP personnel access portal
{
PASS postaccident sampling. system i
PCCS passive containment cooling system PDHR passive decay heat removal l
PDHRS passive decay heat removal system PGA peak ground acceleration PGC power generation complex 3
PGP procedures generation package PIN project information network r
P!V pressure isolation valve l
PM preventive maintenance j
PMF probable maximum flood PMP probable maximum precipitation
{
PORV power-operated relief _ valve PRA probabilistic risk assessment l
PSD power spectrum density l
l PSF performance shaping factor l
PSIS passive safety injection system PVC polyvinylchloride I
PWR pressurized water reactor i
PWSCC primary water stress corrosinn cracking t
QA quality assurance
[
t RAI request for additional information RAM random-access memory i
RAP reliability assurance prooram j
RCA radiological control area-t RCIC reactor core isolat1.. cooling 3
RCP reactor coolant oump
.RCPB reactor coolari pressure coundary
)
RCS reactor coolant system RFI radi Jrequency interference l
RFPY reactor full-power year RG regulatory guide l
RHR residual teat removal s
RIP reactor idernal pump RISCC radiation-induced stress corrosion cracking ROM read-only w mory-r ogram Summary C-5 r
t RPS rer-tor proteuton system RPV reactor pressure vessel
[
RSDC reactor shutdown cooling i
RIO reactor trip breaker RID resistance temperature detector
}
R1 nil ductility temperature RTIm reactor trip system RV reactor vessel
[
RVLIS reactor vessel level instrumentation system l
RWCS reactor water cleanup system i
[
i SAf0L specified acceptable fuel design limit SAR safety analysis report SAS secondary alarm station SBWR simplified boiling water reactor SCA sneak circuit analysis SCC stress corrosion cracking i
SDV safety depressurization and vent i
SDV scram discharge volume-l SDVS safety depressurization and vent system j
SECY NRC Office of the Secretary (of the Commission) t SEP Systematic Evaluation Program SER safety evaluation report
{
SFA single-f ailure analysis i
SFC single-failure criterion j
SG steam generator SGI safeguards information l
SG0F steam generator overfill 1
SGTR steam generator tube rupture
[
SGTS safety gas treatment system i
SHARP systematic human action reliability procedure SI safety injection SIS safety injection system SLC standby liquid control l
SLCS standby liquid control system i
SMB safety margin basis
}
SOER significant operating event report SPDS safety parameter display system i
SQAP Sofware Quality Assurance Program i
SQUG Seismic Qualification Utilities Group i
SRM staff requirements memorandum (a) l SROA safety-related operator action i
- SRP Standard Review Plan (NUREG-0800)
I SRV safety / relief valve l
SSAR standard safety analysis report j
SSC structures, systems, and components
~
SSE safe-shutdown earthquake-
-SSI soil-structure interaction SSW safety service water 51CP.
Source Term Code Package r
l SWAP Service Water Assistance Program SWS service water system
[
i 1
Program Summary C !
.__c__,
I
TBCCWS turbine building component cooling water system j
TDI Transamerica Delaval, Inc.
i THERP technique for human error rate prediction l
TID technical information document TIP traversing in-core probe TMI Three Mile Island THI-2 Three_ Mile Island Unit 2 TS technical specification (s)
ISC technical support center UBC Uniform Building Code i
UHS ultimate heat sink UPS uninterruptible power supply URS ultimate rupture strength 051 unresolved safety issue UT ultrasonic test i
V volt V&V vertfication and validation
}
WS wetwell spray i
i I
e t
t i;
I t
e P
i
-I
- Program Summary
'C-7
... - - =
i i
)
I APPENDIX D PRINCIPAL CONTRIBUTORS NRC Personnel Review Area H. Ashar structural engineering R. Borchardt project management B. Bordinick legal counsel J. Brammer mechanical. engineering K. Campe probablistic risk assessment f
l C. Carpenter reliability assurance l
P.
Caruso systems l
T. Chandrasekaran plant systems T. Cheng structural engineering M. Chiramal instrumentation and controls O. Chopra electrical systems R. Correia human factors, reliability assurance-R. Dube physical security R. Eckenrode human factors A. El-Bassioni probablistic risk assessment F. Eltawila research (plant systems) i T. Essig radiation protection E. Fox emergency planning J. Gallagher instrumentation and controls R. Gallo operator licensing G. Georgiev materials engineering C. G~dman human factors J. Guo plant systems l
B. Hardin research l
C. liinson radiation protection l
T. Hiltz project management l
G. Hsii reactor systems J. Joyce
-instrumentation and controls F. Kantor emergency planning G. Kelly.
-probablistic risk assessment S. Kim structural engineering
~T.
Kim project. management L. Kopp reactor ~ systems J. Kudrick plant systems
-J. Lazevnick electrical systems
.E.
Lee mechanical engineeering J. Lee radiation protection S.: Lee structural engineering A. Levin reactor systems J. Levine meteorology C. Li plant systems Y. C. Li mechanical engineering J. Lyons.
plant systems
-M. Malloy project management J. Martin radiation protection Program Summary D-1 1
.. - -. = -.
-.=. a.
- -.,...- ~.
E. McKenna quality assurance B. Mendelsohn physical-security J. Monninger plant systems J. Moore legal counsel D. Notley plant systems R. Palla probablistic risk assessment K. Parczewski chemical engineering L. Phillips reactor systems R. Pichumani structural engineering j
T. Pohida instrumentation and controls T. Polich reliability assurance
- f. Rinaldi structural ergineering j
R. Rothman geosciences M. Rubin reactor systems G. Schwenk reactor systems J. Sharkey reliability assurance P. Shea licensing assistance L. Shotkin research (plant systems)
B. Siegel project management
- f. Skopec radiation protection D. Smith human factors P. Sobel geosciences L. Soffer research-(source term)
- 1. Spickler radiation protection J. Spraul quality assurance J. Stewart instrumentation and controls D. Terao mechanical engineering D. Thatcher electrical systems G. Thomas reactor systems C. Tinkler research J. Tsao material engineering M. Waterman instrumentation and controls J. Wigginton radiation protection F. Witt chemical engineering R. Woods research-(generic safety issues)
P. Worthington research (plant systems)
Program Summary D-2
APPENDIX E COMMISSION PAPERS APPLICABLE TO ADVANCED LIGHT WATER REACTORS SECY-77-439, " Single failure Criterion," August 17, 1977.
SECY-86-228. " Introduction of Realistic Source Term Estimates into Licensing,"
August 6, 1986.
SECY-88-147, " Integration Plan for Closure of Severe Accident Isues," May 25, 1988.
SECY-88-203, " Key Licensing issues Associated With DOE-Sponsored Advanced Reactor Designs," July 15, 1988.
SECY-89-012, " Staff Plans for Accident Management Regulatory and Research Programs," January 18, 1989.
SECY-89-013. " Design Requirements Related to the Evolutionary Advanced Light Water Reactors (ALWRs)," January 19, 1989.
SECY-89-153, " Severe Accident Design Features of the Advanced Boiling Water Rcactor (ABWR) " May 10, 1989.
SECY-89-228, " Draft Safety Evaluation Report _on Chapter 5 of the Advanced Light Water Reactor Requirements Document," July 28, 1989.
SECY-89-341, " Updated Light Water Reactor (LWR) Source Term Methodology and Potential Regulatory Applications," November 6, 1989.
SECY-90-016, " Evolutionary Light Water Reactor _(LWR)_ Certification _lssues and Their Relationship to Current Regulatory Requirements," January 12, 1990.
SECY-90-241, " Level of_ Detail Required for Design Certification Under Part 52," July 11, 1990.
SECY-90-307, " Impacts of Source Term Timing on NRC Regulatory Positions,"
August 30, 1990.
SECY-90-313, " Status of Accident Management Programs and Plans for Implementa-tion," September 5, 1990.
SECY-90-329, " Comparison of the General Electric Advanced Boiling Water Reactor (ABWR) Design and the Electric Power Research Institute's (EPRI's)
Advanced Light Water Reactor (ALWR) Requirements Document," September 20, 1990.
7 SECY-90-341, " Staff Study on Source Term Update and Decoupling' Siting from Design," October 4,=1990.-
Program Summary E-1 1
l t
i l
SECY-90-353 " Licensing Review Basis for the Combustion Engineering, Inc.,
System 804 Evolutionary Light Water Reactor." October 12, 1990, t
SECY-90-377, " Requirements for Design Certification Under 10 7 Part 52,"
l November 8, 1990.
l SECY-90-406, " Quarterly Report on Emerging Technical Concerns," December 17, 1990.
SECY-91-074, " Prototype Decisions for Advanced Reactor Designs," March 19, 1991.
SECY-91-078, " Chapter 11 of the Electric Power Research Institute's (LPRI's)
Requirements Document and Additional Evolutionary Light Water Reactor (LWR)
Certification Issues," March 25, 1991.
SECY-91-135, " Conclusions of the Probabilistic Seismic Hazard Studies Conducted for Nuclear Power Plants in the Eastern United States," May 14, 1991.
SECY-91-178, " Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications and Combined Licenses," June 12, 1991.
i SECY-91-210. " Inspections, Tests, Analyses, and Acceptance Criteria (11AAC)
Requirements for Design Review and Issuance of a Final Design Approv d '
July 16, 1991.
SECY-91-229, " Severe Accident Mitigation Design Alternatives for Certified Standard Designs," July 31, 1991, SECY-91-262, " Resolution of Selected Technical and Seveie Accident issues for i
Evolutionary Light Water Reactor (LWR) Designs," August 16, 1991.
(
SECY-91-272, " Role of Personnel and Advanced Control Rooms in Future Nuclear
~
Power Plants," August 27, 1991.
SECY-91-273, " Review of the Vendor's Test Programs To Support the Design certification of-Pastive Light Water Reactors," August 27, 1991.
l l
SECY-91-292, " Digital Computer Systems for Advanced Light Wat r Reactors,"
l September 16, 1991.
SECY-91-348, " Issuance of Final Revision to Appendix J._to 10 CFR 50, and Reslated Final Regulatory Guide 1.XXX (MS 021-5)," October 25, 1991.
SECY-92-030, " Integral System Testing Requirements for Westinghouse's AP600 Plant," January 27, 1992.
j SECY-92-037, "Need for NRC-Sponsored Confirmatory Integi?al-System Testing of-the Westinghouse AP600 Design," January 31, 1992.
SECY-92-053,'"Use of Design Acceptance Criteria During 10 CFR D rt 52 Design Certification Reviews," February 19, 1992.
Program Summary E-2 l
l
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SECY-92-092, "The Containment Performance Goal. External Events Sequences, and the Definition of Containment f ailure for Advanced Light Water Reactors,"
March 17, 1992.
SECY-92-120 "NRC Staff Raviews for the Westinghouse AP600 and the General Electric Simplified Boiling Water Reactor (SBWR) Designs," April 7, 1992.
SECY-92-127, " Revised Accident Source Terms for Light Water Nuclear Power Plants," April 10, 1992.
SECY-92-133, " Draft Safety Evaluation Reports for Volume I and Volume til of the Electric Power Research Institute's Advanced Light Water Reactor Requirements Document," April 14, 1992.
SECY-92-134, "NRC Construction Inspection Program for Evolutionary and Advanced Reactors under 10 CFR Part 52 " April 15, 1992.
SECY-92-170, "Rulemaking Procedures for Design Certification," May 8, 1992.
SECY-92-172, " final Safety Evaluation Report for Volume 11 of the Electric Power Research Institute's Advanced Light Water Reactor Requirements Document," May 12, 1992.
SECY-92-211. "NRC Confirmatory integral System Testing for the General Electric SBWR Design," June 5, 1992.
SECY-92-214, " Development of Inspections Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications," June 11, 1992.
SECY-92-219, "NRC-Sponsored Confirmatory Testing of the Westinghouse AP600 -
Design," June 16, 1992.
4 l'
I i
i f
Program Summary E-3 l
I APPENDIX F REPORT BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS h
a
(
l l
t
. Program Summary
.---.;.--..--.-.---.;i
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UNITED STATES l'
F NUCLEAR REGUL ATORN COMMISSION i
7 ) t.,i 3#,
ADVISORY COMMIT TEE ON HE ACTOR SAFEGUARDS t
s W A&HING T ON, D. C. 20%S
"'99.....,0 August 18. 1992 The lionorable Ivan Selin Chairman U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Chairman S:
lin'
SUBJECT:
ELECTRIC POWER RESEARCH I!JSTITUTE ADVANCED LIGitT WATER REACTOR UTILITY REQUIREMEll'1 S DOCUMENT VOLUME II, EVOLUTIO!1ARY PLANTS During the 387th and 388th meetings of the Advisory Committee on Reactor Safeguards, Ju)y 9-11 and August 6-8, 1992, we reviewed the NRC ataff's Final Safety Evaluation Report (FSER) for Volume II of the Electric Power Research % Otitute's (EPRI) Advanced Light Water Reactor (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 liRC staf f and EPRI. We also had the benefit of the documents referenced.
In the early 1980s, EPRI catablished the ALWR program to, support the United States utility industry efforts to ensure a viable nuclear power generation option for the 1990s and Deyond.
The objective of the program was to ensure that future nucicar 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 objectivo, and addressing the policiis 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.
The URD consists of three volumes:
Volume I, "ALWR Policy and Summary of Top-Tier Requirements,"
e is a management-level synopsis of the URD, including the design objectives and philosophy, the overall physical configuration and features of a future nucicar plant design, and the steps neeussary to take the proposed ALER design criteria beyond the conceptual design state to a completed, functioning power plar.t.
Program Summary F-1
The Honorable Ivan Selin 2
August 18, 1992 e
Volume II, "ALWR Evolutionary Plant," consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant (approximately 1350 Hwe).
Volume III, "ALWR 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 of f ered suggestions regarding safety improvaments 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 staf f's review of the URD was conducted as described in NUREG-1197.
As noted therein, the staff used HUREG-0800, " Standard Review Plan (SRP) for the Review of Safety Analysis Renorts for Nuclear Power Plant." for review guidance.
In addition, the staff's review r(14ccia 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 EPHI 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 staf f conducted its review with the understanding that EPRI design criteria would meet all current regulations, except where deviations were ide.itified.
The staff's review of the URD focused primarily on determining whether the EPRI critoria did or did not conflict with current regulatory requirements.
In its review of Volume II 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 categoriest (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 draf t Commission paper, " Issues Pertaining to Evolutionary and passive Light Water Reactors and Their Relationship to Current Regulatory Requirements" that was issued on February 27, 1992.
We provided our recommendations on the open policy issues portaining 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
i I
The Honorable Ivan Selin 3
August 18, 1992 j-i l
The staff has handled the remaining 410 open issues which were I
identified in the FSER for Volume II by classifying them as " Vendor or Utility Specific Items" which aust be satisfactorily addressed during the staff's review of a
vendor-or utility-specific application.
The staff plans to issue a supplement to the FSER
[
af ter 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 l
generically, and to include in a supplement any which are resolved.
(
We recommend generic resolution of as many of these issues as possible.
I l
We commend EPRI for developing a comprehensive set of requirements.
l These will aid in the design of nuclear plants which will be safer, simpler, more robust, and more easily operated and maintained.
l i
We commend the NRC staff for a very thorough review of the EPRT ALWR Evolutionary URD, and its work with EPRI to identify and resolve many issues relevant to licensing future LWRs.
Wo i
recognize the NRC staff's position that its review necessarily is incomplete.
i sincerely, j
David A. Ward h
Chairman
[
References:
J 1.
SECY-92-172, dated May 12,
- 1992, from James M.
- Taylor, i
Executive Director for Operations, for the Commissioners, i
Subject:
Final Safety Evaluation Report for Volume II of the Electric Power Research Institute's Advanced Light Water
?
Reactor Requirements
- Document, including the following
{
cnclosurest Draft Safety Evaluation Report for Volume I,
" Program i
Summary of the NRC Review of the Electric Power Research Institute's Advanced Light Water Reactor Utility i
Requirements Document," prepared by the office of Nuclear i
Reactor Regulation, U.S. Nuclear Regulatory Commission, dated May 1992 Safety Evaluation Report for Volume II, "NRC Review of the Electric power Research Institute's Advanced -Light i
Water Reactor Utility Requirements Document for Evolutionary plant Designs," prepared by the Office of l
Nuclear Reactor Regulation, U.S.
Nuclear Regnlatory Commission, dated'May 1992 i
f Program Summary F-3 i
~,
.. - ~
J i
)
l The lionorablo Ivan Selin 4
August 18, 1992 i
i 2.
Advanced Light Water Reactor Utility Requirements Document, Volume II, "ALWR Evolutionary Plant," Chapters 1-13, through I-Revision 4,
dated April 1992, - Prepared for Electric Power i
Research Institute 1
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NRC Review of Electric Power Research Institute's l
Advanced Light Water Reactor Utility Requirements Document i
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Program Summary August 1992 l
4 f IN OH LH AN T NU"Si R i
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Same as above.
10 SUPPLEME NT ARY NOTES
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Pro ject Number 669 i
M AB5T R ACI um.we or em/
The staff of the U.S. Nuclear Regulatory Commission has orepared Volume 1 of a safety evaluation report (SER), "NRC Review of Electric Power Research Institute's Advanced l
Light Water Reactor Utility Requirements Document - Program Summary," to document the I
results of its review of the Electr:c Power Research Insti tute's "AdvancM Light Water
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