Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design Appendices A -c U- A U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1994 AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most. documents cited in NRC publications will be available from one of the following sources: 1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2. The Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, DC 20402-9328 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publica- tions, it is not intended to be exhaustive. 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Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, for use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American Na- tional Standards Institute, 1430 Broadway, New York, NY 10018. NUTREG-1 503 Vol. 2 Final Safety Evaluation Report Related to the Certification *of the Advanced Boiling Water Reactor Design Appendices Manuscript Completed: July 1994 Date Published: July 1994 Associate Directorate for Advanced Reactors and License Renewal Office,' of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 I xl ABSTRACT This safety evaluation report (SER) documents the ABWR design features that are substantially the same as technical review of the U.S. Advanced Boiling Water those previously considered. The SERs for the other BWR . eacI&POftor(AIIWR) standard design by the U.S. Nuclear designs have been published and are available for public Wegulatory Commission (NRC) staff. The application for inspection at the NRC Public Document Room, the ABWR design was initially submitted by the General 2120 L Street, N.W., Washington, D.C. 20037. Unique Electric Company, now GE Nuclear Energy (GE), in features of the ABWR design include internal recirculation accordance with the procedures of Appendix 0 of Part 50 pumps, fine-motion control rod drives, microprocessor- of Title 10 of the Code of Federal Reoulations (10 CFR based digital logic and control systems, and digital safety Part 50). Later GE requested that its application be systems. considered as an application for design approval and subsequent design certification pursuant to On the basis of its evaluation and independent analyses, the 10 CFR § 52.45. NRC staff concludes that, subject to satisfactory resolution of the confirmatory items identified in Section 1.8 of this SER, GE's application for design certification meets the The U.S. ABWR design is similar to the international requirements of Subpart B of 10 CFR Part 52 that are AB3WR design, which was being built at the Kashiwazaki applicable and technically relevant to the U.S. ABWR Kariwa Nuclear Power Generation Station, at the time of standard design. A copy of the report by the Advisory the staff's review, by the Tokyo Electric Power Company, Committee on Reactor Safeguards required by Inc. The ABWR is a single-cycle, forced-circulation, 10 CFR § 52.53 is provided in Chapter 21. A final design boiling water reactor (BWR) with a rated power of approval, issued on the basis of this SER, does not 3926 megawatts thermal (MWt) and a design power of constitute a commitment to issue a permit or license, or in 4005 MWt. Many features of the ALBWR design are any way affect the authority of the Commission, the similar to those of BWR designs that the staff had Atomic Safety and Licensing Board, and other presiding previously approved. To the extent feasible and officers, in any proceeding pursuant to Subpart G of appropriate, the staff relied on earlier reviews for those 10 CFR Part 2. iii iiiNUREG- 1503 b. CONTENTS VOLUME 2: APPENDICES Page *APPENDIX A..................... A -i... LIST OF ABBREVIATIONS............... .. A ...-i APPENDIX B..................... .I . B-i... REFERENCES........................ .. B-I... APPENDIX C..................... .. C-1 ... CHRONOLOGY OF CORRSPONDENCE .... C-i ... APPENDIX D..................... .. D-1 ... FSER CONTRIBUTORS......... ........ D-1 ... APPENDIX E..................... .. E-1 ... STAFF POSITION ON SHELL BUCKLING DUE TO INTERNAL PRESSURE... E-i ... APPENDIX, F.............................................. .. F-1 ... STAFF POSITION ON STEEL EMBEDMENTS.......................... F-i ... kAPPENDIX G.......................................... .. G-1 ... STAFF POSITIONS AND TECHNICAL BASES ON THE USE OF AMERICAN NATIONAL STANDARDS INSTITUTE (ANSI)/AMERICAN INSTITUTE OF STEEL CONSTRUCTION (AISC) N690, -NUCLEAR FACILITIES - STEEL SAFETY-RELATED STUTRS...... G-1 APPENDIX H...................................................... H-1 DYNAMIC LATERAL SOIL PRESSURES ON EARTH RETAINING WALLS AND EMBEDDED WALLS OF NUCLEAR POWER PLANT STRUCTURES.................................... H-i APPENDIX I........................................................I1-1 EVALUATION OF ABWR PUMP AND VALVE INSERVICE TESTING PLAN (SSAR TABLES 3.9-8 AND 3.9-9)....................................................11 APPENDIX J.............................. I .......................... J-i HUMAN FACTORS ENGINEERING PROGRAM REVIEW MODEL AND ACCEPTANCE CRITERIA FOR EVOLUTIONARY REACTORS......................................................J1 APPENDIX K....................................................... K-1 IMPORTANT SAFETY INSIGHTS.................................................... K-1 V ~NUREG-1503 V FIGURES page G-1 Structural stability curves for the axially loaded compression numbers. .. .. .. .. .. .. .. .. .. ....... G-3 W J. I HFE program review model elements .. .. ..... .. .. .. .. .. .. .. .. .... .. .... .. .... .......... J-4 J.2 BIFE program review stages.............................................. ........ J-5 NUREG- 1503 viv Appendix A LIST OF ABBREVIATIONS A- - C AB)WR advanced boiling water reactor CA compressedl air AC alternating current CAMS containment atmosphere monitoring system ACC accumulator CAV cumulative absolute velocity ACI automatic closure and interlock CBERU contrbuilding emergency recirculation unit ACIWA ac-independent water addition CBRU control-building recirculation unit ACRS Advisory Committee on Reactor CBSREA control building safety-related equipment Safeguards area ACS atmospheric control system CCDF complementary cumulative distribution ACU air conditioning units function ADS automatic depressurization system CCI core concrete interaction AHU air handling units CCS condensate cleanup system AISC American Institute of Steel Construction CDF core damage frequency ALARA as low as is reasonably achievable CDM Certified Design Material ALWVR advanced light water reactor CDRL core damage radiation level AM accident management CE ABB-Combustion Engineering ANL Argonne National Laboratory CED common engineering documents ANS American Nuclear Society CET containment event trees ANSI American National Standards Institute CF&CAE condensate feedwater, and condensate air AQO(s) anticipated operational occurrences extraction APR automatic power regulator CFR Code of Federal Re~yulations APRM average power range monitor CFS condensate and feedwater system ARI alternate rod insertion CH chugging O"ASamplified response spectra CIV containment isolation valve *ASB Auxiliary Systems Branch cm centimeters ASCE American Society of Civil Engineers CMAA Crane Manufacturers Association of ASD allowable stress design America ASD adjustable speed drive CMP configuration management plan ASF automatic suppression function CMU control