ML20140F080

Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design.Supplement No. 1.Docket No.52-001.(General Electric Nuclear Energy)

ML20140F080
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Site:	05200001
Issue date:	05/31/1997
From:	Office of Nuclear Reactor Regulation
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References
NUREG-1503, NUREG-1503-S01, NUREG-1503-S1, NUDOCS 9706120299
Download: ML20140F080 (49)

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l l AVAILABILITY NOTICE l 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 l 20555-0001 1 2. The Superintendent of Documents. U.S. Government Printing Office, P. O. Box 37082 Washington, DC 20402-9328 3. The National Technical Information Service, Springfield, VA 22161-0002 l Although the listing that follows represents the majority of documents cited in NRC publica-l tions, it is not intended to be exhaustive. Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; vendor reports and correspondence; Commission papers; and applicant and licensee docu-ments and correspondence. The following documents in the NUREG series are available for purchase from the Government Printing Office: formal NRC staff and contractor reports, NRC-sponsored conference pro-ceedings, international agreement reports, grantee reports, and NRC booklets and bro-j chures. Also available are regulatory guides, NRC regulations in the Code of Federal Regula-j tions, and Nuclear Regulatory Commission Issuances. Documents available from the National Technical Information Service include NUREG-series reports and technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission. Documents available from public and special technical libraries include all open literature items, such as books, journal articles, and transactions. Federal Register notices, Federal l and State legislation, and congressional reports can usually be obtained from these libraries. Documents such as theses, dissertations, foreign reports ar d translations, and non-NRC con-j ference proceedings are available for purchase from the organization sponsoring the publica-l tion cited. l Single copies of NRC draft reports are available free, to the extent of supply, upon written i request to the Office of Administration, Distribution and Mail Services Sectioil, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001. l l Copies of industry codes and standards used in a substantive manner in the NRC regulatory i process are maintained at the NRC Library, Two White Flint North 11545 Rockville Pike, Rock-l ville, MD 20852-2738, for use by the public. Codes and standards aro usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018-3308. i i l

NUREG-1503 Supp.1 Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design Manuscript Completed: April 1997 Date Published: May 1997 Division of Reactor Program Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 f# "*%,,,

ABSTRACT This report supplements the final safety evaluation report (FSER) for the U.S. Advanced Boiling Water Reactor (ABWR) standard design. 'Ihe FSER was issued by the U.S Nuclear Regulatory Commission (NRC) staff as NUREG 1503 in July 1994 to document the NRC stafTs review of the U.S ABWR design. The U.S. ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the prou: dun:s of Subpart B to Part 52 ofTitle 10 of the Code of Federal Remdations. This supplement documents the NRC staffs review of the changes to the U.S. ABWR design documentation since the issuance of the FSER. OE made these changes primarily as a result of first-of a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis ofits evaluation, the NRC stafTconcludes that the confumatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the U.S. ABWR design. t iii NUREG-1503 Supplement 1

CONTENTS l l Page ABSTRACT., .............. iii 1 INTRODUCTION AND GENERAL DISCUSSION. .1-1 1.1 Introduction.. ......1-1 1.2 General De3ign Description 1-1 1.2.2 Precertification and Postcertification Design Control Procedures... 1-1 1.5 Summary OfPrincipal Review Matters 1-2 1.6 Index of Applicable Regulations and Exemptions 1-2 l 1.7 Index ofTier 2* Information . 1-2 1.8 Index ofConfirmatory Items. 1-3 t 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS..... ... 3-1 3.8 Design of Seismic Category I Structures... ... 3-1 4 REACTOR. . 4-1 4.6 Functional Design of Fine Motion Control Rod Drive System 4-1 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS . 5-1 5.4.6 Reactor Core Isolation Cooling System (RCIC) . 5-1 9 AUXILIARY SYSTEMS . 9-1 9.2 Water Systems.. . 9-1 9.4 Heating, Ventilating, and Air Conditioning Systems. ... 9-1 12 RADIATION PROTECTION 12-1 i 12.3 Radiation Protection Design 12-1 14 INITIAL TEST PROGRAM 14-1 14.3 Certified Design Material 14-1 l 14.3.7.5 Reliability Assurance Program 14-1 1 15 TRANSIENT AND ACCIDENT ANALYSES . 15-1 15.4 Radiological Consequences of Accidents .. 15-1 l 16 TECIINICAL SPECIFICATIONS.. 16-1 i 19 SEVERE ACCIDENTS. 19-1 l 20 GENERIC ISSUES . 20-1 20.5 10 CFR 50.34(f), Additional TMl Requirements .20-1 20.5.1.3 Identification of Potential Design Improvements. . 20-1 21 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS .. 21-1 22 CONCLUSION .22-1 Appendix C CONTINUATION OF CHRONOLOGY OF CORRESPONDENCE C-1 Appendix D CONTRIBUTORS TO THIS FSER SUPPLEMENT . D-1 i 7 Appendix L ERRATA TO THE ABWR FSER . L-1 v NUREG-1503 Supplement 1

1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction 1.2 General Design Description This report supplements the final safety evaluation repen 1.2.2 Precertification and Postcertification (FSER) for the U.S. Advanced Boiling Water Reactor Design Control Procedures (ABWR) standard design. The FSER was issued by the U.S. Nuclear Regulatory Commission (NRC) stafTas NUREG-GE coordinated certain design issues with its international 1503 in July 1994 to document the NRC staffs review of the associates that affected the ABWR design being built in ABWR design. This supplement documents the NRC staffs Japan using a system of common engmeering documents review of the changes to the ABWR design documentation (CEDs). Design changes identified in CEDs were since the issuance of the FSER. OE made these changes maintained by GE in a controlled list for future action called primarily as a result of first-of-a-kind-engineering (FOAKE) the Design Action List (DAL). GE intended to incorporate and as pan of the design cenification mlemaking for the the items on the DAL into the supporting documentation for ABWR design. Specifically, this supplement documents the the U.S. ABWR design as part ofits ongoing design resolution of confirmatory items in the FSER relating to the actisities. preparation of the Design Control Document (DCD); it provides an evaluation of changes to the ABWR design made In the FSER, the staff stated that GE must certify to the NRC as part of FOAKE; it provides an additional evaluation of that the U.S. ABWR DCD was not afTected by any changes radiological release infonnation not included in the FSER; to the ABWR CEDs. This was based on a GE letter of June and it provides errata to the FSER. I 1,1993, stating that, after completing the DCD, GE would cenify that the Tier I and Tier 2 information had not been GE Nuc! car Energy (GE, the applicant) submitted the afTected by any subsequent changes made in the CEDs since ABWR design documentation under Subpart B of Part 52 of the final Tier I and Tier 2 submittals. This was FSER Title 10 of the Code of Federal Reculations. The Confirmatory Item Fl.2.2-1. dccumentation and infomiation penaining to this supplement were submitted on Docket No. 52-001. The ABWR design In a letter of December 22,1994, GE certified that the de:umentation includes the standard safety analysis report information in the DCD had not been affected by any (SSAR), certified design material (CDM), and the DCD. changes in the CEDs since the final submittals of the SSAR and CDM. GE further stated in a letter ofJanuary 26,1995, Each of the following sections or appendices of this that all peninent information from the DAL and CEDs had supplement is numbered and titled the same as the section or been incorporated in Revision 2 of the DCD for the ABWR appendix of the FSER that is being updated. The discussions design. This is acceptable and resolved FSER Confirmatory are supplementary to and not in lieu of the discussion in the Item F1.2.2-1. FSER unless othenvise noted. Accordingly, Appendix C is a continuation of the chronology of correspondence for the In the FSER, the stafTstated that GE must give the staff a list review, Appendix D is a list of principal contributors to this of the ABWR CEDs and the DAL that applied to the U.S. supplement, and Appendix L contains errata to the FSER. ABWR design and their effective dates. In a letter ofJune No significant changes were made to FSER Appendices A, 11,1993, GE stated that, after completing the DCD, it would B, E, F, G,11, I, J, and K by this supplement. finalize the DAL and submit the corresponding effective dates of CEDs and the DAL. His was Confmnatory item This supplement is issued by the Standardization Project Fl.2.2-2. Directorate in the Ollice of Nuclear Reactor Regulation. The NRC's licensing project manager for the U.S. ABWR design In a letter of January 26,1995, GE described the design is Dino C. Scaletti lie may be reached by calling (301) configuration supporting the most current DCD resision for 415-1104, or by writing to the Ollice of Nuclear Reactor design cenification (DCD Revision 2). Further, GE stated Regulation, U.S. Nuclear Regulatory Commission, that it maintains a Wster Parts List (MPL) of record that Washington, DC 20555-000). Copics of the ABWR design contains a complete list of all ABWR certification dccumentation and all amendments and revisions are documents, including CEDs and the DAL, thereby available for public inspection at the NRC's Public maintaining records of the documents supporting the design Document Room,2120 L Street NW. (Lower Level), certification. For future design efTorts, GE will follow Washington, DC. Copies of the ABWR FSER and this procedures to control the detailed design work to ensure supplement are also available at the NRC's Public Document conformance with the DCD. Therefore, although the design Room. documentation supporting the design may change, these changes are traceable and provide the ability to retrieve the design documentation suppo! ting the design cenification. 1-1 NUREG-1503 Supplement I

This is acceptable and resolved FSER Confirmatory item Section 1.7 below); proprietary infonnation in the SSAR that Fl.2.2 2. was not included in the DCD (refer to SSAR proprietary volumes); safeguards information in the SSAR that was not The staffissued a notice of proposed rulemaking for the included in the DCD (refer to SSAR Section 13.6); and ABWR design certification in the Federal Register that detailed probabilistic risk assessment (PRA) information in incorporated DCD Revision 2 by reference. Subsequently, the SSAR that was not included in the DCD. The treatment GE submitted additional changes to the ABWR design, as of these issues is described in the ABWR design certification discussed in Section 1.5 of this supplement. Therefore, in a rule and the statements of consideration (SOC) and are, letter dated June 10,19%, GE confirmed that the MPL therefore, not discussed in this supplement. For further continued to derme the design documentation implementing information, see the final rule and the SOC section-by-the NRC approved design, including the CEDs and the section discussion for Section IV," Additional Requirements DAL This is acceptable. and Restrictions." 1.5. Summary of Principal Review Matters GE submitted Revision 4 to the DCD on March 28,1997. This revision includes changes, such as the DCD GE prepared the CDM and SSAR as part ofits design introduction, that were made to conform the ABWR DCD certification application for the U.S. ABWR standard design. with the final design certification rule. This version of the The NRC stafTprovided its evaluation of the design in the ABWR DCD is approved by this supplement to the ABWR FSER (NUREG 1503). Subsequently, GE prepared a FSER and is the version that will be incorporated by separate document called the Design Control Document reference into the fmal design certification rule for the U.S. (DCD) to be incorporated by reference into the ABWR ABWR standard design. design certification rule. The DCD has two tiers of infonnation that were derived from and include most of the information in the CDM and the SSAR. 1.6 Index of Applicable Regulations and The staff stated in the FSER that, after issuance of the FSER, Exemptions the applicant for design certification will submit a DCD for the staffs review. This was Confirmatory item F1.5-1, GE In the FSER, the NRC staffidentified new standards for suomitted Revision 0 of the DCD on October 28,1994. In selected technical and severe accident issues for the U.S. general, GE followed the NRC staff guidance in letters of ABWR design that were addressed and resolved durmg the August 26,1993 and August 3,1994 regarding the format of design certification review. These new design standards the DCD. The stafTprovided comments on the DCD, and were consequently included as additional applicable GE addressed all of the NRC stafTcomments in DCD regulations in the proposed rule for the purposes of 10 CFR 1 Revisions I and 2. This is acceptable and resolved FSER 52.48,52.54,52.59, and 52.63. The Commission decided Confirmatory item F1.5-1. not to codify the additional applicable regulstions in the final rule, but the Commission did set forth its intent with regard DCD Revision 2 was the last revision the NRC received to these new design standards in its SOC for the final design belbre issuing the notice of proposed rulemaking for the certification rule (See the SOC public comment summary ABWR design in the Egderal Register on April 7,1995. and resolution section on the need for additional applicable Subsequently, GE proposed additional changes to the regulations). ABWR design documentation resulting from detailed design work in FOAKE. These proposed changes were documented 1.7 Index of Tier 2* Information in a letter dated April 16,1996, a meeting summary dated May 8,19%, and a letter dated June 10,19%. GE provided In the FSER, the stafistated that any changes to certain a revised submittal on July 1,1996. The staffs review of SSAR commitments would require prior NRC approval these changes is set fbrth in the appropriate sections of this before the change was implemented by a COL applicant or supplemental FSER. licensee who referenced the ABWR certified design. The stafflisted these SSAR commitments in the FSER, and GE submitted Revision 3 to the DCD, Revision 3, on August required that they be identified in the DCD as 30,1996. Revision 3 incorporated the changes documented " Tier 2*" information. This was FSER Confirmatory item in its letters of June 10 and July 1,1996. GE also submitted Fl.7-1. SSAR Amendment 37 (Revision 9) and CDM Revision 8 on August 30,19%, to make these documents consistent with GE identified the Tier 2

• information in the appropriate Revision 3 to the DCD. Ilowever, some difTerences between sections of the DCD, although GE did not designate this the DCD and the SSAR/CDM remained that GE documented information in the SSAR. The stafTdid not require the in its letter of August 30,1996. These difTerences include designation in the SSAR because the Tier 2* infonnation designation of Tier 2* information in the DCD (refer to NUREG-1503 Supplement I l-2

was described in the staffs FSER. This is acceptable and 1.8 Index cf ConfirmItory Items resolved FSER Confirmatory item Fl.7-l. Section 1.8 of the FSER listed a total of five confirmatory in various locations in the FSER, the stafT stated that any issues The FSER stated that these items would te resolved changes to Tier 2* information would involve an unreviewed during the staffs review of the ABWR design control safety question (USQ) and, therefore, require NRC review document. In the FSER, each confirmatog item was r.nd approval prior to implementation. This statement assigned a unique number that identified the section of the regarding USOs was used simply to indicate that the change FSER where the item was discussed. This number was listed process for Tier 2* information would be the same as that for in parentheses. For example, Confirmatory item Fl.5-1 was proposed changes to other Tier 2 information that is discussed in Section 1.5 of the FSER. All of these issues detennined by an applicant or licensee to be a USQ. have been resolved as discussed in the corresponding flowever, a determination of whether or not a proposed sections of this report. change to the Tier 2* information would constitute a USQ hes not been made by the NRC, and the actual process for Item Number Descrintion ofItem ch:nging Tier 2* information is described in the final design certification rule. Therefore, the language in the FSER has Fl.2.2-1 Certification that the DCD was not been modified to conform with the language of the final rule affected by changes to CEDs and its SOC (See tie rule and the SOC section-by section analysis regarding the processes for changes and departures, F l.2.2-2 Submittal of a list oiCEDs and DALs and the SOC public comment summan and resolution section regarding the Tier 2 change process) by the errata in Fl.5-1 Submittal of DCD Appendix L to this supplement. Fl.7-1 Identify Tier 2* infbrmation F 14.3.7.5-1 Reliability Assurance Program i f 1-3 NUREG-1503 Supplement 1 J

3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS 3.8 Design of Seismic Category I Structures GE proposed changing the DCD to use higher strength materials for the submerged portions of the lower drywell access tunnels and reactor pressure vessel (RPV) pede.nal. The submerged portions will be clad with stainless steel to improve the corrosion resistance of the materials. GE determined that the changes were necessary based on considerations for the ability to clad and the strength to withstand high thermal stresses. The submerged portion of the RPV pedestal shell required a stainica steel clad and the AS72 material specified in the DCD was not suitable for cladding. The access tunnel shells must withstand a high thermal stress as predicted by detailed structural analyses and the A516 material specified in the DCD had a relatively low yield strength at room temperature. GE proposed ASTM A533, Type B, Class 2 material for the clad portions of the RPV pedestal and lower drywell access tunnels. This material is a fully killed steel, which haa very good cladding properties and high strength. Therefore, the NRC stalTfound these changes to be acceptable because the properties of the proposed material are suitable for the intended application. I 3-1 NUREG-1503 Supplement 1

4 REACTOR 4.6 Functional Design of Fine Motion Control Rod Drive System GE submitted a change to increase the design pressure of the fine rnotion control rod drive (FMCRD) scram piping, based on tests and evaluations of water hammer effects. The changes are consistent with the ASME Code, which requires the use of equipment events rather than plant events in detennining the design pressure. The NRC staff reviewed the change and found it to be acceptable. e 1 4-1 NUREG-1503 Supplement 1

5 REACTO.R COOLANT SYSTEM AND CONNECTED SYSTEMS SA.6 Reactor Core Isolation Cooling System (RCIC) GE submitted a change to delete the rupture disks that were originally intended to protect tlw low pressure exhaust side of the RCIC turbine case and exhaust line from over pressurization. OE detennined that the rupture discs created the potential for an interfacing systems LOCA if the system were subjected to an over pressure situation. This potential would be counter to the purpose for the previous upgrade of the RCIC turbine exhaust sysem to withstand maximum system pressure. Therefore, GE removed the ruptr.re disks from the design to reduce the potential for a LOCA s the turbine exhaust system. The i NRC staff fmds tiss change to be acceptable because it does not cha..ge the findings in the FSER. i i l l t 5-1 NUREG-1503 Supplement I

9 AUXILIARY SYSTEMS 9.2 Water Systems 9.4 Heating, Ventilating, and Air Conditioning Systems l GE determmed that the federally mandated phaseout of l several commonly used refrigerants required the following GE modified the Reactor Building and Radwaste Building l design changes: (1) changing the description of the iIVAC systems to improve system reliability and 4 chillers for the !Icating, Ventilating, and Air Conditioning maintainability. Specifically, GE used electric heating in (liVAC) Emergency Cooling Water System so that either a place of hot water heating to protect against freezing in the centrifugal or positive displacement type pump could be pipes of the hot water system during adverse outside used in the system;(2) changing the description of the environmental conditions for the Reactor Building l Reactor Building Cooling Water (RCW) heat exchangers secondary containment and Radwaste Building IIVAC l ca that either a shell and tube or plate type heat exchanger systems. This also eliminated the relatively large hot water could be used in the system; and (3) lowering the ultimate piping that was routed from the reactor building to the heat sink (Ulis) design temperature from 37.8 *C (100 'F) turbine building and back. GE also split the secondary to 35 *C (95 'F). GE submitted these changes to the containment 11VAC system air intake into three 50-percent ABWR design documentation as part ofits DCD Revisions air handling units, thereby providing redundancy to 0,1, and 2. The staff reviewed the above changes made in enhance system reliability and facilitate sysem the ABWR design documentation and determined that the maintenance. Further, GE used high efficiency filters in changes to the heat exchangers and refrigeration systems place of medium grade bag-type filters to provida. were not safety significant because they did not affect the additional filtering capability and reduce maintenance performance requirements of the systems. Further, the stafT problems associated with bag-type filters. The staff had based its safety evaluation in the FSER on a Ulls concludes that the changes are design improvements that temperature of 95 'F, thus bounding the UilS design enhance the system reliability and maintainability, did not temperature change. Some changes to the FSER are alTect the fundamental safety decisions in the FSER, and contained in the errata in Appendix L to this supplement. are acceptable. Therefore, although the above changes affect background discussions in the FSER, the stafTconcludes that none of GE modifiul the control building and reactor building them affect the fundamental safety decisions in the FSER. safety-related llVAC systems to change the smoke removal methods, replaced centrifugal fans with vane axial fans, Subsequently, GE submitted the following additional and switched divisional power supply to the fine motion change resulting from its first-of a-kind-engineering control rod drive (FMCRD) panel rooms. The first change 1 (FOAKE) work and included it in the final DCD. GE provided positive exhaust ventilation rather than provided an additional chiller / pump set for Division A of pressurization for smoke removal from the areas served by the IIVAC Emergency Cooling Water (llECW) System. the 11VAC systems. A dedicated smoke removal fan was The change provided function r1 redundancy to avoid the added to the reactor building safety related electrical loss of cooling to safety-related electrical equipment area equipment (SREE)liVAC system. In addition, both l IIVAC systems for the contrt g and reactor exhaust fans in the control room habitability area (CRilA) n building (which could potentially cnallenge electrical and control building safety-related equipment area equipment environmental qualification temperature limits), (CBSREA)IIVAC systems are activated fbr smoke and satisfied system maintenance needs. Consistent with removal. Finally, the reconfigure. tion of duct work Divisions B and C, an additional chiller / pump set for arrangements in the control building 11VAC systems and Division A of the IIECW System provided for the reactor building SREE 11VAC system provides a single perfomiance of chiller / pump on-line maintenance, whereby cross-connect from the retum to the exhaust duct work, a controlled environment can be sustained in Class 1 E except that the diesel generator, day tank, chiller, and equipment rooms in Division A of the reactor building and battery rooms in the above systems exhaust directly to the control building. The stafTconcludes that the change is a outside atmosphere. For the second change, the centrifugal l design improvement that enhances system reliability and exhaust fans were replaced with 2-speed vane axial fans maintainability, did not change the fundamental safety and centrifugal supply fans were replaced with single I decisions in the FSER, and is acceptable. speed vane axial fans in the CRllA system for space conservation. Also, the centrifugal supply and exhaust fans were replaced with single speed vane axial fans in the { CBSREA system. Finally, the distribution of the cooling loads were balanced by assigning FMCRD panel rooms to Divisions A and B. The NRC stafTconcludes that the design changes enhanced the system ftmetions, and did 9-1 NUREG-1503 Supplement 1 1

not affect the fundamentoi safety decisions in the FSER. Therefore, the changes are acceptable. GE reassigned the main control rcxxn ilVAC exhaust fans according to their respective divisional space. The changes reassigned the main control room HVAC exhaust fans ("B" as "C" and "C" as "B") accordmg to their respective divisional space to eliminate the potential divisional cross-over ofcooling water and power, to enhance exhaust fan performance due to less complexity in duct work, and to climint.te potential for breaching fire barriers and fire proofing the duct work. The staff concludes that the changes are design improvements that enhance the systems reliabil ty and maintainability, did not afTect the fundamental safe.ty decisior:a in the FSER, and are acceptable. GE corrected various inconsistencies between Tier 1 and Tier 2 of the DCD in the reactor building secondary containment IIVAC and reactor building non-safety related equipment ilVAC systems. The NRC staficoncludes that the changes ensure that the design documentation more accurately represents the design, did not afTect the fundamental safety decisions in the FSER, and are acceptable. GE mcxlified the CRilA !!VAC system to provide two independent Class I E power sources for each pair of boundary isolation dampers, except for the motor-operated isolation dampers which are designed to fail as-is. GE also added a cross-tie in Tier i between the two intet ducts of emergency filtration unit in each division. The changes assure the necessary aligmnent of dampers and prevents the infiltration of unfdtered in-leakage during a design basis accident and loss of one division of power. The NRC staff concludes that the changes meet SRP Section 9.4 regarding single failure criteria and are,'herefore, acceptable. 1 1 1 l l i NUREG-1503 Supplement 1 9-2

12 RA!!ATIONPR3TECTION 12.3 Radiation Protection Design GE submitted changes to the reactor building radiation zone maps in the DCD. The changes involved minor detailed aspects of the drawings. The NRC staff fmds these changes acceptable because they do not change the findings in the FSER. 4 9 12-1 NUREG-1503 Supplement 1

i 14 INITIAL TEST PROGRAM 14J Certified Design Material 14.3.7.5 Reliability Assurance Program i In the FSER, the NRC staff required a high-level commitment to a reliability assurance program in the design documentation. GE committed to provide the required information in a letter dated July 12.1994. This l was FSER Ccafumatory Item 14.3.7.$-l. GE prosided the l required information in a modification package to the SSAR and CDM on July 20,1994, and also included this information in the DCD. This is acceptable and resolved FSER Confirmatory item F14.3.7.5-1. ) i l \\ l l 14-1 NUREG-1503 Supplement 1

. - = - - 15 TRANSIENT AND ACCIDENT ANALYSES 15A Radiclogical Consequences of Accidents Several tables in Section 15.4 of the FSER inadvertently contained incorrect values that were used in the radiological analyses for the ABWR design. These values have been corrected as part of the errata in Appendix L of this supplement. Subsequent to issuance of the FSER in July 1994, GE changed the ABWR control room ventilation and filtration system design as reflected in the revisions to FSER Table 15.9. These changes resulted in a decrease in the control room filtered air intake, an increase in the intake filter (flicacy; an increase in charcoal adsorber iodine removal efficiency; and a reduced rate of filtered air recirculation flow. These changes have been incorporated into the SSAR and the DCD. The NRC staff has reviewed these design changes using similar methodology as that described in FSER Section 1S.4, and finds that the changes result in lower radiation doses to control room personnel than previously calculated. The revised Chi /Qs are reflected in the errata in Appendix L to this supplement. Therefore, the NRC staficoncludes that the design changes do not alter the conclusions reached in FSER Section 15.4 and in FSER Section 6.4, "Contn>l Room Habitability Systems," and are acceptable. l t l l l 15-1 NUREG-1503 Supplement 1 l l i

16 TECHNICAL SPECIFICATIONS GE modified the technical specifications (TS) by changing the location for temperature measurement from the average UllS temperature to the permanently installed temperature . elements in the reactor service water (RSW) system at the inlet piping to the heat exchangers for the reactor building l-cooling water system. Additionally, the TS temperature ) limit was reduced to 33.3 'C to provide sufficient operating margin to ensure that the RSW inlet temperature did not exceed the 35 *C assumed for tim loss of coolant design basis accident. The TS temperature is sufficiently above j _ the expected maximum normal operation RSW inlet temperature of 32.8 *C to minimize the potential of reaching the TS limit during normal operation. Also, GE I modified the surveillance frequency for testing the set 1 l points of safety relief valves from 18 months to that l specified in the Inservice Testing program, consistent with 'i the BWR standard TS. The NRC staficoncludes that the changes ensure system performance, meet the requirements of 10 CFR 50.34 and 50.36 for technical specifications for the design, and are acceptable. 4 i l l l l l l 1 ) l a 16-1 NUREG-1503 Supplement I l,

i 19 SEVERE ACCIDENTS GE submitted changes to the DCD to make the probabilistic risk assessment (PRA) and severe accident I analyses more accurately reflect the approved design. The changes were primarily to (1) correct the description of the emergency diesel generator and combustion turbine generator load carrymg capabilities, (2) correct discussions of the turbine service water isolation features, and (3) update the seismic margins analysis. The NRC staff reviewed the changes to Chapter 19 and concludes that none of the changes significantly affected the PRA or severe accident analyses or results and, therefore, do not change the fundamental safety decisions in the FSER. The NRC staff also modified its FSER to reflect the revised descriptions in the DCD as part of the errata in Appendix L to this supplement. I 19-1 NUREG-1503 Supplement I

20 GENERIC ISSUES 20.5 10 CFR 50.34(f), Additional TMi The NRC stafT reviewed the design alternatives identified Requiremenis in the TSD using $2,000/ person-cSv averted for health effects and adopting a $3,000/ person-cSv supplemental 20.5.1.3 Identification of Potential Design allowance for ofTsite property (See NUREG/CR-6349, Improvements " Cost benefit Considerations in Regulatory Analysis"). Assuming a base case 7% real discount rate as prescribed The numerical values and discussion in this FSER section in NUREG/BR-0058, Revision 2, the present value of the were based on the values in Section 19P of the SSAR. As health and safety benefits attributable to a cost-beneficial part of the design certification rulemaking, GE updated design improvement would approximate $233,000. Thisis SSAR Section 19P, but did not include it in the DCD. a factor of about 1.2 times higher than the $200,000 Instead, GE rekicated it to GE's " Technical Support estimate identified in the FSER. A comparable estimate Document (TSD) for the ABWR", Revision 1, December for the health and safety benefits of a cost-beneficial design 1994, which was contained in an attachment to a Ictier modification based on a 3% real discount rate, which is from GE to the NRC dated December 21,1994. In the recommended for sensitivity analysis purposes, is errata in Appendix L to this supplement, the stafrupdated $460,000, or 2.3 times greater than the $200,000 estimate the FSER to correspond to the latest information in the in the FSER. TSD. Also, the discussion in this supplement conforms with the fmal environmental assessment issued with the Most of the candidate design attematives in the TSD were design certification rule, and is based on the updated estimated to cost more than $460,000 and, therefore, were information in the TSD. not cost-beneficial. Of the design alternatives that cost less than $460,000, the drywell head flooder was the most cost. In the FSER, the NRC staff utilize 1 a value of beneficial design m(xiification ($1.7 million/ person-cSv $ 1,000/ person-cSv ($ 1000/ person-rem) averted to averted), as shown in Table 20.5.1-3 of the FSER. estimate that a design improvement that cost more than However, given that the drywell head flooder was $200,000 would not be cost beneficial. This figure estimated to cost on the order of $ 100,000, under either the conservatively assumed that the total 60-year lifetime risk 7% or 3% discount rate scenario, this design alternative for the ABWR was eliminated by the design improvement would have to eliminate at least 43% or 22%, respectively, (200 person-cSv averted risk x $1,000/ person-cSv = of the total lifetime risk. Since the drywell head flooder $200,000). Since the FSER was issued, the NRC issued was estimated to only account for less than 10% of the total " Regulatory Analysis Guiddines of the U.S. Nuclear risk, even for this most cost-beneficial design modification, Regulatory Commission" (NUREG/BR-0058, Revision 2, the total costs continued to be well in excess of the total November 1995). This guidance document adopted a benefits. $2,000/ person-cSv ($2000/ person-rem) conversion factor, subject to present worth considerations, and is limited in In summary, the NRC staff concludes that with the scope to health efTects. Limiting the conversion factor significant margins in the results of the cost-benefit solely to health efTects required that the regulatory analysis analysis, consideration of severe accident design include an additional dollar allowance for averted ofTsite attematives using the new values provided in NUREG/BR-property damage. 0058 do not change the findings in the previous analysis in the FSER. 20-1 NUREG-1503 Supplement 1

21 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Committee on Reactor Safeguards considered the information discussed in this supplement to the ABWR FSER during their 433rd meeting on August 8, 19%, and subsequently issued its letter on August 15, ' 1996. The letter, which follows, reflects approval of the l application for design certification and includes no i rewmmended actions for either the NRC staff or GE Nuclear Energy. l l i l l 1 l l l i e 21-1 NUREG-1503 Supplement 1

-~ me po m[q,, UNITED STATES NUCLEAR REGULATORY COMMISSION n ( ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8, WASHlHGTON, D. C. 20555 U,* S August 15, 1996 l The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20$55-0001

SUBJECT:

DESIGN CHANGES PROPOSED BY GENERAL ELECTRIC NUCLEAR ENERGY RELATING TO THE CERTIFICATION OF THE U.S. ADVANCED BOILING WATER REACTOR DESIGN During the 433rd meeting of the Advisory Committee on Reactor Safeguards, August 8-10, 1996, we reviewed recent design changes proposed by General Electric Nuclear Energy (GENE) relating to the certification of the U.S. advanced boiling-water reactor (ABWR) design. These " design changes

• censist of both actual modifications to the design and corrections to the documentation to remove inconsistencies and typographical errors. We had the benefit of discussions with representatives of the NRC staff and of GENE. We also had the benefit of the documents referenced.

Conclusjsni Our review of Supplement I to NUREG-1503, " Final Safety Evaluation Report Related to the Certification of the U.S. ASWR Design," did not change the conclusion reached in our earlier report of April 14, 1994. We continue to believe that acceptable bases and requirements have bei established in the application to assure that the U.S. ABWR Standard Design ro be used to engineer and censtruct plants that with reasonable assurance canf4 sperated without undue risk to the health and safety of the public. Backaround and Discu111gn We have been involved in the review of the U.S. ABWR design since GENE applied for certification. This review was carried out in accordance with 10 CFR Part 52, which requires ACRS to report on those portions of 10 CFR Part 52 applications that concern safety. In our April 14, 1994 report to the Commission, we supported the certification of the U.S. ABWR design. This report was included in the staff Safety Evaluation Report (NUREG-1503). The present review is intended to supplement our earlier review of this ABWR application. Sincerely, l J S. / w l T. S. Kress e Chairman l NUREG-1503 Supplement 1 21-2

l 2 Referettcal: l. U. S. Nuclear Regulatory Cossaission, NUREC-1503, Supplement No.1, ' Final j Safety Evaluation Report Related to the Certification of the Advanced 3 Boiling Water Reactor Design," dated July 1, 1996 2. Staff Requirements Memorandum dated June 11, 1996, from John C. Hoyle, Secretary, to John T. Larkins, ACRS, regarding meeting with Advisory Committee on Reactor Safeguards, Nay 24, 1996 l 3. ACRS Report dated April 14, 1994, from T. S. Kress, Chairman, ACRS, to i Ivan Selin, Chairman, NRC,

Subject:

Report on the Safety Aspects of the General Electric Nuclear Energy Application for Certification of the Advanced Boiling Water. Reactor Design 4. Letter dated April 16, 1996, from J. F. Quirk, GE Nuclear Energy, to Dennis N. Crutchfield, Nuclear Regulatory Commission, regarding ABWR design changes 5. Letter dated July 1,1996, from J. F. Quirk, CE Nuclear Energy, to the Nuclear Regulatory Commission,

Subject:

ABWR Design Control Document Changes 1 ? i 1 1 a 4 f 21-3 NUREG-1503 Supplement I

22 CONCLUSION The NRC staff performed its review of changes made to the U.S. ABWR design documentation by GE Nuclear Energy in its letters dated June 10 and July 1, 1996 and other changes made to conform the U.S. ABWR Design Control Document (DCD) to the fmal design certification rules. The design changes were reviewed by the Advisory Committee on Reactor Safeguards as described in Chapter 21 of this report. On the basis of the evaluation described in NUREG-1503 and this report, the NRC staff concludes that the l confinnatory issues in NUREG-1503 are resolved, the changes to the ABWR design documentation are acceptable, and GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the U.S. ABWR design. I 1 22-1 NUREG-1503 Supplement I

Appendix C CONTINUATION OF CHRONOLOGY OF CORRESPONDENCE This appendix contains an update of the chronological list of routine licensing correspondence in Appendix C of NUREG 1503. The correspondence is between the U.S. Nuclear Regulatory Commission (NRC) staff and GE regarding the review of the Advanced Boiling Water Reactor (ABWR) under Project 671 and Docket Numbers 50-605 and 52-001. Correspondence regarding the ABWR design certification rulemaking is not included here, but may be found in the rulemaking records. July 20,1994 Jack N. Fox, GE, Forwards Revision 7 to ABWR SSAR and Revision 6 to CDM Fiche: 80480-222/80480-297 acn: 9407220186 l July 25,1994 Dennis M. Crutchfield, NRC, Letter responding to NEI letters of 6/24/94 and 6/23/94 Fiche: 80755-297/80755-297 acn: 9409010086 July 28,1994 Thomas 11. Boyce, NRC, Summary of meeting on 7/12/94 to discuss preparation of the ABWR Design Control Document Fiche: 80473-136/80473-151 acn: 9408050011 July 31,1994 Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design, Vol. I Fiche: 80681-001/80683-148 acn: 9408260011 July 31,1994 Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design, Vol 2 Fiche: 80659-098/80659-306 acn: 9408250023 August 2,1994 Jack N. Fox, GE, Forwards ten copies of drafi ABWR Design Control Document Fiche: 80563-001/80564-030 acn: 9408120078 Augusi 3.1994 Jack N. Fox, GE, Forwards Chapter 21 17x22 inch drawings to replace temporary 11 x17 drawings provided in 6/23/94 letter l Fiche: 80589-340/80589-342 acn: 9408150260 August 3,1994 Dennis M. Crutchfield, NRC, Updates guidance on preparation of design control document contained in 8/26/93 NRC letter Fiche: 80489-199/80489-202 acn: 9408080093 August 12,1994 Joseph F. Quirk, GE, Letter regarding fee regulations for design certification Fiche: 80626-356/80626-356 l acn: 9408180170 l l August 23,1994 Thomas H. Boyce, NRC, Forwards stafTcomments on sections of the drall Design Control Document (DCD) for the Advanced Boiling Water Reactor (ABWR) l Fiche: 8%92 254/80692-283 i acn: 9408290030 i C-1 NUREG-1503 Supplement 1

l l August 25,1994 ItW. Borchardt, NRC, Letter forwarding ten copies of ABWR FSER, NUREG-1503 (Vol.1 & 2) Fiche: 80728-28480728-287 acn: 9408300058 /sugust 30,1994 Joseph F. Quirk, GE, Submittal of revision to draft Introduction to the Design Control Document (DCD) Fiche: 80841-00140841-016 acn: 9409090214 September 7,1994 Joseph F. Quirk, GE, Forwards Rev 0 to Advanced BWR Design Control Document." Fiche: 81150-010/81182-010 acn: 9409190302 September 8,1994 Thomas 11. Boyce, NRC, Summary of Meeting on 8/23/94 to Discuss Staff Comments on Sections of the Draft Design Control Document (DCD) Fiche: 80936-098/80936-099 acn: 9409160074 September 20,1994 Joseph F. Quirk, GE, Letter requesting that FDA for the ABWR be issued for a period of fifteen years Fiche: 83349-352/83349-356 acn: 9503290326 September 27,1994 Dennis M. Crutchfield, NRC, Letter informing GE of results of resiew of the Design Control Document (DCD) for the Advanced Boiling Water Reactor (ABWR). Identified multiple discrepancies. Fiche: 81078-355/81078 356 acn: 9409300084 October 4,1994 R.W. Borchardt, NRC, Forwards stafTcomments on the Introduction to the Design Control Document (DCD) Fiche: 81245 350/81245-356 acn: 9410110200 October 5,1994 Joseph F. Quirk, GE, Letter discussing root cause and con ective measures for unidentified changes in the Design Control Document (DCD) Fiche: 81363-302/81363-304 acn: 9410180178 October 13,1994 Joseph F. Quirk, GE, Letter regarding GE being designated as the source for the DCD in the ABWR notice of proposed rulemaking Fiche: 83349-351/83349-351 acn: 9503290330 l l NUREG-1503 Supplement 1 C-2

Appendix C October 20,1994 William T. Russell, NRC, Letter responding to NEl letter of 9/20/94 to Chainnan Selin, discussing proposed design certification rules for ABWR and System 80+ Fiche: 81454 190/81454-191 acn: 9410270084 l i l October 21,1994 David T. Tang, NRC, Summary of meeting on 10A)6/94 on the ABWR Design Control l Document (DCD) Fiche: 81444-356/81444-360 acn: 9410260214 l October 28,1994 Joseph F. Quirk, GE, Forwards Rev 0 to "ABWR Design Control Document (DCD).* DCD comprised ofintroduction, certified design material & approved safety analysis material. Responses to NRC comments requested by 9/27/94 letter also enclosed. Fiche: 81680-001/81695-360 acn: 9411020070 November 1,1994 Steven P. Frantz, GE, Forwards description of proposed process for controlling changes to severe accident evaluations & explains bases for proposed process. Fiche: 81752-339/81752-352 acn: 9411140186 November 1,1994 Thomas It Boyce, NRC, Summary of meeting on 9/27 28/94 to discuss stafT comments on draf1 DCD for the ABWR. Fiche: 81622-001/81622-161 acn: 941107027i November 2,1994 Dennis M. Cmtchfield, NRC, Response to GE letter of 10/13/94 regarding designated source for the ABWR DCD in notice of proposed miemaking Fiche: 81658-178/81658-180 acn: 9411080085 November 4,1994 S.P. Frantz, GE. Forwards proposed revision to Section 3.8 of DCD Introduction for ABWR regarding GE meeting on i 1/02/94 Fiche: 81075-357/81075 358 acn: 9411140188 November 15,1994 Thomas 11. Boyce, NRC, Sununary of meeting on 941102 to discuss treatment of severe accidents in the Design Control Document (DCD) Fiche: 81822-303/81822-307 acn: 9411220426 November 18,1994 Joseph F. Quirk, GE, Forwards NEPA/SAMDA submittal for the ABWR. Attachment contains Technical Support Document for the ABWR Fiche: 83349-281/83349-350 acn: 9503290334 November 23,1994 William T. Russell, NRC, Forwards revised final design approval (FDA) for the ABWR Fiche: 81849-332/81849-337 l l acn: 9411250204 1 C-3 NUREG-1503 Supplement 1

December 8,1994 R.W. Borchardt, NRC, Fonvards staff comments on Revision 0 of ABWR Design Control Document. Fiche: 82061-157/82061263 acn: 9412190091 December 16,1994 N.E. Garber, " Plain Carbon Steel." Fiche: 82144-330/82144-345 acn: 9412270184 December 21,1994 Joseph F. Quirk, GE, Forwards NEPA/SAMDA submittal for the ABWR. Attachment contains Revision I to the Technical Support Document (TSD) Fiche: 83349-216/83349-280 acn: 9503290339 December.22,1994 Joseph F. Quirk, GE, Closure of ABWR FSER Confirmatoiy Items Fiche: 83353-001/83353-003 acn: 9503290346 December 22,1994 Joseph F. Quirk, GE, Submittal of Revision 1 to the ABWR Design Control Document Fiche: 82126-001/82127-225 acn: 94122800 % January 17,1995 Jack N. Fox, GE, Submittal of Revision 2 to the ABWR Design Control Document Fiche: 82484-281/82484-346 acn: 9501230195 January 26,1995 Jack N. Fox, GE, Provides information for closure of ABWR FSER Confinnatory item Fl.2.2 2 previously addressed in 12/22/94 letter Fiche: 82616-278/82616-279 ) acn: 9502020103 March 16,1995 R.W. Borchardt, NRC, Forwards environmental assessment (EA) for ABWR design certification and severe accident mitigation design alternatives (SAMDAs) Fiche: 83238-272/83238-293 acn: 9503240250 ) May 31,1995 Jack N. Fox, GE, Fonvards revised effective pages listing for ABWR DCD j Fiche: 82484-281/82484-346 acn: 9501230195 l February 7,19% Dennis M. Crutchfield, NRC, Discusses resolution of confirmatory items in NUREG-j 1503 regarding ABWR Fiche: 87081-079/87081-080 acn: 9602090004 March 22,19% Joseph F. Quirk, GE, Submittal of Amendment 36 (Revision 8) to the ABWR SSAR, and Revision 7 to the CDM i Fiche: 87680-001/87680-238 acn: 960328189 NUREG-1503 Supplement 1 C-4

~. l l April 3,19% Joseph F. Quirk, GE, Forwards marked up proposed changes to ABWR design description l Fiche: 88175109/88175-113 acn: 9605080030 i i April 16,19% Joseph F. Quirk, GE, Forwards proposed changes to ABWR DCD developed during j first-of-a-kind-engineering (FOAKE) Fiche: 88077-00I/88007-146 acn: 9604240082 April 25,19% Brian K. Grimes, NRC, Discusses review of changes to ABWR design documentation in GE letter of 4/16/96 Fiche: 88046-355/88046-358 acn: 9604290399 April 26,19% Joseph F. Quirk, GE, Forwards responses to staffletter regarding ABWR DCD change package Fiche: 88175185/88175-186 acn: 9605080090 May 8,1996 Thomas 11. Boyce, NRC, Summary of meeting on 5/1/96 to discuss changes to the ABWR Design Control Document (DCD) Fiche: 88270-163/88270-261 acn: 9605150239 June 10,19% Joseph F. Quirk, GE, Forwards changes to ABWR design documentation Fiche: 89075-163/89075-312 acn: 9607190012 August 30,1996 Joseph F. Quirk, GE, Forwards ABWR DCD, Revision 3, ABWR CDM, Revision 8, j and ABWR SSAR Amendment 37, Revision 9 Fiche: 89647-001/89649-180 acn: 9609090224 C-5 NUREG-1503 Supplement 1 l l

l ABWR DESIGN CONTROL DOCUMENT (DCD) REVISIONS N DCD REVISIONS DATE Revision 0 10/28S4 Revision i 12/22/94 1 Revision 2 1/17/95 l Revision 3 8/30/96 Revision 4 3/28S7 ABWR CERTIFIED DESIGN MATERIAL (CDM) REVISIONS CDM REVISIONS DATE Revision 0 8/31/92 Revision 1 (submitted with SSAR Amendment 32) 9/17/93 Revision 2 (submitted with SSAR Amendment 33) 12n/93 Revision 3 (submitted with SSAR Amendment 34) 3/31N4 Revision 4 (submitted with SSAR Amendment 35) 5/25N4 Revision 5 (submitted with SSAR Amendment 35 6/23/94 modification package (SSAR Revision 6)) Revision 6 (submitted with SSAR Amendment 35 7/20/94 modifiation package (SSAR Revision 7)) Revision 7 (submitted with SSAR Amendment 36) 3/22/96 Revision 8 (submitted with SSAR Amendment 37) 8/30/96 NUREG-1503 Supplement 1 C-6

ABWR STANDARD SAFETY ANALYSIS REPORT (SSAR) AMENDMENTS AMENDMENTS DATE AMENDMENTS DATE 1 3/29/88 19 12/13SI 2 6/2988 20 3/13/92 Drawings 4/8/92 3 12/3048 21 7/6/92 Drawings 7/13/92 Modification package 7/22/92 4 1/3149 22 9/18/92 Drawings 9/21/92 5 2/28/89 23 11/20 S 2 6 3/3149 24 In/93 7 6/249 25 1/29N3 8 7/2849 26 3/24/93 9 11/1749 27 4/23/93 10 3/28/90 28 5/14/93 11 5/2/90 29 5/28/93 12 6/4/90 30 7 4/93 13 7/3/90 31 (Revision 1) 7/28/93 14 10/2/90 32 (Revision 2) 9/17/93 Modification package 9/27/93 15 11/30/90 33 (Revision 3) 12n/93 Modification package 12/13S4 16 2/22/91 34 (Revision 4) 3/3I/94 4/16/91 drawings Modification package 4/11/94 17 6/28/91 35 (Revision 5) 5/25/94 Modification package (Revision 6) 6/23/94 Modification package (Revision 7) 7/20/94 18 10/11/91 36 (Revision 8) 3/22/96 37 (Revision 9) 8/30/96 C-7 NUREG-1503 Supplement 1 l.

Appendix D l CONTRIBUTORb TO THIS FSER SUPPLEMENT ) 1 I IMJdB RESPONSIBILITY Thomas Boyce Project Manager l Bernard Bordenick Legal Review William Burton Plant Systems Matthew Chiramal Instrumentation and Controls Angela Chu Technical Specifications Richard Emch Radiological Analyses George Georgiev Materials Engincenng JeffHolmes Fire Protection J:yLee Radiological Analyses ' James Lyons Section Chief, Plant Systems Janice Moore LegalReview l John Monninger Containment Systems and Severe Accidents Roger Pedersen - Radiation Protection Janak Raval Plant Systems Nicholas Saltos Probabilistic Risk Assessment Dino Scaletti Senior Project Manager James Stewart Instrumentation and Controls Frank Talbot Initial Test Program and Reliability Assurance Dale Thatcher Electrical Engineering l l George Thomas Reactor Systems . Jern Wilson Senior Policy Analyst l ' Ron Young Plant Systems ) l l l 1 1 l ~ l s [ ( k 2 r j D-1 NUREG-1503 Supplement 1 r

i Appendix L ERRATA TO THE ABWR FSER Pane. Column. Paranraoh Chanac Page 1 1,1st column,3rd paragraph Delete the sentence beginning with " Amendment 35", and replace it with " Amendment 35 (Revisions 5,6 and 7), the last revision of which was submitted to the Commission on July 20,1994, was the last amendment." Also, dele *e the sentence "GE submitted Revision 4 to the CDM on May 25,1994.", and replace it with "CDM Revision 6, submitted to the Commission on July 20,1994, was the last revision to the CDM." P:ge 1 1,1st column,4th paragraph In the first sentence, add the phrase "(Revisions 5,6 and 7)* aRer " Amendment 35." Also change the CDM revision from " Revision 4" to Revision 6." Prge 1-4, Ist column,3rd paragraph Delete the phrase "the core spray,." Page 17,2nd column. Ist paragraph Change "..NRC review and approval" to "NRC approval" Page 211, Table 2 2 Change the values for LPZ Chi /Q to the following: 0-8 hours 1.56E-4 8 24 hours 9.61E-5 1-4 days 3.36E-5 4-30 days 7.42E-6 Page 3 37,2nd column,2nd paragraph Change "RG l.61" to "RG 1.60." Pzge 3 38,2nd column,3rd paragraph Add parentheses around the phrase "the RO 1.60 shapes anchored to 0.3g peak ground accelerations" and delete the word "and" in front of that phrase. l Page 3-42,1st column,2nd paragraph In line 4, change "25.9m (85 ft)" to "25.7m (84.3 ft)." Page 3 42,2nd column, Ist paragraph In line 6, change "25.9m (85 ft)" to "25.7m (84.3 A)." ] Page 3-43,2nd column, Ist paragraph Change "(Case RZU)" to "(Case R2U)." Page 3 50,1st column,1st paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and approval prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Furthermore, any requested change., " Page 3 50,1st column,3rd paragraph In line 2, Change ".. finger pin closure," to *.. flanged closure," Also, in line 3, change "..was analyzed.. " to "..will be analyzed...", and in line 8 change "..were evaluated " to...will be evaluated." Page 3 50,2nd column,1st paragraph Change *.would involve an unreviewed safety question and, therefore, l require NRC review and approval prior to implementation." to "..would j require NRC approval prior to implementation." Also, delete the next sentence beginning with "Furthermore, any requested change.. Page 3 52,1st column,4th paragraph Change the second sentence to read ".. finite element method will be used for the analysis and the design will be accomplished. " i L-1 NUREG-1503 Supplement I

I Appendix L i l Page 3 53,2nd column,last paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and approval prior to implementation." to *..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Furthermore, any requested change..

• Page 3-55,2nd column,2nd paragraph Change "NASTRAN" to "STARDYNE" computer code.

j Page 3-57, Ist column,3rd paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and approval prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next l sentence beginning with "Furthermore, any requested change.. " j Page 3 80,1st column,2nd paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and approval prior to implementation " to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Furthermore, any requested change.. " Page 3-87,2nd column,2nd paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and approval prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Furthermore, any requested change.. " Page 3-123,2nd column, last paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and approval prior to implementation." to "..would reqrire NRC approval prior to implementation." Also, delete the next sentence beginning with "Furthermore, any requested change.. " Page 4 1,2nd column,3rd paragraph in the first sentence, change ".. prior NRC review and approval." to l " prior NRC approval." Page 4 2,2nd column,3rd paragraph Change "..with NRC review and (gproval." to "..with prior NRC approval." Page 4 3,1st column,last paragraph In the second sentence, change "..with NRC review and approval." to "..with prior NRC approval." Page 6-12,2nd column,3rd paragraph in line 3, change " swell" to " condensation." Page 6-39,2nd column, Ist paragraph Change "(150 psig)" to "(150 psia)." Page 6-39,2nd column,4th paragraph Modify the sentence beginning with "SSAR Table 6.3 1. " to read "SSAR Table 6 3-1 states that the rated HPCF flow of 12,11311 min (3,200 gpm) will be attained at a differential pressure (between the reactor vessel and the air space of the companment containing the water source for the pump) of approximately 689 kPad (100 psid), Page 6-41,1st column,2nd paragraph In line 3, after ".. primary containment.. ", add "(except the low Pressure Flooder loop A check valve which is outside containment because it is connected to the feedwater line)" l l l NUREG-1503 Supplement 1 L-2

I Appendix L Page 6-53,2nd column,2nd paragraph Replace the paragraph with the following: "The HPIN system has both non-safety-related and safety-related portions. The non-safety-related portion provides a continuous nitrogen supply to all pneumatically operated components in the primary containment during normal operation. This nitrogen is supplied by the nitrogen gas evaporator / storage tank via the makeup line from the ACS The safety-related portion has two independent divisions. Each division contains a safety-related emergency i stored nitrogen supply capable of supplying 100 percent of the requirements of the division being seniced. Normally, nitrogen gas for l the safety-related portion is also supplied by the nitrogen gas evcporator/ storage tank via the makeup line from the ACS. If that supply is not available, then it will be supplied from IIPIN gas storage bottles. There are ticlines between the non-safety related portion and each division of the safety-related portion. Each tieline has a motor-operated shutoff valve." Page 7-1, Ist column,3rd paragraph Change " Chapter 1. " to " Chapter 21. " Page 7 13, Ist column,1st paragraph In the indented paragraph, change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested changes.. " Pege 713,2nd column,3rd paragraph Change ". GE NUMARC, " to "..GE NUMAC. " Page 719,2nd column,4th paragraph Change "For manual scram, " to "For manual trip test, Page 7-22, Ist column,2nd paragraph Change "MPL ABBE-4080. " to *MPL-A32-4080. " Also, change " shall be a synchronous between. " to ". shall be asynchronous between.. " Page 7 23, Ist column, Ist paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested changes.. " Page 7-27,1st column,2nd paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested changes.. " Page 7 27,2nd column,5th paragraph Change " Figure 7. A.2. I in SSAR Appendix 7A shows.

• to " Table 7A-1 in SSAR Appendix 7A lists. " Also change "This drawing was listed. "

to "This table was listed. " Page 7 28, Ist column,5th paragraph Change "There are a total of 2 I input channels. " to "There are nominally a total of 21 input channels.. " Page 7-28,2nd column, Ist paragraph in subparagraphs a. d, change " interface boards" to " interfaces". Page 7 29, Ist column,2nd paragraph Change "Section 7.1.2.6" to "Section 7.1.2. l.6." l Pcge 7-29,2nd column, Ist paragraph in subparagraph 1., change "one set of scram pilot valve solenoids" to " l one half of the scram pilot valve solenoids." L-3 NUREG-1503 Supplement 1

Appendix L Page 7-30, Ist column,2nd paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested changes.. " Page 7 32, Ist column,1st paragraph Change " Division III 480 Vac source" to " Division II 480 Vac source." Page 7 32, Ist column,4th paragraph In the last two lines, change "The A.1WR design requires the coils of both scram pilot solenoid valves of each QD. " to "7he ABWR design requires both coils of the scram pilot solenoid valve of each CRD.

• Page 7 32,2nd column,2nd paragraph Change " Portions of the RPS and SSLC (in particular, the RMUs) are located.
• to " Portions of the RPS and the RMUs are located. " Also, delete the word " control
• from " control building" in the sentence.

Page 7-33,2nd column,2nd paragraph in subparagraph 1., change " commonality" to " correlation" and "an error" to "a common-cause error." Page 7 34, Ist column,2nd paragraph Change "DTM" to " logical." Page 7-41,2nd column, last paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested changes.. " Page 7-42,2nd column 2nd paragraph Change " Table 7B. I" to "Apperxlix 7B." l Page 7-44,1st column,4th paragraph Change "..would involve an unreviewed safety question and, therefore, l require NRC review and acceptance prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested changes.. " Page 7-44,2nd column, last paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would l require NRC approva! prior to implementation." Also, delete the next I sentence beginning with "Any requested changes.. " Page 7-45, Ist column,2nd paragraph Change "DMil-4270* to "1123-4010." Page 7-46, Ist column,4th paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested changes.. " Page 7-48,1st column,3rd paragraph Change each "and" in the paragraph to "or" in all 5 places where it occurs. Page 7-51,2nd column,3rd paragraph Change " reactor flow control" to'" recirculation flow control." Also change " dome pressure and low reactor" to " dome pressure or low reactor." Page 7-52,1st column,3rd paragraph Change the last sentence to read "The ARL initiation variables in the RFC system also initiate a. " NUREG-1503 Supplement 1 L-4

i Appendix L Page 7-56,2nd cohimn,3rd paragraph Change the first sentence to read "Three SRNM channels per division provide input to each of SSLC Divisions I and III, and two SRNM channels per division provide input to each of SSLC Divisions 11 and IV." Page 7 57, Ist column,3rd paragraph After the sentence ending in ". IEEE-279 1971," add "The OPRM is a functional subsystem of the APRM. Each OPRM receives the identical LPRM signals from the corresponding APRM channels as inputs." Page 7-57,1st column,4th paragraph In the second sentence, put a period after the word " function" and delete the remainder of the sentence. Also, delete the sentence beginning with "The SRI function. " Also, aner the phrase ". four separate inputs," insert "(combined with other APRM trip outputs)." Page 7-58, Ist column,5th paragraph Delete the first three sentences and replace with "The SPTM system is a four division system consisting of temperature sensors at eight sensor locations around the circumference of the suppression pool, with two groups of sensors at each sensor location. Each group has four sensors located at different elevations in the suppression pool. The signal processing for the SPTM system is performed by EMS and SSLC microprocessors, which are powered by four divisionally separated electrical buses." Page 7-60, Ist column In the title for section 7.7.1.5, change " Generator" to " Generation." Page 8-1,2nd column,4th paragraph Change " gas turbine generator" to " combustion turbine generator." Page 8-5,2nd column,4th paragraph In the second sentence, delete the phrase " routed on opposite sides of the room and will be." Page 8-13,2nd column,2nd paragraph Change the sentence beginning with " Simultaneously,. " to read " Simultaneously, a timer will be started, allowing the operator to take necessary corrective action (The actual set points will be established as part of an overall system voltage and load analysis)." Also modify the following sentence to read "After the time delay, the feeder breaker affected by the degraded vohage will be tripped." Page 8-14,1st column,2nd paragraph In the bulletized section, delete the phrase "for 5 minutes with voltages at the load at 70 percent of the nominal voltage rating", and insert the phrase "for degraded vohages below 90% for the time period established in the load analysis for the degraded voltage protective time delay." Pcge 9-4, Ist column,2nd paragraph Change the sentence that begins with "Specifically, the system includes two. " to read "Specifically, the system includes two 100 percent capacity circulating pump =, two 50 percent capacity heat exchangers, two 100% filter /demineralizers, two post demineralizer strainers,. " Page 9-6, Ist column,1st paragraph Delete the prefix "non " from "non-safety-related suction portion." Page 9-22, Ist column,2nd paragraph in the last sentence, change " train" to "section." Page 9-23,1st cohimn,1st paragraph At the end of the sentence add ". and with a design UIIS temperature of 35 "C (95 *F)." Pege 9-24, Ist column,4th paragraph Delete the phrase ", the portions of the system that are part of the secondary containment boundary,". L-5 NUREG-1503 Supplement 1

1 i Appendix L Page 9-26, Ist column, Ist paragraph In line 6, change " ..of one refrigerator and pump, a surge tank

• to "..of two refrigerators and two pumps, a surge tank.. "

Page 9-26,2nd column,1st paragraph In the last sentence, change " refrigerant" to "IIECW system." Page 9-28, Ist column,4th paragraph In the second sentence, delete the phrase " corresponding disision of the." Page 9-30,2nd column,2nd paragraph In the second sentence, change " heat exchangers" to " pumps." l Page 9-33, Ist column,6th paragraph At the end of the sentence ending in ". safety-related portion of the system." add "in the event that nitrogen gas is not available from the ACS." Page 9-46, Ist column, top paragraph In line 2, change "..in which the exhaust fan is stopped, the recirculation damper is closed, and the exhaust bypass damper is opened." to "..in which both exhaust fans are started at high speed and the recirculation damper is closed." j l Page 9-47,2nd column,2nd paragraph In line 8, change ".. closing the exhaust fan, and opening the exhaust fan bypass damper to allow.. " to "..and starting both exhaust fans in conjunction with a supply fan to allow.. " Page 9-49,2nd column, Ist paragraph in line 10, delete "(bag-type filter)", and in line 13 delete "(bag-type;." Also, in line 14, change the sentence beginning with "The supply system.. " to read "The supply system consists of three 50-percent capacity air handling units consisting of a filter, a cooling and heating coil, and supply fans,. " Also, in line 18, delete " bag-type." l l Page 9-51, I st column,4th paragraph In the first sentence, delete "..FCUs and four.. ", add a comma aner the word " coil", and add "..and filter as required." to the end of the sentence. Also, in line 6, change "10" to "6." Page 9-52,1st column,2nd paragraph In line 5, put a period aner the word " fan," and delete "..and an electric I heater." Page 9-52,1st column,2nd paragraph Change the DG room temperature from "below 45 "C (113 "F)" to below 50 "C (l 22 "F)." ( Page 9-52,1st column,2nd paragraph In the sixth line from the bottom, delete the word "and," and add *..and starting the smoke removal fans in conjunction with the supply fans." to the sentence ending in "..to purge the afTected area." Page 9-53, Ist column,1st paragraph Change the DG room temperature from "below 45 "C (113 "F)" to "below 50 *C (l 22 "F)." Page 9-65,1st column,1st paragraph In line 5, put a period aner the word " areas" and delete the rest of the paragraph. Page 9-65,2nd column, Ist paragraph In line 2, delete " secondary containment and the " Also, in line 4 change ..these HVAC systems, they wilL" to...the llVAC system, it will.. " Page 10-2,1st column,2nd paragraph In the second sentence, change " disk / pump" to " disk / dump." Page 10-2,1st cohunn,4th paragraph in the first sentence, delete the rest of the sentence aner the phrase ". will be provided" and insert "in accordance with the Boiling Water Reactor Owners Group turbine surveillance program." NUREG-1503 Supplement 1 L-6

Appendix L Page Il-2,2nd column,2nd paragraph In the second sentence, change " filtered in one or two" to " filtered in one of two." Page 14-28, I st column,4th paragraph Change "..would constitute an unreviewed safety question, and therefore, would require NRC review and approval prior to implementation of the change." to "..would require NRC approval prior to implementation of the change." Page 14-30,2nd column,last paragraph in the first sentence, change "..would constitute an unresiewed safety question." to "..would prior NRC approval." Also, in the second sentence, change ". prior NRC review and approval." to ".. prior NRC approval." Page 14-31,1st column,2nd paragraph Change ".. prior NRC review and approval.. " to ".. prior NRC approval.. " Page 14-31,2nd column,last paragraph Change " basis configuration" to " basic configuration". Page 14-37,1st column,2nd paragraph Change "..would constitute an unreviewed safety question, and therefore, would require NRC review and approval prior to implementation of the change " to "..would require NRC approval prior to implementation of the change." Page 14-37,2nd column, I st paragraph Change "..would constitute an unreviewed safety question, and therefore, would require NRC review and approval prior to implementation of the change " to "..would require NRC approval prior to implementation of the change." Page 14-40,1st column,last paragraph Change "..would constitute an unreviewed safety question, and therefore, would require NRC review and approval prior to implementation of the change." to "..would require NRC approval prior to implementation of the change " i Page 14-49,2nd column,1st paragraph Change ". constitutes an unreviewed safety question and, therefore, must be submitted to the NRC for review and approval prior to implementation." to "..would require NRC approval prior to implementation of the change " Page 14-52,1st column,3rd paragraph Change ". constitutes an unreviewed safety question and, therefore, must be submitted to the NRC for review and approval prior to implementation." to "..would require NRC approval prior to implementation of the change." Page 14-55,1st column,1st paragraph Change ". constitutes an unreviewed safety question and, therefore, must be submitted to the NRC for review and approval prior to implementation " to "..would require NRC approval prior to implementation of the change." Page 15-3, Ist column,4th paragraph in the first and last sentences, ch'ange "38 *C (68"F)" to "55.6 *C (100"F)" Page 15-7, Section 15.4, line 8 Change "(25 rem), or a whole body dose of no more than 3000 mSv (300 rem)," to "(25 re n) whole body dose," Page 15-11, Table 15.3,3rd parameter Change the value for the mass of primary coolant released through small line from "5.5E+3 kg (1.2E+4 lb)" to "1.4E+4 kg (3.1E+4 IL)". L-7 NUREG-1503 Supplement 1

Appendix L Page 15-12, Table 15.4,9th parameter Add "* *" aAer " Standby gas treatment system" and add, at the bottom of the Table, "" No credit given for lower flow rate after 20 minutes pressure drawdown time." Page 15-12, Table 15.4,13th parameter Change the value for suppression pool water volume from "3.785E46 m' (1E+6 gal)" to "3.625E+3 m'(9.576E+5 gal)". Page 1513 Table 15.6,4th parameter Change the value for condenser volume from "9.85E+8 cc (3.47E+4 A')* to "6.23E+3 m'(2.2E+5 A')". i Page 1513, Table 15.6,7th parameter Change the value for main steamline thickness from "2.5 (1.0)" to "3.6 (1.4)" and the value for drain line length from "610 (240)" to "7160 (235)". Page 15-14, Table 15.8,7th parameter Change the value for Kr-85 inventory released from damaged rods from "35%" to "30%". Page 15-14, Table 15.8,8th parameter Change the values for lodine fraction from "0.25" to "0.25%" and *0.75" to "99.75%". Page 15-15, Table 15.9, I st parameter Change the value for control room free volume from "7000,m'(2E+5 ft')* to "5509 m' (1.95E+5 A')". Page 1515, Table 15.9,2nd parameter Change the value for the following recirculation rates: filtered intake from "1.8 m'/sec" to "0.994 m'/sec"; filtered recirculation from " 0.8 m'/sec" to "0.47 m'/sec"; and filter ellicacy from "95%" to" 99%." Page 15-15 Table 15.9,3rd parameter Change "1.0 E-1 A3/ min" to"1.0E+1 A3/ min." Page 15-15, Table 15.9,6th parameter Add a ")" after " sectors" and change the meteorology values to the following: 00-008 3.1OE-3 sec/m'(8.8E 5 sec/A') 08-024 1.83E 3 sec/m'(5.2E-5 sec/ft') 24-096 1.16E-3 sec/m'(3.3E-5 sec/ft') J %-720 5.12E-4 sec/m'(1.5E-5 sec/ft') Page 15-15, Table 15.9. 7th parameter Change to Iodine protection factor from "27" to "100". Page 15-15 Table 15.9, i1th parameter Change the thyroid doses to control room personnel to the following: 00-008 7 (0.7) 08-024 7 (0.7) I 24-0 % 37 (3.7) l l 96-720 52 (5.7) j Total 103 (10.2) l Page 15-16,1st column,2nd paragraph Change "SSAR 15.4.9.6" to "SSAR Section 15.4.10.6". Page 1516, ist column,3rd paragraph Change "SSAR 15.4.9.2" to "SSAR 15.4.10.2". Page 15-16,2nd column,3rd paragraph Change "5448 kg (l 2,000 lbs)" to "14,000 kg (30,000 lbs)" in 2 places. Page 15-19,1st column,1st paragraph Change "28 BWR sites" to 28 geological sites". NUREG-1503 Supplement 1 L-8

Appendix L Page 18-28,1st column,2nd paragraph Change "..by the COL applicant would involve an unresiewed safety question and, therefore, require NRC review and acceptance prior to implementation." to "..would require NRC approval before implementation." Also, delete the next sentence beginning with "Thus, any change.. Page 18-35,1st column,3rd paragraph Change "..would involve an unreviewed safety question and, therefore, require NRC review and acceptance before implementation." to "..would require NRC approval prior to implementation." Also, delete the next sentence beginning with "Any requested change..

• Page 19-6, I st column, last paragraph Change "3.8E-8" to "3.2E-8" in line 8.

Page 19-6,2nd column,2nd paragraph Change "fussell-vesely" to "Fussell-Vesely". Page 19-6,2nd column,2nd paragraph Change "when shutdown occurs" to "during shutdown" in line 10. Page 19-6,2nd column, Ist paragraph Change "The analysis identified scram function and its attendant equipment.. " to "The analysis identified multiplexed safety system logic and control equipment..

• in line 5.

Page 19-7,2nd column,1st paragraph Change "SSAR Section 19.9" to "SSAR Section 19.8". Page 19-8,1st column,2nd paragraph Delete the word "many" in line 3. Page 19-11, Ist column, Ist paragraph Add " Appendix D" to the list of appendices in line 4. Page 19-11,2nd column,5th paragraph Change "RWCU" to "CUW". Page 19-12, I st column,1st paragraph In the last sentence, change "(See SSAR Section 19.9) to (See SSAR Section 19.11) Page 1912,1st column,4th paragraph in line 2, add "or in the emergency procedure guidelines" after "(See SSAR Section 19.9)" Page 19 12,1 st column, last paragraph In the second sentence, change "..is documented in the table. " to "..is documented in SSAR Section 19.8.. " Page 19-13,1st column, top paragraph Change "3.8E-8" to "3.2E-8" in line 4. Page 19-13,2nd column, I st paragraph In the second line, change "fussell-vesely" to "Fussell-Vesely". Page 19-14,1st column,3rd paragraph In the ninth line, change ".. diverse (resulting from the.. " to " diverse (compared to the.. " L-9 NUREG-1503 Supplement 1

Appendin L Pages 1918 and 1919, Table 19.17 Move or insert the "/DP** between " LOP ** and "APW** or "/APW** in sequences 5,9,18,20,3,7,17,19,11,12,22,24,25,26,4,6,8, and 10. In Damage Class 1B, change the seismic IICLPF value from *0.62" to "Of>0" and the seismic / random IICLPF from "None" to "0.50g

• 1.6E-3" for both sequence 3 and the total. In sequences 17 and 19, delete the

"*V2" at the end of the sequence description. In sequence 7 and the total, change *0.62g" to "0.60g " In sequence 11 and 12, delete the "AW*." In sequences 21,13,22, and 24, change "PC" to "PC/." Modify all sequence numbers by subtracting 2. Page 19-19,1EGEND Change "PC= Failure of SRVs to Close" to "PC= Failure of SRVs to Close Given LOP with Scram" and add "PC!= Failure of SRVs to Close Given LOP without Scram" Page 19-20, Table 19.1-8 In Damage Class 1B-2, add "SW Pump House (0.60g) or" after " Diesel Generator (0.62g) or." Also in Damage Class 1B-2, in the second column, delete " Fire Pump (0.62g)" and add "FW Tank (0.51 g) or FW Pump (0.51 g) or Injection Valve (0.50g) or FW Piping (0.50g) or Manual Valve (0.50g)." Also, partition the column titled "Scismic/ Random " Dominant Cut Sets" into 2 subcolumns. In Damage Class 1B-2, add "FW System (0.50g)" to the first subcolumn and " Support System (1.6E-3)" to the second subcohunn, and in Damage Class IC, add " Fuel Assembly (0.62g)" to the first subcolumn and "SRVs Close (1.0E-1)" to the second subcolumn. Page 19-21, Table ',9.1-8 In Damage Class 1D, delete "or HPCF Pump (0.62g)" and "V2 (0.62g)." Also, in the 3 places where it occurs in Damage Classes 1D and 1 A P through 1 E-P, change " Fire Pump (0 62g)" to "FW Tank (0.51 g) or FW Pump (0.51 g) or Injection Valve (0.50g) or FW Piping (0.50g) or Manual Valve (0.50g)." Also, change "LPCF Pump (0.56g)" to "LPCF Pump (0.62g)." In addition, in Damage Class 1 V (Incl 1 V-P), partition the column titled " Seismic / Random " Dominant Cut Sets" into 2 subcolumns, and add " Fuel Assembly (0.62g)" to the first subcolumn and "LPL (1.0E-l 2)" to the second subcolumn. Page 19-25,2nd column, Ist paragraph Change "7E-9" to "2E-8" in line 4. Page 19-25,2nd column,4th paragraph Change "RSW" to "RCW" in lines 7 and 8. Also, clunge "2E-9" to "9E-9" l in the last line. Page 19 26,1st column,2nd paragraph Change "RSW" to "RCW" in the 7th line from the bottom. Page 19-26,2nd column,2nd paragraph In the second sentence, delete "and TSW" and "in both systems", add "CWS" between "close" and " isolation." Also, in the last sentence, delete "the TSW and." l l Page 19-26,2nd column,3rd paragraph In line 4, change ".. reactor water service water /RCW.. " to ".. reactor service water.. " Also, in the last sentence, change "RCW" to "RSW". Page 19-27, I st column,2nd paragraph In the first sentence, delete "and TSW" and "..and the two TSW pumps.. " In the second sentence, change ".. trip all five pumps, and ek>se all 1 isolation valves in both systems." to ".. trip the CWS pumps, and close all CWS isolation valves." l NUREG-1503 Supplement 1 L-10 l i

Appendix L Page 19 31, ist colunm,2nd paragraph Add to the end of the paragraph "It is unlikely to lose RPS logic power because the power for the safety system logic control is backed up by batteries." Page 19-31, Ist column,6th bullet Delete "the reactor water cleanup system," in line 2. Page 19-31,2nd column,3rd bullet Change "..a feedwater pump or other pump.. " to "..any of several pumps.. " in line 3. Page 19-33,2nd cohunn, Ist paragraph Add "with no active heat removal" after "about 20 hours from accident initiation." Page 19-35,2nd column,2nd paragraph In the last sentence, change ".. chances of a flooding a following.. " to ".. chances of flooding the cavity following.. " b Page 19-37,2nd column, I st paragraph Change " Table 19K.11-1" to " Table 19K-4". Page 19-37,2nd column,2nd paragraph Change " Table 19K.3-1" to " Table 19K-1" and change " Table 19K.3-2" to " Table 19K-2". Page 19-38, Ist column,2nd paragraph In line 10, delete "..and the NRC..

• Page 19-38, Ist column,3rd paragraph Add the containment overpressure protection system to the last sentence.

Page 19-41, Section 19.1.3.9.2 Replace "RWCU" with "CUW" in all places where it occurs. Page 19-41,2nd column,1st paragraph In line 8, after...is actuated." add "GE also added a remote manual shutoff valve to the CUW system inside containment to allow the operator to manually isolate the system from the control room." Page 19-47, I st column, I st paragraph in the last sentence, delete " relief valves," Page 19-49,1st column,1st paragraph Add the phrase "If the AC independent water addition system is used, to the sentence beginning with "The water from the containment spray system.. " Page 19-49, I st column,3rd paragraph In the first sentence, change "will" to "may". Page 19-51,2nd column,2nd paragraph In the third sentence, change the units of minimum flow rate from "10.8 Kg" to "10.8 Kg/sec". Page 19-54,2nd column,2nd-4th paragraphs Delete the word "an" from the phrase "an uncertainty analyses" in 3 places where it occurs. i Page 19-55, Ist column,2nd paragraph in the last line, change " system" to " systems". Page 19-58,2nd column,4th paragraph in the third sentence, add the phrase "from the AC independent water addition system" to the end of the phrase "..from 35 to 31 hours if drywell sprays are available." Page 19-59, Ist column,2nd paragraph In the fourth sentence, add the phrase "..with no active heat removal" to i the end of the phrase "..on the order of 15 to 20 hours, Page 19-60,2nd column,2nd paragraph Change the units from " liters /m" to " liters / min" in two places. L-11 NUREG-1503 Supplement 1

Appendix L Page 19-70,2nd column,last paragraph In the fiflh line, change "..should be Table.. " to "..should be used in Table..

• Page 19-73,1st column,3rd paragraph in the first line, change "A.2 above" to "19.2.6.2.1.2 above".

Page 19-73,2nd column, last paragraph In the last line, change "..Section 2D of this report." to "Section 19.2.6.2.4 of this report." Page 19-74,1st column,2nd paragraph Delete the phrase "In a facsimile dated September 7,1993,.

• Also delete the last sentence of the paragraph.

Page 19-74, Ist column,3rd paragraph In the first sentence, change "five operable penetrations" to "six operable penetrations" and "two pressure unseating hatches" to "three pressure-unscating hatches." Also, in lines 5 and 10, change "three pressure-unseating hatches" to "four pressure-unseating hatches." Page 19-74,2nd column, first paragraph Change "As discussed in Section 28.. " with "As discussed in Section I9.2 6.2.2.." Page 19-76,2nd column,3rd pat agraph At the beginning of the paragraph, add "GE indicated that the probability of a loss of shutdown cooling due to loss oflogic power (which can occur in operating BWRs if the RPS MGs trip) has been significantly reduced since the shutdown cooling isolation logic (i.e., SSLC) power is backed up by batteries. Nonetheless, Page 19-77,2nd column, first paragraph In the eleventh line, change " containment" to "RPV". Page 19-78,2nd cohunn, first paragraph in the fourth line, change "..in such a way " to "..in some other way." Page 19 78,2nd column,3rd paragraph In the last sentence, change "..this COL action item.. " to ".. COL Action item 19.9.23.. " Page 19 79, Ist column,3rd paragraph In the 3rd line from the bottom, change "The staff, therefore, requires.. " to "The stafTs position was that it required.. " Page 19 80,1st column,1st paragraph in line 2, change "the same" to "similar". Page 19-80, Ist column,3rd paragraph In line 4, delete the word "the" at the end of the line. Page 19-80,2nd column, I st paragraph In the last sentence, change "RIIR" to "the". Page 19-81,1st column, top paragraph In line 4, add "(Section 19.9.11)" after the word " item." Page 19-82,1st column,3rd paragraph In the ninth line, add " adjacent" afler "a second". Page 19-82,2nd column, top paragraph In line 6, change "is" to "are." Page 19-82,2nd column,2nd paragraph In the fifth sentence, change "..more than one control rod.. " to "..two adjacent control rods.. " ~ Page 19-82,2nd column,3rd paragraph In the first sentence, change "..more than one control rod.. " to "..two adjacent control rods.. " Page 19-82,2nd column,4th paragraph In line 8, change "..a feed pump or other pump.. " to "..one of several pumps.. " NUREG-1503 Supplement 1 L-12

Appendix L Page 19-83,1st column,2nd paragraph Change "..a second control rod. " to "..two adjacent control rods.. " Page 19-84, I st column,4th bullet Add " Water in the. " to the beginning of the sentence and delete %xfing level". Page 19-87,1st column,3rd paragraph la the last sentence, change "..this COL action item.. " to ".. COL Action Item 19.9.25.. " Page 19-88,1st column,2nd paragraph in the last sentence, change "..this COL action item.. " to ".. COL Action Item 19.9.11.. " Page 19-88,1st column, last paragraph in line 12, change ".. reduced inventory.. " to ".. shutdown. " Page 20-47,1st column, last paragraph Change the maximum snowload from "2.354 kPa (.341 psi)" to "2.394 kPa (0.35 psi)". Page 20-78,1st column,4th paragraph Change "..SSAR Table 19P-1." to ".. Attachment A of the " Technical Support Document (TSD) for the GE ABWR," Revision I, dated December 1994,in Table A-1. The TSD was submitted to the staffin a letter from GE to the NRC of December 21,1994." Page 20-79, Note 1 Change *..SSAR Table 19P-1." to ".TSD Table A-l." Page 20-81,1st column,2nd paragraph Change "..SSAR Table 19P-3." to "..TSD Table A-3." Page 20-81,1st column,4th paragraph Change *..(SSAR Table 19P-3). " to "..(TSD Table A-3).. " Page 20-81,2nd column,4th paragraph Change "..SSAR Sections 19P.3 and 19P.4, " to "..TSD Sections A-3 and A-4, Page 20-82, Table 20.5.1-3 Item Ic., change the Cost / Person-Sieven Averted value for improved maintenance procedures / manuals from "l880 (18.8)" to "I 870 (18.7)". Page 20-82, Table 20.5.1-3 Item 2a, change the Person Sievert Averted value for passive high pressure system from "0.00138 (0.138)" to "0.00069 (0.069)", and the Cost / Person-Sv Averted value from "1270 (12.7)" to "2530 (25.3)". Page 20-82, Table 20.5.1-3 Item 3b., change the Person-Sievert Averted value for increased containment pressure capacity from *0.00020 (0.02)" to "0.0016 (0.16)". Page 20-82, Table 20.5.1 3 Item 3c., change the Person-Sievert Averted value for improved vacuum breakers from "0.0000003 (0.00003)" to "0.0000004 (0.00004)". Page 20-82, Table 20.5.1-3 Item 7a., change the Cost / Person Sv Averted value for drywell head flooding from "1700 (1.7)" to "170 (1.7)". Page 20-82, Table 20.5.1-3 Item 9b., change the Cost / Person-Sv Averted value for altemate pump power source from "1740 (17.4)" to "1730 (17.3)". Page 20 82, Table 20.5.1 3 Item 13a., change the Cost / Person-Sv Averted value for reactor building sprays from "5900 (5.9)" to "590 (5.9)". Page 20-84,2nd column,1st paragraph Change "..SSAR Section 19P.4, " to "..TSD Section A.4,. " L-13 NUREG-1503 Supplement 1

/ Appendix L Page 20-84,2nd column,2nd paragraph Change ". 3E-7 person-Sv (0.00003 person-rem)." to ". 4E-7 person-Sv (0.00004 person-rem)." Page 20-85, Table 20.5.1-4 Change the Person-Sv averted value for passive high pressure system from "0.00138 (0.138)" to "0.00069 (0.069)". Page 20-85, Talle 20.5.1-4 Change the Person-Sievert Averted value for increased containment pressure capacity from "0.00020 (0.02)" to "0.0016 (0.16)". Page 20-85, Table 20.5.1-4 Change the Person-Sv averted value for improved vacuum breakers from "0.0000003 (0.00003)" to "0.0000004 (0.00004)". Page 20-87,1 st column,1 st paragraph Change "..SSAR Section 19P.I.3." to "..TSD Section A.I.3.1." Also, change "..SSAR Section 19P.S.. " to "..TSD Section A.S.. " Pages B-12 to B-15 Reorder the listed Regulatog Guides in numerically ascending order. Page C-1 Replace the entire page with Pages C-1 through C-3 of this supplement. Page C-25, first entry Delete the first enty. Page C-46 In the 3rd entry dated May 5,1992, change "J. Duncon" to J. Duncan." Also, delete the last enty on the page regarding a May 18,1992 letter from C.B. Brinkman. Page C-117,3rd entry Change "SSAR Amendment 5" to "SSAR Amendment 35". Page C-117,6th entry Insert the Fiche number "80331:057-80331:083". Page D-1 Add "C. Thomas, Electrical Engineering", "R. Pichumani, Geoscience", and *C. Li, Plant Systems" to the list of FSER contributors. Page E-2, Ist column, Ist paragraph Change "NE-3322.6" to "NE-3324.8". Page K-1, Ist column,last paragraph In the eleventh line, change the note to read "EDGs cannot power condensate or feed pumps; CTGs cannot power feedwater pumps)". l Page K-1, Ist column, last paragraph In the last line, add "(Section 19.9.19)" after " COL applicant". l Page K-3,2nd column, last paragraph In the last sentence, change "..the circulating water and turbine service water pumps and close isolation valves in both systems." to "..the circulating water pumps and close CWS isolation valves." Also add "They will also alert the operators to other floods from TSW". l Page K-4, Ist column,last paragraph in the last sentence, change "..the circulating water and turbine service water pumps and close isolation valves in both systems " to "..the circulating water pumps and close CWS isolation valves. The TSW system must be isolated manually within one hour to prevent the water from reaching safety equipment." NUREG-1503 Supplement 1 L-14

Appendix L Page K-7, Ist column,2nd paragraph In subparagraph (2) pertaining to SRVs, insert a period after ".. Class 2" and create a new subparagraph (3) with the remainder of the subparagraph. Page K-7, Ist column,4th paragraph In subparagraph (2) pertaining to LOCAs outside containment, change "RWCU" to "CUW". L-15 NUREG-1503 Supplement 1

- - - - - - - - - - - - - - - - - - - - - - ~ - - - ~ ~ ~ - ' ~ ~ ' ~ NRC FORM $$$ U.s. NUCLEAR REGULATORY Commission

1. REPORT NUMBER Q49)

(Assigned by NRC. Add Vol. Supp., Rev., 7$* BIBLIOGRAPHIC DATA SHEET ""d ^dd*"* """**# *"W rsee meucaans on me =wse)

2. TITLE AND SUBTITLE NUREG-1503 Supplement No.1 Final Safety Evaluation Report Related to the Certification of the Advanced Boiling Water Reactor Design 3.

DATE REPORT PUBUSHED Supplement No.1 l "C" YEAR May 1997

4. FIN OR GRANT NUMBER

5. AlJTHOR(S)

6. TYPE OF REPORT Regulatory

7. PERIOD COVERED (Jncsusive Desse)

August 1994-April 1997

8. PERTORMING ORGANIZATlON - NAME AND ADORESS tWNRc powde Omm. omeo a Rega. U S Nucteer Requiewy comnssa, and moolmg address. Icontecw, prowde nome and nwomg address)

Division of Reactor Program Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

e. SP.ONS.ORING ORGANIZATION - NAME AND ADORESS tWNRC. type "Same as above9 dcontecear. ponte !*RC Omm, o me of Regm U S Nuclev Requieky Comns e,

ng s) Division of Reactor Program Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Io. SUPPLEMENTARY NOTES Docket Nos. 52-001 and STN 50-605 Project No. 671

11. ABSTRACT (xo words or tesa)

This report supplements the final safety evaluation report (FSER) for the U.S. Advanced Boiling Water Reactor (ABWR) standard design. The FSER was issued by the U S. Nuclear Regulatory Commission (NRC) staff as NUREG-1503 in July 1994 to document the NRC staffs review of the U.S. ABWR design. The U.S. ABWR design was submitted by GE Nuclear Energy (GE) in accordance with the procedures of Subpart B to Part 52 of Title 10 of the Code of Federal Regulations. This supplement documents the NRC staffs review of the changes to the U S. ABWR design documentation since the issuance of the FSER. GE made these changes primarily as a result of first-of-a-kind-engineering (FOAKE) and as a result of the design certification rulemaking for the ABWR design. On the basis of its evaluation, the NRC staff concludes that the confirmatory issues in NUREG-1503 are resolved, that the changes to the ABWR design documentation are acceptable, and that GE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the U.S. ABWR design.

12. KEY WORDS/DESCRIPTORS (List wads or phrases met we essest researchers en exe6ns the reporf) 13 AvA(AsiUTY STATEMENT Advanced Boiling Water Reactor (ABWR) unlimited GE Nuclear Energy (GE) y SECURITY CLASSAFICAT)oN Final Safety Evaluation Report (FSER) crn,s Rege)

Design Certification (DC) unclassified F.inal Design Approval (FDA) 10 CFR Part 52 crn,s Report) unclassified

15. NUMBER OF PAGES 16 PRICE NRC FORM 335 p.89)

This form was electroncally produced by Ehte Federal Forms, Inc L__.___-.._

i i Printed on recycled Paper Federal Recycling Program

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