ML20210R213

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Final Safety Evaluation Report Related to the Certification of the System 80+ Design.Docket No.52-002.(Asea Brown Boveri-Combustion Engineering)
ML20210R213
Person / Time
Site: 05200002
Issue date: 05/31/1997
From:
Office of Nuclear Reactor Regulation
To:
References
NUREG-1462, NUREG-1462-S01, NUREG-1462-S1, NUDOCS 9709020345
Download: ML20210R213 (32)
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, U.S. Nuclear Regulatory Commission '

Office of Nuclear Reactor Regulation May 1997

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AVAILABILITY NOTICE Availability of Roforence Materials Cited in NRC Publications Most docements cited in NRC publications wn! be available from one of the following sources:

1. Th's NRC Public Document Room,2120 L Street, NW., Lower Level, Washington, DC N555-0001

.I, 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 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to bo exhaustivo.

Referenced documents available for inspection and copying for a f ae from the PRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information noticos, inspection and investigation noticos; licensee event reports; vondor reports and corresponconco: Commission papere; and applicant and licensee docu-monts and correspondenco.

The following documonts in the NUREG serios are availablo for purchase from the Government Printing Offico: formal NRC staff and contractor reports NRC-sponsored conference pro-

+

coodings, international agrooment reports, grantoo reports, and NRC booklets and bro-chures, Also availablo are regu'atory guidos NRC regulations in the Code of Federal Regula-tions, and Nuclear Regulatory Commission Is'suances.

Documents available from the National Technical Information Sorvice include NtiREG-series reports and technical reports prepared by other Federal agenclos and reports prepared by the Atomic Energy Commisslori, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, journal articlos, and transactions. Federal Register noticos. Fodoral and State logislation, and congressional reports can usually be obtained from those librarlos.

Documents sveh as thosos, dissertations, forolgn reports and translations, and non-NRC con-forence procoodings are available for purchase from the organization sponsoring the publica-tion cited.

' Single copies of NP'. 4 e ports ara available froo, to the extent of supply, upon written  !

request to the Office W v #tration, Distribution and Mall Services Section, U.S. Nuclear

, Regulatory Commission,Nashins, ton DC 2055S-0001.

Copios of industry codes and standards used in a substantivo manner in the NRC regulatory process are maintained at the NRC Ubrary, Two Whito Flint North.11545 Rockville Pike, Rock-ville, f.10 2P52-2738, for use by the public. Codes and standards are usually copyrighted and mty 60 purchased frcm the originating organization or, if they are American National Standards,~ from the American National Standards Institute,1430 Broadway, New York, NY 10018-3308.

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NUREG-1462 Supp.1 Final Safety Evaluation Report Related to the Certification of the System 80 + Design Docket No.52-002 Manuscript Completed
May 1997 Date Published: May 1997 1

Dh4k,; r1 deactor Program Management Omu 4 *.uclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

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ABSTRACT This report supplements the final safety evaluation report (FSER) for the System 80+ standard design. The FSER was issued by the U.S. Nuclear Regulatory Commission (NRC) staff as NUREG 1462 in August 1994 to document the NRC

. staff's review of the System 80+ design. he System 80+ design was submitted by Asea Brown Boveri-Combustion Engineering (ABB-CE), in accordance with the procedures of Subpart B to Part $2 of Title 10 of the Code of Federal Renulations nis supplesnent documents the NRC staff's review of the changes to the System 80+ design documentation since the issuance of the FSER. ABE CE made these changes as a result of its review of the System 80+ design details.

The NRC staff concludes that the changes to the System 80+ design documentation are acceptable, and that ABB-CE's application for design certification meets the requirements of Subpart B to 10 CFR Part $2 that are applicable and technically relevant to the System 80+ design.

iii NUREG-1462 Supplement 1

_ __ _ _ _ _ _ .~ _ _ _ _ - _ _

s CONTENTS Page

- A B S TRA CT . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i l l

- 1 INTRODUCTION AND GENERAL DISCUSSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.1 I ntrod uct ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 1.5 Summary Of Principal Review Matters . . . -, . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 - -

1.6 Index of Applicable Regulations and Exemptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 1.9 Index of Tier 2' information . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, & SYSTEMS . . . . . . . . . . . . . . 3-1 3.2 Classification of SSCs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3- 1 3.6.2.1 Pipe Break Criteria for High Energy Piping Systems . . . . . . . . . . . . . . . . 3 1 3.6.3.5 Review of ABB-CE Bounding LBB Analyses . . . . . . . . . . . . . . . . . . . . . . 3 1 3.9.3.1 Loading Combinations, System Operating Transients, and Stress Limits . 3-1

- 3.12.5.4 Damping Val ues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 4 REA CTO R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 - 1 4.2 Fuel System Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1 5 REACTOR COOLANT SYSTEM AND CONNECTED SYETEMS . . . . . . . . . . . . . . . . . . . . . 5 1  :

5.2.1 Applicable Code Cases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 5.4.3 Shutdown Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 6 ENGINEERED S AFETY FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 6 1 6.5 Containment Spray System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1 .

6.7 Safety Depressurization System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4 9 AUXI LI ARY S YSTEM S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9- 1 9.3.4 Chemical and Volume Control System . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . 9 1 10 STEAM AND POWER CONVERSION SYSTEM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 10.3 Main Steam Supply System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1

- 16 TECHNICAL SPECIFICATIONS ~ . . . . . . . . . . . . . . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . 16 1 19 SEVERE ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19- 1 19.1 - Probabilistic Safety Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19-1 19.2 Severe Accident Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19-1 19.3 Shutdown Risk Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19-1 19.4- Consideration of Potential Design improvements Under Requirements of 10 C F R 5 0.3 4(f) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19-2 19.4.6 Cost Benefit Comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19-2 21 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS . . . . . . . . . 21-1

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22 CONCLUSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 - 1

- Appendix A- CONTINUATION OF CHRONOLOGY OF CORRESPONDENCE . . . . . . . . . . . A-1 Appendix D CONTRIBUTORS TO THIS FSER SUPPLEMENT . . . . . . . . . . . . . . . . . . . . . . . . . D-1

Appen<lix F ERRATA TO THE SYSTEM 80+ FSER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1 v NUREG-1462 Supplement 1 -

1 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction 1.5 Summary Of Principal Review Matters nas report supplements the final safety evaluation report ne NRC staff stated in the FSER that, subsequent to (FSER) for the System 80+ standard design. The FSER the completion of the staffs review of the SSAR and {

was issued by the U.S. Nuclear Regulatory Commission CDM for the System 80+ design, ABB-CE will submit (NRC) staff as NUREG 1462 in August 1994 to a DCD for the staffs review, ne DCD, which will te document the staffs review of the System 80+ design, incorporated I;y reference into the final design Asea Brown Boveri Combustion Engineering (ABB-CE) certification rule, has two tiers of information that were made changes to the System 80+ design documentation, derived from and include most of the information in the after issuance of the FSER, as a result of its review of CDM and the SSAR.

the System 80+ design details, his supplement documents the NRC staffs evaluation of these changes ABB-CE submitted the DCD for the staffs review on to the System 80 + design and it also provides errata to December 16, 1994. In general, ABB-CE followed the the FSER. NRC staff guidance in letters dated August 26,1993, and August 3,1994, regarding the format of the DCD.

Combustion Engineering, Inc. (ABB-CE, the applicant) The staff provided comments on the DCD in a letter submitted the System 80+ design documentation under dated January 27,1995. ABB-CE submitted a revision Subpart B of Part 52 of Title 10 of the Code of Federal to the DCD on February 22,1995, which addressed the halations The documentation and information staffs comments. Additional revisions to the DCD, pertaining to this supplement were submitted on Docket based on additional discussions with the staff, were No.52-002. The design documentation includea the submitted by ABB-CE on March 24 and March 27, standard safety analysis report (SSAR), certified design 1995. Rese revisions to the I'CD are noted by a bar in material (CDM), and design control document (DCD). the margin next to the change and a [2/95] footnote at the bottom of the page. The February 1995 revision Each of the following sections or appendices of this was the last revision the NRC staff approved before supplement is numbered and titled the same as the issuing the notice of proposed rulemaking for the System section or appendix of the FSER that is being updated. 80+ design in the Federal Recister on April 7,1995.

The discussions are supplementary to and not in lieu of Subsequently, ABB-CE proposed additional changes to the discussion in the FSER unless otherwise noted. the System 80+ design documentation as a result of its Accordingly, Appendix D is a list of the principal review of the System 80+ design details. These contributors to this supplement and Appendix F contains changes were proposed in letters dated June 11 and July errata to the FSER. No changes were made to FSER 17,1996, and finalized in letters dated June 27 and July Appendices A, B, C, and E by this supplement. 25, 1996, respectively. The staffs review of this information is included in the appropriate sections of this This supplement is issued by the Standardization Project FSER supplement.

Directorate in the Office of Nuclear Reactor Regulation.

The licensing project manager for the System 80+ ABB CE submitted revised DCD pages, which design is Jerry N. Wilson, PE. lie may be reached by incorporated the above design changes and corrected calling (301) 415-3145, ,r by writing to the Office of various editorial and typographical enors. for the staffs Nuclear Reactor Regulation, Mail Stop O-10-D-22, U.S. verification by letter dated December 13,1996. The Nuclear R.:gulatory Commission, Washington, DC. substantive changes to the DCD were identified by a 20555-0001. Copies of the System 80+ design margin bar adjacent to the change and a footer date of documentation and all amendmenta and revisions are [11/96]; editorial or typographical changes also have a cvailable for public inspection at the NRC's Pu5lic footer date of (11/96] but do not contain margin bars.

Document Roons 2120 L Street, NW. (Lower Level), ABB-CE submitted the final version of the DCD on Washington DC. Copies of the FSER (NUREG 1462) April 30th and provided corrections on May 7,1997.

and this supplement are also available at the NRC's His revision of the DCD includes conforming changes Public Document Room, to the DCD introduction, seismic site parameters, and inspections, tests, analyses, and acceptance criteria (ITAAC). Rese revisions are identified with a footer date of [l/97]. The final versian of the System 80+

DCD is approved by this supplement to the FSER and is the version that will be incorporated by reference into the design certification rule for the System 80+ design.

1-1 NUREG-1462 Supplement 1

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' l.6 Index of Applicable Regulation ~

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and Exemptions
In the FSER, the NRC staff identified new as-tards for selected technical and severe accident issues for the Systeen 80+ design that were addressed and resolved during the design certification review, nees new

- design standards woes consequently included as additional applicable regulations in the proposed rule for ,

the purposes of 10 CFR 52.48,52.54,52.59, and $2.63.

b Conunission decided not to codify the additional . .

applicable regulations in the final rule, but the l Coaunission did set forth its intent with regard to these ,

new design standards in its SOC for the final design certification rule (See the SOC public conunent summary and resolution section on the need for additional applicable regulations).

1.9 Index of Tier 2* Information

- In the FSER, the NRC staff stated that any changes to i certain SSAR conunitments would require prior NRC approval before the change was implemented by a COL ,

applicant or licensee who referenced the System 80 +

design certification. % stafflisted these SSAR commitments in the FSER, and required that they be l identifialin the DCD as " Tier 2*' information. ABB.

CE identified the Tier 2* information in the appropriate -

acetions of the DCD.

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in various locatiot.s in the FSER, the NRC staff stated that any changes to Tier 2* information would involve an unreviewed scfety question (USQ) and, therefore, require NRC review and approval prior to implementation. His statement regarding USQs was used simply to indicate that the change process for Tier 2* information would be the same as that for proposed changes to other Tier 2 information that is determined by an applicant or licenses to be a USQ. However, a determination of whether or not a proposed change to the Tier 2* information would constitute a USQ has not been made by the NF.C. and the actual process for changing Tier 2* information is described in the final

- design certification rule.. brefore, the language in the FSER has been modified to conform with the language

. of the final rule and its SOC by the errata in Appendix F to this supplement. See the rule and the SOC section-by-section analysis regarding the process for changes and departures, and the SOC public comment summary and j

- resolution nection regarding the Tier 2 change process.  ;

I NUREG-1462 Supplement 1 1-2 I

3 DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, ce SYSTEMS 3.2 Classification of SSCs 3.9.3.1 leading Combinations, System Operating Transients, and Stress Limits ABB CE proposed revisions to DCD Table 3.21, which changed the seismic and safety claanifications of the ABB-CE proposed revisions to DCD Table 3.9 2 Safety Depressurization System (SDS) spargers, vacuum ' leading Combinations ASME Code Class I,2, and 3 breakers, and discharge piping so that all components in Components' to change the table title to ' Loading the discharge portion of the SDS are cleasified Combinations for ASME Class 1,2, and 3 Components consistently, in addition, the apargers and vacuum and Component Supports.' The NRC staff agrees that breakers were deleted from the list of safety class I,2, the loading combinations in Table 3.9 2 are applicable to and 3 valves in Table 3.2 2. As a result of thene component supports. Therefore, this change is revisions, the apargers, vacuum breakers, and piping in acceptable and does not affect the findings in the FSER.

the discharge portion of the SDS will be classified as non-nuclear safety (NNS), but Seismic Category 11. He 3.12.5.4 Damping Values Seismic Category 11 cleanification nasures that a failure or interaction of any of these NNS components will not ABB-CE proposed revisions to DCD Section 3.7.1.3, degrade the functioning of a Seismic Category i Figure 3.7 32. Table 3.71, and Appendix 3.9A. Rese structure, system, or component to an unacceptable revisions changed the maximum allowable damping safety level, which meets Position C.2 in RO l.29. He value for piping analyzed using the uniform envelope NNS classification of the discharge piping downstream response spectrum method from the ASME Code Case of the preuurizer safety valves is consistent with that of N-4t t 1 values to a 5 % value for all modes of vibration, current operating PWRs and is acceptable. The staff De revised Table 3.71 contains a footnote stating that conchidea that these changes are for consistency and do when the 5 % value is used for such piping, the not affect the findings in the FSER. conditions in RO 1.84 for using CC N-411 1 will apply even though Code Case N-411 1 is not being used.

3.6.2.1 Pipe Break Criteria for High Energy Piping analyzed using either the time history or Piping Systems independent support method will use the appropriate values in Table 3.7-1.

ABB-CE proposed revisions to DCD Table 3.6-3, 'High Energy Lines Within Containment.' These revisions in section 3.12.5.4 of the FSER, the NRC staff reported update the table to reflect the current design of the that as an alternative to the RO 1.61 damping values, pressurizer and Safety Depressuriration System. All which are in Table 3.71, variable damping values in four pressurizer safety valves are now mounted directly sacordance with the requirements and limitations of the on the pressurizer and the Rapid Depressurization Line ASME Code Case N 411 1 may be used, subject to the extends from the pressurizer to relief valves RC-408 and conditions given in RO 1.84 relative to the use of Code 409. Herefore, since the discharge portion of the Case N-411 1. In its evaluation of the above changes, S:fety Depressurization System is not classified as high the NRC staff considered the following inherent energy piping, the high energy lines in items 40,41,42, conservatisms implicit in the overall DCD criteria:

and 43 of Table 3.6-3 have been deleted, in addition, Itcins 58 and 59 have been revised to agree with the 1. Implementation of the conditions specified in RO 1.84 current design of the Rapid Depressurization Line, will generally result in a conservative design.

Rese changea result in criteria that are consistent with the guidelinen in SRP Section 3.6.2, and are acceptable. 2. He use of the uniform 5% value could result in a small underprediction of support loads and piping 3.6.3.5 Review of ABB-CE Bounding LBB deflection at higher frequencies. However, because the Analyses DCD (and other ALWR) seismic criteria are (1) based on ground response spectra as defined in RO 1.60 that ABB-CE proposed revisions to DCD Sections 3.6.3.7, are enhanced in the high frequency range (approximately 3.6.3.8, and Appendix 3.9A, which committed to 8-40 Hz), and (2) anchored at a relatively high peak combine the normal operating loads and the maximum ground acceleration value of 0.3g, the NRC staff finds design loads by the absolute summation method and that the use of the uniform 5% damping is acceptable change the factor on load for the leakage crack size from only for use on ALWRs.

V2 to 1. As discussed in criteria #5 of FSER Section 3.6.3.5.2, this is an acceptable alternative criterion for leak before-break (LBB) and does not change the findings in the FSER.

31 NUREG-1462 Supplement I

On the basis of the above evaluation, the staff has

concludal that use of the uniforam 5 % damping value when innpleumated with the seissnic and piping design criteria la the DCD will provide piping designs with snargins which are consistent with' thoes of designs using Code Case N411 1, as limited by RO 1.84, and is therefore acceptable.

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NUREG-1462 Supplement 1 3-2

i 4 REACTOR 4.2 Fuel System Design ABSG peuposed changes to section 4.2.2.4, ' Control Element Asseeably,' (CBA) of the DCD and Figwee 4.2.11,4.3 46, and 4.3 47. Does changes required confereeing changas to the Tier i design descripticas in Section 2.2.2 and Figwe 2.2.13. De objective of the changes is to allow. (1) for the possibility of having 4-element CEAs at twelve spacinc core Wh; and (2) for the possibility of replacing 4h CEAe with 12 element CEAe at specific core locatione. 'Ibse changes neo covmod and M by the W- criteria in Tables 4.1 and 4.2 of the FSER and are, therefore,

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ABS-CE also proposed as addition to Table 4.2 3 in the DCD. Die change did not affect the fladiage in the FSER and, therefore, the change le = = ,- " .

41 NUREG-1462 Supplement 1

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.2.1 Applicable Code Cases ABB-CE proposed a change to Table 1.8 7, 'ASMB Code Canee Applicable to System 80+' of the DCD, which added ASMB Code Case N 498, ' Alternate Rules for 10 Year Hydrostatic Pressure Testing for Class 1 and 2 Systems,Section XI, Division 1, Code Case N-498 was endorsed by the staff in RG 1.147, Revision 9, dated May 13,1991. *fnerefore, the staff finds the change of adding Code Case N-498 to be acceptable, implementation of ASME Code Case N 498 reduces the number of hydrostatic tests by five during the life of a plant that references this design. Therefore, changing the number of hydrostatic tests in Table 3.91 from 15 to 10 is acceptable. In addition, some conforming changes to page 5 4 of the FSER are provided in the errata in Appendix F to this supplement, 5.4.3 Shutdown Cooling System ABB-CE proposed a change to Table 5.4.7 2, Item 3.

The shutdown cooling pump discharge valve failure mode was changed from the " Fails Open" to the " Fails Closed" position, which accurately reflects the intended design purpose of the shutdown cooling system (SCS),

as indicated in the FSER. 'the NRC staff concludes that the change will not alter the intended design of the SCS, which is used to provide cooling capability to the reactor during plant shutdown and transients. Therefore, the staff finds the proposed change to page 5.4-40 of the DCD acceptable.

51 NUREG-1462 Supplement 1

6 : ENGINEERED SAFETY FEATURES I . -

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- 6.5 - Containment Spray System 6.7 Safety Depressurization System ABS CB proposed changes to Section 6.5.3.4 of the  : ABB-CE's preliminary design analysis for the safety DCD for Systees 90+ canc= ning the minissum available depressurization system (SDS) used 6 inch piping and not positive auction hand (NPSH) for the shutdown valves _that would allow the reactor coolant system cooling (SC) and ra=*===-* spray (CS) pumps. N (RCS) pressure to be sufficiently reduced from 2500 psia pg-r I changes are masted on page 5.5 23 and en to 250 psia following a total loss of feed water

, Figures 6.3.2.l A and IB. (TLOFW) event, in which the most limiting condition was assumed that steam generator feedwater was not Based on prospective pump vendors' data, ABB-CE recovered and feed and bleed for once-through cooling

- found that the NPSH available to the SC pumps was was not initiated. De SDS also provides other rapid insufficient when aligned for containment spray. RCS depressurisation flows in TLOFW events with Howeyw, there was adequate NPSH available to the SC safety injection available to prevent core uncovery while pumps during shutdown cooling operation. As a result, maintainmg a minimum required mixture level of 2 feet ABB-CE proposed to increase the sias of b crossover above the reactor core, pipe between the shutdown cooling system (SCS) line and the containment spray system (CSS) line, and the CS A recent detailed engineering study by ABB-CE pump suction line from 18 inch to 20-inch. The licensee concluded ht b SDS using 4 inch piping and valves has recalcuinted the available NPSH and the piping flow would provide enough RCS pressure relief capability to rates for the SC and CS pumps bened on the new pipe preserve the validity of the original TLOFW analysis, size. De maximum allowable containment spray flow he resized rapid deptemurization valves (4 inches) rate was reduced from 6500 gpm to 5500 gpm which would open no lens than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the pressurizer was the value used in the safety analysis for containment safety valves first lift and allow the RCS pressure to be spray. De minimum available NPSH to the SC pump reduced from 2500 psia to 250 pois prior to the reactor was calculated to be 19.6 feet at 5500 gpm which vessel melt-through for a severe accident.

saceeds the required NPSH of 18 feet for a typical CS or SC pump at runout flow. Derefore, ABB-CE The NRC staff concluden that the SDS using 4-inch concludes that the CSS would have adequate NPSH piping and valves would provide adequate RCS during all modes of operation, depressurization capability as required to mitigate TLOFW events. Testing requirements to validate the ne NRC staff reviewed ABB CE's submittal and SDS valve flow capacity and other related test concludes that the proposed design chanae is acceptable requirements as indicated in SDS inspection, tests, on the basis that: analyses, and acceptance criteria (ITAAC) will remain valid for the new piping and valve size. Therefore, the

. De change is required to conform with the SC NRC staff finds the proposed changes on pages 5.4-45 pump design. and 6.717, and Figure 5.1.2 3 of the DCD acceptable.

. The revised NPSH available to the SC pumps will exceed b required NPSH when the SC pump is aligned for containment spray,

. The reduced containment spray flow rate is the value used in the previous safety analysis.

61 NUREG-1462 Supplement 1

9 AUXILIARY SYSTEMS 9.3.4 Chemical and Volume Control Syilem The NRC staff concludes that a modification to the CVCS to allow both charging pumps on-line, ne currently approved System 80+ chemical and momenNrily, with a maximum allowed combined volume control system (CVCS) design includes an charging limit of 160 gpm, while maintaining the RCP interlock in the charging pump controls so that both seal injec; ion, is acceptable in place of the currently charging pumps cannot be operated at the same time gprovrJ interlock signal, whereby one pump must be during all modes of operation. De interlock was added completely shutdown before the standby pump is allowed na part of a protection feature that prevents an to be on-line, resulting in a momentary loss of seal inadvertent boron dilution during Mode 5 operation, in injection. De new de,ign feature does not alter the which the lowered reactor coolant volume leads to a results of the p. cat safety evaluation. Therefore, the smaller dilution time constant end results in the fastest NRC staff finds the proposed changes to DCD pages dilution rate and, therefore, yields the shortest time tt> a 9.3 29 and 15.411 and pages 2.7 56 and 59 of Tier 1 complete loss of shutdown margin, acceptable.

A recent study performed by ABB-CE identified this ABB-CE also proposed Circe additional changes to feature as a potential operational problem because the Section 9.3.4 of the DCD. The first change corrects an interlock requires that one charging pump must be inconsistency in Section 9.3.4.2.1. In Figure 9,3,4-1, completely shut down in order to switch to the standby Sheet 2, of the DCD, it is shown that the fluid leaving charging pump. In the process of shutting down the the purification ion exchanger is returned to the Reactor operating pump and switching to the standby pump, Coolant System by the charging pumps and not by the there will be periods in which reactor coolant pump shutdown cooling pumps, as it was erroneously stated on (RCP) seal injection cannot be maintained. To eliminate page 9.3-30 of the DCD.

this potential problem, ABB-CE proposes to delete the interlock signal and implement minor modifications to The second change corrects some errors in Section the CVCS, which would still validate the upper linait 9.3.4.3.1 of the DCD, which addresses the redundancy mwned in the boron dilution analysis. De boron of components in the chemical and volume control dihtion analysis for plant operations in Modes 2, 3, 4, system. The description of the redundancy for the seal

, and 5 indicates that, with a maximum charging flow rate irdection and purification filters on page 9.3 37 referred of 160 gpm, the dilution time to reach the minimum to ' pumps' instead of " filters.' This was obviously margin is between 2.5 and 3.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, as long as the incorrect.

RCPs are operating. In Mode 5, with no RCPs operating, the dilution time to reach the minimum The third change increases the normal operating pressure margin is between 1.2 and 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, for the Volume Control Tank (VCT) from 20 psig to 20-50 psig. This higher operatiag pressure is needed to To preserve the maximum charging flow rate of no more maintain sufficient hydrogen pressure in the VCr gas than 160 gpm used in the design-basis accident analysis space, to keep dissolved hydrogen in the VCT water at for inadvertent boron dilution, a design modification is between 15 and 50 cc H2 (STP)/kg of water. This value proposed to include a flow indicator controller and is specified in Table 9.3.4 l A of the DCD. De isolation valves in the charging pump piping discharge, increase in the normal operating pressure in the VCT to which would limit the maaimum combined charging flow 50 peig does not pose any safety concern because there rate to 160 gpm when both pumps are on-line. The is still a 50% margin left to the design preasure, system flow will be controlled and monitored in the control room, his system nulification will provide the The NRC staff finds these three additional changes to be flexibility for plant personnel to switch from the acceptable because they do not change the findings in the operating pump to the standby charging pump for FSER.

maintenance purposes, by bringing both charging pumps nunentarily on-line with the combined maaimum flow of no more than 160 gpm. RCP seal injection will also be maintained.

9-1 NUREG-1462 Supplement !

l ,

10' STEAM AND POWER CONVERSION SYSTEM ,

10.3 Main Steam Supply System ABB{E proposed changes to Section 10.3.2.3.2.1 and 10.3.4 of the DCD. The changes corrected the snaia steem isolation valve trypass valve closing time froin 10 seconds to 5 seconds or less. The NRC staff reviewed the changes and found theen acceptable because the 5 second closure tiene was used in ths afety analysis.

)

10 1 NUREG-1462 Supplement I

16 TECHNICAL SPECIFICATIONS ABB-CE proposed changes to Tecksical Specification 3.5.4 In containment Refueling Water Storage Tank (IRWST) Speci6cally Figure 3.5.4-1, which provides a curve of contal===8 :^- - / :. temperature vs. the IRWST water ^ _ _ , . :.. De current Agure has an IRWST temperature range from 407 to 1107. De proposed change is to revise the scale for the IRWST  !

temperature range from 607 to 1107. %ere is no cl.mge to the curve itself. De basis for the change is to achieve consistency between the Technical Specifications and the assumptions in the safety analysis. A minimum allowable IRWST temperature of 607 was assumed in the containment pressure analysis in Section 6.2.1.5.3.4,

' Active Heat sinks,' in the DCD.

%: other affected DCD sections are Appendix 16A, Section B 3.5.4, and Chapter 6. Table 6.2.122, An example of IRWST temperature of $37 corresponding to a containment temperature of 907 is given in the Technical Speufication bases on page B 3.5 25; it will be revised to conform with the change on Figure 3.5.41 of Chapter 16. The revised example shows an IRWST temperature of 817 and a corresponding containment temperature of 1107. For consistency, Table 6.2.122 of Chapter 6 will also be corrected to change the refueling water temperature from 807 to 817.

De NRC staff has reviewed the above changes and finds them accepteble W== the original curve has not been changed and the change were made to achieve consistency in the System 80+ documentation, i

16-1 NUREG-1462 Supplement I s

19 SEVERE ACCIDENTS 19.1 Probabilistic Safety Assessment provide an RCS hot leg water level indication when the reactor vessel head is detensioned and removed, as ABB-CE proposed changes to Section 19.7,

  • External compared with the currently approved system in which Eventa Analysis," of the DCD. ABB-CE deleted the operation of the IUTCs connected to the reactor vessel component and human error failure probabilities from head is limited only to those pericxis when the reactor Tables 19.7.5.1 1 and 19.7.5.31 but retained the high vessel head is installed.

confidence of low probability of failure (HCLPF) values.

W deletion of the quantitative portions of the design- nis new mid-loop IUTC system consists of an specific probabilistic safety assessmenta is consistent instrument installed in a tank connected to the RCS hot-with the NRC's guidance for preparation of a DCD, as leg piping near the shutdown cooling suction connection.

discussed in the statements of consideration for the final  % piping of the system tank is connected directly to design certification rules, b retention of the llCLPF the top and bottom of the RCS hot-leg and is isolated by vclues is necessary to meet commitment r2 in Table a series of isolation valves, with appropriate valve 19.151 of the DCD ABB-CE also deleted Tables position controls, indications, and displays in the cotrol 19.7.5.3 2 and Tables 19.7.5.4 1 through 19.7.5.4 7. room. Each RCS hot-leg will have a permanentl)

These Tablea dict not provide significant insights and installed and separate mid-loop IUTC system.

would not be needed by an applicant referencing the System 80+ design, h NRC staff has reviewed thene ne connecting pipe, up to and in.luding the second proposed changes to the DCD and found them to be system isolation valve from the RCS hot leg is designed acceptable. in acconlance with ASME Section Ill, Class I requirements, he system's tank and piping downstream 19.2 Severe Accident Performance of the econd isolation valve are designed for RCS operating pressure and temperature in accordance with ABB CE proposed a revision to Section 19.11.5.4.6.1 of ASME Section Vlli, including the system drain valve, the DCD in order to achieve consistency between the This RCS mid-loop water level system is available only deacription of the reactor coolant system (RCS) response during reduced inventory and mid loop conditions (Mode characteristics and Table 19.11.5.4.6 1 and Figure 5) and is usually isolated during normal operating 19.11.5.4.6.12 that are referenced in this section of the conditions (Modos 1 through 4). The mid-loop IUTC DCD. .e analysis of the RCS response is not affected instrument consists of a vertical array of the heated and by these changea and, therefore, the NRC staff finds the unheated junction thermocouples that nrovide alann change to page 19.11 145 of the DCD acetptable, setpoints for high water levels (water level approaching the steam generator nor21es) and low water levels (water 19.3 Shutdown Risk Evaluation level approaching loss of shutdown cooling suction),

The IUTC design will retain the same level of accuracy ABB-CE proposal changes to the system-level to within plus or minus 1 inch of the RCS hot leg water monitorin; of the reactor vessel coolant level. One of level indication and is displayed in the control room.

the reactor water level monitonng capabilities is Each IUTC system will have a separate power supply provided by the currently approved refueling heated and heater controller to prevent common-nxxle failure, junction thermocouples (IUTCs). This IUTC system providen narrow range indications, with an accuracy to The new permanently installed mid-loop level monitoring within plus-or-minus I inch, of the reactor vessel water system will not be affected by refueling activities in twel during mid loop operations via nessurement of the which the reactor vessel head is detensioned and reactor water level in the hot leg region. Operation of rer.nved. The new design should result in fewer water this refueling IUTC system is limited to rhone periods level reading errors and higher instrument reliability by when the reactor vessel head is installed. relocating instmments to a more benign area.

A study by ABB-CE concluded that a permanently b NRC staff concludes that the proposed mid loop installed mid-loop reactor water level measurement IUTC water-level monitoring system provides a better system using submerged IUTCs in a tank connected to altemative for measuring the reactor coolant level during the reactor coolent system (RCS) hot leg piping will reduced inventory and mid-loop conditions as compared continuously measure the RCS water level during to the currently approved HJTC system and, therefore, refueling operations when the RCS is in a reduced. approven the proposed changes to pages 5.11,7.719 inventory or mid loop condition, his system will also 19.8A-47,19.8A 156, and 19.8A-193 of the DCD.

19-1 NUREG-1462 Supplement 1 1

19,4 Consideration of Potential Design Improvements Under Requirements of 10 CFR 50.34(f) 19.4.6 Cost Benefit Comparison - Most of the candidate design alternativw wwe estimated to cost more than $20,000 and, therafore, were not cost-la the FSER, the NRC staff utilised a value of beneficial. The only design alternative that cost less

$1,000/ person.csv ($1,000/ person mm) svartud to - then $20,000 is the hookup for portable generators, ne estimate that a design improvement that cost more than estimated cost for this design alternative is $10,000 as

$17,000 would not be cost beneficial. Dis figure shown in Table 19.6 of the FSER. However, given that t-conservatively assumed that the total 60-year lifetime the hookup for portable generators was estimated to cost risk for the system 30+ design was elimiaawl by the on the order of $10,000, under either the 7% or 3%

design improvement (17 person cSv avwied risk a discount rate scenario, this design alternative would have

$1,000/ person c5v = $17,000) Since the FSER was to eliminate at least 50% or 25 %, respectively, of the ineued, the NRC issued ' Regulatory Analysis Guidelines totallifetime risk. Since the hookup for portable of the U.S. Nuclear Regulatory Commission

  • generators was estimated to only account for less than ,

(NUREO/BR 0058, Revision 2, Novernher 1995). His 1 % of the total risk, even for this most cort-beneficial j guidance document adopted a $2,000/ person-cSv design modification, the total costs continued to be well

($2000/ person rom) conversion factor, subject to present in excess of the total benefits. i worth considerations, and is limited in scope to health ]

. effects. Limiting the conversion factor solely to health In summary, the NRC staff concludes that with the effects required that the regulatory analysis include an significant margins in the results of the cost benefit additional dollar allowance for averted offsite property analysis, consideration of severe accident design damege, alternatives using the new values provided in ,

NUREG/BR-0058 do not change the findings in the I

The NRC staff reviewed the design alternatives previous analysis in the FSER.

identified in the SSAR using a $2,000/ person cSv averted for health effects and adoping a $3,000/ person.

cSv supplemertal allowance for offaite property (See NUREO/CR i349, ' Coat benefit Considerations in Regulatory Analysis"). Assuming a base case 7 % real discount rate as prescribed in NUREG/BR-0058 Revision 2, the present value of the health and safety benefits attributable to a cut beneficial design improvement would appromin.ste $20,000. A coinparable estimate for the health and safety benefits of a cost beneficial design modification based on a 3 % real '

discount rate, which is recommended for sensitivity analysis purposes, is $40,000.

NUREG-1462 Supplement i 19 2

1 21 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Comunittee on Reactor Safeguarde ec "- M the infonnetion discussed in this r:;;'-- ^

to the System 90+ FSER dunng their 433rd W on AugtW 8,1996, and subsequen;!y issued its letter on August 14,1996. The letter, which follows, reflects approval of the application for design certification and (aciudes no i+:-- "-f actione for either the NRC sistf or ABBCE.

21 1 NUREG-1462 Supplement 1

/ pa neog#g*

(MdITED sT Af ts

! NUCLEAR REGULATORY COMMISSION

{ ADV180RY COMMITTEt Oed RE ACTOR SAFEGUANDS waamsuctoes o c.aesea '

August 14, 1996 The Honorable $htrisy Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washingto1, D.C. 20555 0001

Dear Chairsan Jackson:

SU6JCCT: DE51GN CHANCES PROPOSED BY ASEA BROWN BOVERI - COM8U$Tl0N ENGIN(( RING R[tATING TO THE CfRTiflCAil0N Of THE SYSTEM 80+ DESIGN During the 433rd meeting of the Advisory Coanittee on Reactor Safeguards, August 8-10, 1996, we reviewed recent design changes proposed by A$tA Brown Bovert -

Combustion Engineering (ABB-CE) relating to the certification of the System 80+

design. These ' design changes

  • consist of both actual modifications to the destgre and corrections to the documentation to remove inconsistencies and typographical errors. We had the benefit of discussions with representatives of the NRC staff ar.d of ABS-CE. We also had the benefit of the documents referenced.

Conclusions Our review of Supplement 1 to NUREG 1462, ' final Safety Evaluatton Report Related to the Certification of the Systes 80+ Design,' did not change the conclusion reached in our earlier report of May ll,1994. We continue to believe that acceptable bases and requirements have been established in the application to assure that the System 80+ Standard Design can be used to engineer and construct plants that with reasonable assurance can be operated without undue risk to the health and safety of the public.

Backaround and Discussion We have been involved in the review of the System 80+ design since ABB-CE applied for certification. This review was carried out in accordance with 10 CFR Part i 52, which requires ACR$ to report on those portions of 10 CfR Part 52 applications that concern safety. In our May ll,1994 report to the Commission, we supported the certification of the Sys'ta 80+ design. This report was included in the staff Safety Evaluation Repers 'NRtG-1462). The present revicw is intended to supplement our earlier review f this ABB-CE application.

Sincerely, h j. W T. 5. Kress Chairman j l

l NUREG-1462 Supplement 1 21-2

. Nuclear Regulatory Commissten, NUREG-1462, Supplemet.t No.1, ' Final Safety Evaluatten Report Related to the Certification of the System 40+

Desta=,' dated July 1, 19M

t. ACRS E. pert dated May 11 1994, free T. $. Kress, Chatraan, ACRS, to Ivan selin, Chairman, Nat, subject: . Report en the Safety Aspects of the ASEA Breen Govert-Cos6vstion Engineering App 1tcation for Certification of the systes e6+ standard Plant Design
3. Letter dated June 27, igm, free C. B. Brinkman, Alt-Combustion Engineering Nuclear Systems to U.S. Nuclear Regulatory Commission, regardingSysten80+5tandardPlantDesignChanges
4. Letter dated July 17, 19H, from C. B. Ortakaan, ABS-Combustion Engineering Nuclear Systems, to U.S. Nuclear Regulatory Commission, regarding st additional design changes for Systen 80+ Standard Plant Design 3

S-l 21-3 NUREG-1462 Supplement 1

22 CONCLUSION The NRC staff performed its review of changes made to the System 110+ design documentation by ABB-CE

- in its letters dated June 27 July 25, and December 13, 1996 and other changes made to ceaform the System 80+ Dwign Control Docua4mt to the Commission's ,

guidance not the final design certification rules. The design chages wees rwiraint by the Advisory Commi: tee on Reactor Safeguards as described in Chapter 21 of this regnrt. On the basis of the evelustion described in NUREO 1462 and this report, the NRC staff concludes that the changes to the System 80+ design documaan=alon are acceptable, and ABB-CE's application for design certification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the System 80+

design f

22-1 NUREG-1462 Supplement 1

Appendix A CONT 7NUATION OF CHRONOLOGY OF CORRESPONDENCE This appendix contains an update of the chronological list of routmo licensing wi- - We in Appendia A of NUREO-1462. The w; , - '=+ S between the NRC etaff and ABB-CE regarding the review of the System 80+

design under Project 675 and Docket Number 52 002. Correspr=d==ce regarding the system 80+ design certification rulemaking to not included here but may he found in the rulemaking records.

July 26,1994 W. T. Russell, NRC, Latter transmitting FDA for System 80+ in accordance with Appendix 0 of 10 CFR 52 Fiche: 80366-001/80370-217 aca: 9407280072 July 26,1994 R. W. Borchardt, NRC, Letter responding to a letter requesting a detailed thermal-hydraulic summary analysis Fiche: 80423 310/80423 314 men: 9408020147 August 5,1994 D. M. Crutchfield, NRC, imiter updating guidance on preparation of uesign control document.

Fiche: 80498-192/80498195 aca. 9408100207 September 22,1994 R. W. Borchardt, NRC, Letter forwarding three documents used by NRC staff in evaluating severe accident phenomenon Fiche: 81064-001/81064 174 y acn: 9409280218 October 12,1994 C. B. Brinkman, ABB CE, Transmits input to System 80+ design control document Fiche: 81430 001/81432-207 aca. 9410210016 October 19,1994 R. W. Borchardt, NRC, Letter providing ABB-CE with necessary corrections to design control document introduction Fiche: none aca: 9410240089 October 28,1994 C. B. Brinkman, ABB-CE, Transmits language related to System 80+ design control document 2 Fiche: 81587 075/81587-076 acn. 9411010253 November 23,1994 W. T. Rueesti, NRC, Letter transmitting revised FDA System 80+ in accordance with Appendix 0 of 10 CFR 52 Fiche: 81866-088/81866-093 acn: 9411290113 December 8,1994 C. B. Brinkman, ABB-CE Forwards mart xl up CESSAR DC Figure 5.1.2 2 Fiche: 82110-286/82110-313 acn. 9412230043 December 14,1994 R. W. Borchardt, NRC, latter forwarding documents that form the basis for NRC staff severe accident safety evaluation Fiche: 82029-001/82029 317 aca: 9412140052 A.1 NUREG-1462 Supp!ement 1

l l

December 15,1994 C. B. Brinkman, ABB-CE, Transmits input to System 80+ DCD Fiche: 82164 001/82165140 aca. 9412280001 December 16,1994 C. B. Brinkman, ABB CE, Forwards System 80+ design control document Fiche: 82190 @ l/82216-048

.aca. 9412280070 December 16, 1994 C. B. Brinkman, ABB-CE Transmits justification for Tier 2 seismic & valve testing empiration Fiche: 82168-007/82168-009 aca. 9412280327 January 6,1995 C. B. Brinkman, ABB-CE. Transmits Rev. 2 to " Technical Support Document" Fiche: 82334-001/82334 099 aca. 9501120197 January 27,1995 R. W. Borchardt, NRC, latter forwarding comments on System 80+ DCD Fiche: 82591 065/82551 1787 aca: 9502010303 February 10,1995 C. B. Brinkman, ABB CE, Forwards proprietary parameters list Fiche: 82779 021/82779-026 acn. 9502160100 February 22,1995 C. B. Brinkman, ABB CE, Forwards revised design ces:rol document Fiche: 82907 001/82911239 acn. 9502280272 March 3,1995 C. B. Brinkman, ABB CE, Forwards additional information on renoval of auxiliary throttle coolor Fiche: 83045 290/83045-291 aca. 9503100148 March 14,1995 D. M. Crutchfield, NRC, Letter discussing ABB CE submitted views and positions Fiche: 83153 322/83153 323 acn. 9503170207 March 16,1995 R. W. Borchardt, NRC, Letter forwarding environmental appraisal of the severe accident design alternatives Fiche: 80252 302/80252 322 aca: 9503240046 March 17,1995 C. B. Brinkman, ABB-CE Forwards revision pages to the DCD Fiche: 83178 254/83178 314 aca. 9503200178 March 24,1995 C. B. Brinkman, ABB-CE, Forwards revision to the design control document Fiche: 83329-001/81329142 acn. 9503290050 March 24,1995 T. R. Quay, NRC. I.stter discussing status of a ABB-CE request to withhold proprietary information from public disclosure Fiche: 83681304/83681309 acn. 9504280181 NUREG-1462 Supplement 1 - A-2

March 27,1995 C. 3. Brinkman, ABB-CE, Forwards correcticJ pages to the design control document revisions Fiche: 83393 326/83393 329 acn. 9504040329 ,

January 24,1996 T. R, Quay, NRC, Latter discussing clarification of System 80+ PRA results Fiche: 86893@7/86893 009 aca. 9601290077 May 15,1996 B.K. Orimes, NRC, Latter discussing final DCD for System 80+ design Fiche: 88297-307/88297 308 aca. 9605170020 June 27,1996 C. B. Brinkman, ABB CB, Forwards finalized chattes 'o the System 80+

design control document Fiche: 88842 249/88842 351 aca. 9607010016 July 17,1996 C. B. Brinkman, ABB-CE, Forwards draft changes for staff review and approval Fiche: 89067 009/89067 025 mer. 9607180288 July 25,1996 C. B. Brinkman, ABB CE, Forwards fmalized change pac,kage of six draft changes to the System 80+ design control document Fiche: 89227 220/89227 236 acn. 9607310161 October 1,1996 S. L. Magruder, NRC, Letter forwarding RAI on small break LOCA Fiche: 89911001/8991141 acn. 9610030258 October 22,1996 S. L. Magruder, NRC, Letter forwarding status of a ABB-CE request to withhold company proprietary infornation Fiche: 90546191/90546-195 acn. %10250172 December 13,19% C. B. Brinkman, ABB CE. Forwards revisions to the System 80+ DCD Fiche: 91156-001/91156 324 acn. 9612180372 December 16,1996 F. J. Miraglia, NRC, Letter responding to a requested status of CESSAR DC relative to FDA for System 80+

Fiche: 91101 194/91101 195 acn. 9612160082 A-3 NUREG-1462 Supplement 1

2-r _ m .T i

-Appendix D CONTRIBUTORS TO TIUS FSER SUPPLEMENT l

HAMF, RESPONSIBILITY

/athony Attard Fuel System Design - l Bernard Bordenick legal Review l lierbert Brammer Mechanical Engineering j David Diec Reactor Systems  !

Jic Sien Ouo Plant Systems l John liuang Mechanical Enginaering  !

Shou Nien liou Mechanical Engineering larry Kopp Nuclear Physics Jaylee Radiological Analyses Chang Yang Li Plant Systems Janws Lyons Section Chief, Plant Systems Stewart Magruder Project Manager Janice Moore legal Review Son Ninh Project Manager Robert Palla Severe Accidents Krzysztof Parcrewski Chemical Engineering Janak Raval Plant Systems Nicholas Saltos Probabilistic Risk Assosament Dino Scaletti Senior Project Management, Generic lasues Michael Snodderly Severe Accidents Jerry N. Wilson Senior Policy Analyst and Project Manager f

D-1 NUREG-1462 Supplement 1

Appendix F ERRATA TO THE SYSTEM 80+ FSER Pane. Column. Paranranh Qangt Page 1 12,1:4 colunm, lat paragraph la the 3rd entry, change ' Applicable regulation for electric power system' to ' Applicable regulation for of tsite power source to safety division

  • Also, add the 4th entry with '8.3.1,1 Applicable regulation for alternate power source to non-safety equipment'.

Page 1 12,2rd column,4th paragraph la the 2nd entry, change 'ACI 349 (1985 Edition for desige and construction of internal structures' to 'ACl-349 (1985 Edition) for design and construction of seismic category I structures *, in the 3rd entry, change 'N690 (1984 Edition) for structural design and construction

  • to "N690 (1984 Edition) for d > sign and construction of steel structurea*.

Page 2-9,1st column,1st paragraph Replace the lat paragraph in Section 2.5.2.6 with "The COL

, applicant will compare site-specific earthquake free-field surface ground motiona, assuming a rock outcrop, to the ground motions used an input for the design certification. The COL applicant must verify that these site-specific design response spectra are enveloped by the control motion spectra shown in Figure 2.1 of the System 80+ FSER. ~1his action is identified as COL Action item 2.51."

Page 3 43,2nd column, let paragraph Replace the 1st paragraph with 'ABB CE has preser 'ed the site acceptance criteria in Section 2.5.2.5.3 of the DCD. For a rock site, mite-specific free-field ground surface response spectra at 5 percent of critical damping in the horizontal and vertical directions will be developed and compared to Figure 2.5 38 of the DCD, if the site-specific response spectra are enveloped by the spectra in Figure 2.5-38, then the site is acceptable, if the site-specific response spectra exceed either spectrum in Figure 2.5 38 at any frequency, a site-specific evaluation can be performed. In this evahtation, a site-specific structural dynamic analysis will be performed and the resulting in structure response spectra at six critical elevations

[ foundation basemat elevation (EI) 15.24 m (50 ft), interior structure El 27.97 m (91.75 ft), control room El 35.2 m (115.5 ft), top of steel containment vessel El 76.5 m (251 ft), interior structure El 44.5 m (146 ft), and shield building El 80.31 m (263.5 ft)) will be compared to the respective design response spectra in Figures 3.7D l through 3.7D 21 of the DCD. If the in-structure response spectra from the site-specific evaluation are enveloped by the in-structure design response spectra, for each of the six elevations, the site is acceptable. If the in-structure response spectra from the site-specific evaluation exceed the in-structure design response spectra, for any of the six elevations at any frequency, the design might still meet the design and licensing commitments due to the substantial design margin between the design commitments and the actual bases upon which the plant was designed. To demonstrate that the plant design meets the design and licensing commitments, a confirmatory site-specific evaluation can be performed to demonstrate that the System 80+ design meets the applicable design criteria for structures, systems, and components when subjected to the site-specific response spectra. The results of the confirmatory site-specific evaluation will be reviewed by the NRC staff."

i F.1 NUREG-1462 Supplement i  ;

... .. - . . - - . . . .-. -.- - ~ - . - . . _ ~ . - - - - - . . . ~ - -

Paga. Column. Paragraph Qgage Page 3 43,2nd column,2nd paragraph Replam tim 2nd paragraph with 'For a soil site, site-specific response l spectra at 5 percent of critical damping in the horisontal and_ vertical -

directions at the free-field ground surface will Se developed and compared to Pisures 2.5 39 and 2.5-40 of the DCD. If the site- -

specific ground surface response spectra are enveloped by the spectra

- in Figures 2.5 39 and 2.5 40, then the site is acceptable. If the site-specific response spectra escoed either spictrum at any frequency, a

' site-specific evaluation can be pert'ormed, la this evaluation, in-structure response spectra, at sia critical ekvations defined above,

'g- obtained from the site-specific evaluation will be compared to the respective design response spectra in Figures **.7D-1 through 3.7D-21 of the DCD. If tim in-structure response spectra from the eite-specific evaluation are enveloped by the in structure design response spectra, for each of the sin elevations, the site is acceptable. If the in-structuvo respones. spectra from the site-specific evaluation exceed

- the in structure design response spectra, for any of the sin elevations I

at any frequency, the design might still meet the design and licensing comunitments due to the substantial design margin between the design comunitments and the actual bases upon wiiich the plant was

- designed. To demonstrate that the plant design meets the design and licensing commitments, a confirmatot . specific eva luation can be performed to demonstrate that the System 80+ design meets the applicable design criteria for structures, systems, and components when subjected to the site-specific response 1.pectra. The results of i tin confirmatory site-specific evaluation will be reviewed by the NRC staff.'

f-Page 341, let column,1st paragraph Change "... would involve an unreviewed safety questica and, therefore, require NRC review and approval prior to  ;

t implementation " to *... will require NRC approval prior to 1 implementation." Also, delete the next sentence beginning with 'Any  !

requested change ...' ,

Page 342,2nd column, ist par). graph Change *... would involve an unreviewed safety question and, therefore, require NRC review and approval prior to implenantation.' to "... will require NRC approval prior to i implementation." Also, delete the next sentence beginning with 'Any roquested change ...'

' Page 345, let column, let paragraph Change _ "... would involve an unreviewed safety question and, ,

therefore, require NRC review and approval prior to implementation." to * .. will require NRC approval prist to ,

implementation." Also, delete the next sentence beginning with 'Any__

requested change ...'

' v' age 3 99,2nd coluna, tant paragraph Change * .. would involve an unreviewed safety question and, therefore, require NRC review and approval prior to

implementation." to *... will require NRC approval prior to l - i'-- -- ion." Also, delets the next sentence beginning with 'Any

^

i

( requested change '...'

1

! NUREG-1462 Supplement It F-2 u,-- - ,,,.n.. .* .,--- wr.., - . . . , ,r. .w+.-- ,- w -,,e r - < ram ,--

- . _ ~ . _ _ _ . . ~ _ . _ _ . . _ _ . _ _ _ _ _ _ _ . . _ _ . _ _ _ . . - _ _ . . . .

i ,

i l

Pap. Column. Passamph QItat

. Pay 3104,2nd column,5th paragraph Change "... would involve en unreviewed safety question and,_

therefore, require NRC review and approval prior to i=pe=====a=*ia=

  • to '... will require NRC approval prior to

'n' Also, delete the next sentence beginning with

'Purthernmore, any requested change ...'

i Pay 3136,2nd colussa, let paragraph Qange "... would involve an unreviemod safety question and,

, _ therefore, require NRC review and approval prior to

'::; ' - - ^'=- to '

'... will require NRC approval prior to

' ^

^H' Also, delete the most sentence beginning with 'Any

' _ _/ -

requesud cheap ...'

l Page 4 3,2nd column,2nd and 4th , sc . / - Oeny '... would involve an unreviewed safety question and require prior NRC review and approval prior to implesmentation,' to '... will

. require NRC approval prior to implementation," Also, delete the

- Arut sentence in paragraph 4 beginning with 'Any requested chany Page 5 4, tu column,2nd paragraph Change "... Ros 1.84 and 1.85, ...' to "... Ros 1.84,1.85, and

1.147, ..." and delete the nest sentence beginning with 'None of the

! specified ...".

Page 5 4, let column,4th paragraph Chany "... ROs 1.84 and 1.85, ...' to "... RGs 1.84,1.85, and i 1.I47,...".

Pep 6 4,2nd column,4th paragraph in line 5, change 'Rol.5" to 'RO 1.50'.

l Page 617,2nd column,4th paragraph la line 4, delete '6.3.4*.

! Pap 7 3, let column,2nd paragraph Chany *... would involve an wuoviewed safety question and, therefore, require NRC review and acceptance prior to I-7 ^:^k= ' to "... will require NRC approval prior to irg' ^ ^ ion.' Also, delete the nest sentence beginniit with "Any requested clunge ...'

Page 7 IL, let column, let paragraph Delete "... would involve an unroviewed safety question ..." and the i nemt sentence beginning with 'Any requested change ...'

! Page 7 9, let column, let paragraph Chany *... would involve an unreviewed safety question and.

therefore, will require NRC review and sc--J = prior to

in * "=
  • to *... will require NRC approval prior to 1

' r; ' - - h." Also, delete the next sentence beginning with 'Any requested change ...'

)-

Page 7-9. 2nd column,6th paragraph Qany *... would involve a unroviewed safety question and, therefore, requus NRC review and acceptance before

- -

  • to "... will require NRC approval prior to i,'- -_t=
  • Also, delete the next sentence beginning with 'Any requested cheap ...'

Pap 7 21, let column,5th paragraph Change '... would involve en unreviewed safety question and, .

4 therefore, require NRC review and acceptance before being -

i-;' '

  • to "... will require NRC approval prior to ir;'- ^_= ^* =." Also, delete the nest sentence beginning with 'Any requested change ...'

F-3 NUREG-1462 Supplement 1

Paga. Column. Parmaraph Quan Par 7 32, 2nd column, 4th paragraph Delete '... an unreviewed safety question would result from'...' the 2nd sentence and replace the period at the end of the 2nd sentence j with a comuna. Delete 'nerefore, any change to these issues ...'  !

from the beginning of the 3rd sentence and delete the 4th sentence beginning with 'Any requested changes ...'

Page 7 32, 2nd relumn, last paragraph Change "... would involve an unreviewed safety question and, therefore, will require NRC review and acceptance before being implemented.' to '... will require NRC approval priot to implementation.' Also, delete the next sentence beginning with 'Any requested changes ...'

Page 9 32,1st column,5th paragraph Delete the 3rd sentence and add 'In Amendment Q to the CESSAR-  !

DC, ABB-CE revised Section 9.2.6.2 to state that the CSS consists of a condensate storage tank (CST), piping and two recycle pu'mps.

De minimum capacity of the CSS is based on the maximum usage during startup (e.g., maximum steam generator blowdown vs. startup duration) plus 1009ercent margin. De CSS is constructed of stainless steel and has a stainless floating cover minimize air ingrees".

Page 9-32,2nd column, let paragraph la line 3, change 'CSTs* to ' CST *. In line 5, change "CSTs are' to

' CST is' and change 'are* to 'is".

Page 9 38,2nd column,4th paragraph la line 5, change "10 CFR 50.34(f)(s)(viii)* to '10 CFR 50.34(f)(2)

(viii).

Page 9 56,2nd column,3rd paragraph in line 4, change 'Section !!!.0, Ill.J and Appendix R' to ' Sections 111.0, Ill.J and Ill.O of Appendix R".

Sage 18 79,2nd column,3rd paragraph In line 14, change '18.2(6),18.4.2.l(14),18.4.2.8 and 18.4.2.1l' to

  • 18.7.1.8.1 and 18.7.1.8.2".

Page 18127,2nd column, last paragraph Change '... would involve an unreviewed safety question and, therefore, would require NRC review and approval prior to implementation.* to *... will require NRC approval prior to implementation.' Also, delete the next sentence beginning with " Any requested change ..."

Page 19-60, let column 5th paragraph In line 2, change *2000* to *200*.

Page 19-67,2nd column,2nd paragraph In line 9, change '662.9 m8 ' to '62.9 m8*.

Page 19 77,2nd column,3rd paragraph la line 15, change *19.11.4-l* to "19.11.4.4-l'.

Page 19108. Table 19.6, 4G column Change 8.3xtes for the Bird Diesel Gen ntor to 8.3x10.s and-3.3x10rs for the Filtered Containment Wu ta 3.3x10'8 Page 19-109. Table 19.6,4th column Change 3.6x108 for the Refractory Lined Crucible to 3.6(10+8 I

i Page 201,2nd column,1st paragraph la line 18, delete ' dated December 21, 1992*. In Line 19, change

  • Supplement 15" to " Supplement 15, dated April 1993*.

NUREG-1462 Supplement 1 p.4 l

l

Papa. Canam. M h l Pese A40,6th entry Change 'C.B. BrMam, CE, letter forwarding twined Combustion

"-f q Nuclear Fuel. FICHE: 80016 315' to 'C.B. Brinkman, CE, lesser forweeding copies of Asnandsment W to CESSAR DC. m FICHE:90lM @l'. $

4 F5 NUREG 1462 Supplement 1

latePonW ang UA NUCLaAR Rs00LATORY C0tesatte10N 1. REPORT fchdatR oem i4.ar=e nr ==c, Ade vd., s.oe, n .

            • "*"""*"*'*"'i 7,'E RIBUOGRAPHIC DATA SHEET ts=
Tmi AND SusTITLE Supgdoment 1 Faned Safefy Evalumbon Report related to the CertAcaton of the System 80+ Doeign a. Datt Report PUsusato ucutw l vtan Docket No. 52@2 y,y 3997 4.rv4 oR ORANT NUMr.f R S AUTHOR (8) 8 TYPE OF REPORT Safety Evaluation Report 7,PERIOOCOVERt0 peewepsana)

Sept.1994 May 19g7 A wy c=mesm **.e a'.mne . deme. Fe**=w.

e et p==RroRheNo a-. ww m.oRo.a,NitATioN

= =.; . NAME AND ADDREs3 fr wic, www twom owe a Aspen U a Nww Aere Dtvision of Reactor Program Management Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington. DC 20556-0001 e sponsoRm OnGANirATION . HAME AND ADOREss tr Mec, &c. *sp.w semi ase*we ..ww Mic them the e A.em U a Nwow Revisewy Cwmasm e w .w w medeseJ Same as above to. sueettutNT ARY Not t S Docket Nos $2-002 and 60-470, Project Number 676 51 ABSTRACT pre wees e hae)

This report supplements the final safety evaluation report (FSER) for the System 00+ standard design The FSER was issued by the U S. Nuclear Regulatory Commission (NRC) staff as NUREG 1462 in Augutt 1994 to document the NRC staffs technical rev6ew of the System 80+ design The application for the System 80+ design was submitted by Combustion Engineenng, Inc., now Asea Brown Doverl . Combustion Engineering (ABB-CE) pursuant to Subpart B of 10 CFR Part $2. This supplement documents the NRC staffs review of the changes to the System 80+ design documentation since the issuance of the FSER. ABB-CE made these chianges as a result of its revew of the System 80+ design details. The NRC staff concludes that the changes to the System 80+ design documentation are ecceptable, and that ABB-CE's application for design f,$rtification meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically relevant to the System 80+ design, 12 Krv WoRosocscRefoRs a.e.ed.e ,=. , t e...e, . w.e., y. ej u avaamin sTaimni

" h l' * *' 'd Asea Brown Dovert . OmbusHon Engineering, Inc. (ABB-CE)

Design Certification te sicusurvetassscation Evoluuonary Design au r.e.)

Final Safety Evalua3on f.eport (FSER) unclassified System 00+ Design p,,,, %

Subpart B to 10 CFR Part 52 unclassified 16 NiaM0tR OF PAGE; 1s PRICE Www au a em 33,,,,,,,,,,,,,,,,,g,,,,,,,,,g,,,,,,,,,,,,,,,,

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