Text

Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance.Appendices
	ML20202G181
Person / Time
Issue date:	12/31/1997
From:	; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	;
References
	NUREG-1560, NUREG-1560-PT06, NUREG-1560-PT6, NUREG-1560-V03-P6, NUREG-1560-V3-P6, NUDOCS 9802200077
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NUREG-1560 Vol. 3 Individual Plant Examination Program:

Perspectives on Reactor Safety and Plant Performance Part 6 Appen 'ces A, B, and C U.S. Nuclear Regulatory Commission Office of Nuclear R gulatory Research l

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1560 R PDR

AVAILABILITY NOTICE Availability of Reforonce Materials Cited in NRC Publications Most documents cited in NRC publications will bo available from one of the following sourcos:

] 1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Wathington, DC 20555-0001

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 be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Puolic k 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-monts and correspondence.

The following documents in tho 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, granteo reports, and NRC booklets and bro-chures. Also available are regulatory guides, NRC regulations in the Code of Federal Regula-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 Commissinn, 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 and State legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC con-ference proceedings are available for purchase from the organization sponsoring the publica- 1 tion cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear Regulatory Commission, Washington DC 20555-0001.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, Two White Flint North,11545 Rockville Pike, Rock-ville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted and may be purchased frorr. the originating organization or, if they are Amorican National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018-3308.

NUREG-1560 Vol. 3 3

Individual Plant Examination Program:

Perspectives on Reactor Safety and Plant Performance i Part 6 Appendices A, B, and C Manuscript Completed: October 1997 Date Published: Decanber 1997 Division of Systems Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

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PART 6 Appendices

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ABSTRACT

This report provides perspectives gained by reviewing probabilisticriskassessmentcharacteristics Methods,

. 75 Individual Plant Examination (IPE) submittals data, boundary conditions, and assumptions used in pertaining to 108 nuclear power plant units. IPEs are the IPEs are considered in understanding the probabilistic analyses that estimate the core damage - . differences and similarities obmed among the frequency (CDF) and containmem performance for various types of plants.

accidents initiated by internal events (including internal flooding, but excluding internal fire). This report is divided into six parts. Part I is a q summary repott of the key perspectives gained in

.The U.S. Nuclear Regulatory Commission (NRC), each of the areas identified above, with a discussion Office of Nuclear Regulatory Research, reviewed the of the NRC's overall conclusions and observations IPE ' submittals with the objective of gaining (Chapter 8). Pans 2 through 6 provide a more in-i perspectives in three major areas: (1) improvements depth discussion of the perspectives summarized in made to ind!vidual plants as a result of their IPEs and Part 1. Specifically,Part 2 discusses key perspectives

, the collective results of the IPE program, (2) plant- regarding the impact of the IPE Program on reactor specific design and operational features and modeling safety (summarized in Part 1 Chapter 2). Part 3

[ assumptions that significantly affect the estimates of discusses perspectives gained from the IPE results

.CDF and containment performance, and (3) strengths regarding CDF, containment performance, and human

_J and weakne:ses of the models and methods used in - actions (summarized in Part I, Chapters 3,4, and 5, ;

the IPEs. These perspectives are gained by assessing respectively). Part 4 discusses perspectives regarding

the core damage and containment performance results, the IPE models and methods (summarized in Part 1, including overall CDF, accident sequences, dominant Chapter 6). Part 5 discusses additional;IPE contributions to component failure and human error, perspectives (summarized in Part 1, Chapter 7). Part - i I

and containment failure modes. In particular, these 6 contains Appendices A, B and C whic provide the results are assessed in relation to the design and referencesof the information from the lPEs, updated operational characteristics of the various reactor and PRA results, and public comments on draft NUREG.

containment types, and by comparing the IPEs to 1560 (including staff responses), respectively.

1 i

4 i

A 1

lii NUREG-1560 1

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TABLE OF CONTENTS VOLUME 3 ChBMSI Ea&f ABBREVIATIONS .............................................................vil '

/.ppendix A: IPE References . . . . . . . . . . . . . . . . . . . . . , . . . . ...........................A1 Appendix B: IPE Updates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 1 Appendix C: Public Comments and NRC Responses on Draft NUREG 1560 ................,....C1 y NUREG-1560

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ABBREVIATIONS BWR Bailing Water Reactor CCFP Conditional Containment Failure Probability CDF Core Damage Frequency Cs Cesium ECCS Emergency Core Cooling System IIEP lluman Error Probability llRA Iluman Reliability Analysis I lodine IPE Individual Plant Examination LERF Large Early Release Frequency LOCA Loss of Coolant Accident MAAP Modular Accident Analysis Program NRC Nuclear Regulator Commission NSSS Nuclear Steam Supply System PECO Philadelphia Electric Company PRA Probabilistic Risk Assessment PSF Performance Shaping Factor PWR Pressurized Water Reactor Qllo Quantitative Health Objective RCP Reactor Coolant Pump SER Staff Evaluation Report SRV Safety Relief Valve SBO Station Blackout Te Tellurium VP Vice President WOG Westinghouse Owner's Group vii NUREG-1560

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1 APPENDIX /,

INDIVIDUAL PLANT EXMilNATION REFERENCES l

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App A. IPE References in this Appendix, the references ,*or the Individual additionalinformation are listed in Table A l. 'Ihis Plant Examination (IPE) submittats are provided. lPE information is provided via the submittal's date and submhtals and responses to NRC request (s) tot submittal's public document room accession number.

Table A l IPE References (of Information Used in NUREG.

1560).

Individual pia t esamination Licensee responses to request Plant nstne submittel for additionalinformation Submittal Act walon Submittal Accession Date Number Date Number Arkansas Nuclear One, ! 4/29/1993 9305040339 * .

l Arkansas Nuclear One,2 R/28/1992 9209010212 10/5/1995 9510110067 1

Beaver Valley 1 10!!/1992 9210150272 3/10/1995 9503130375 licaser Valley 2 3/17/1992 9203240301 9/1 t/1992 9210010222 ,

10/26/1992 9211030334 liig Rock Point 3/27/1994 9406080120 * .

tiraidwood l&2 6/30/1994 9408110123 * .

lirowns l'erry 2 9/01/1992 9209030199 9/21/1993 9309280175 4!!4/1995 9504180280 12/23/1993 9401060224 Brunswick 1&2 R/31/1992 9209100204 9/0w . ,94 9409200201 2/27/1995 9503010179 Ilyron I A2 4/28/1994 9405250189 * .

CaHsway 9/29/1992 9210090033 11/22/1995 9511280U8 Cah et Cliffs 1&2 12/30/1993 9401070022 9/12/1995 9509150108 Catawba 1&2 9/10/1992 9209240287 6/07/1993 9306150372 Clinton 9/23/1992 9210050174 11/22/1995 9511300286 Comanche Peak 1&2 10/30/1992 9211050102 * .

Cooper 3/31/1993 9304060035 2/20/1995 9502280017 Crystal Rivei 3 3/09/1993 9303150193 11/22/1995 9511280382 Dnis Besse 2/26/199" 9303030295 9/11/1995 9509150145 4

DC Cook l&2 $/01/1992 9205050329 2/24/1993 9303010355 2/26/1993 9303030121 12/03/1993 9312030217 4/25/1994 9405090139 A1 NUREG 1560

App A. IPE References Table A 1 IPE References (of Information Used in NUREG.

1560).

individual plant esamination 1Jeensee responses to request Plant name submittal for additionalinformation

! Submittal Accession Submittal Accession I

Date Number Date Number Diablo Canyon I&2 4/14/1992 9204240011 1/15/1993 9301250130 Dresden 2&3 1/28/1993 9304130182 10/28/1994 9411010060 Duane Arnold 11/30/1992 9212090167 6/26/1995 9$07100196 l'atley l&2 6/14/1993 91 % 240041 11/09/1994 9411180035 Iermi 2 9/01/1992 9209090121 6/30/1994 9407060029 IitiPatrkk 9/13/1991 9109190203 9/01/1992 9209140256 Iort Calhoun i 12/01/1993 9312070021 11/30/1995 9512040426 Ginna 3/19'1994 9403230240 ' -

Grand Gulf I 12/23/1992 **

9212290071 .

Iladdam Neck 6/29/1993 9307070lf3 * . ,

llatch l&2 11/11/l002 9212230136 10/07/1994 9410120348 Ilope Creek $/31/1994 9406 % 0125 11/06/1995 9511090179 Indian Point 2 8/12/1992 9208200238 10/31/1995 9511210368

!adian Point 3 6/30/1994 9407120222 6/20!!995 9506290190 Kewaunee 12/01/1992 9212090115 1/13/1995 9$01200288

!.aSalle 1&2 4/28/1994 9435090227 ** .

1.imerick I&2 7/30/1992 9208030288 " .

McGuire 1&2 11/04/1991 9111070233 6/30/1992 92070800$0 10/$/1992 9210210155 '

Maine Yankee 8/28/1992 9208030288 2/28/1995 9503080175 Millstone 1 3/31/1992 9204070238 $/25/1993 9306030323 Millstone 2 12/31/1993 9401100239 ' 31/1994 9406070213 9'27/1995 9509250347 Millstone 3 8/31/1990 9009100231 4/22/1991 9104290183 Monticello 2/27/1992 9203090231 2/15/1993 9302220084 NUREG 1560 A2

App A. IPE References Table A.I !PE Iteferences(of Infornistica Used in NUMEG.

1560).

Individe:l p. ant esamination Licensee responses to request Plant name submittel for additionalinformation Submittat Accession Submittal Accession Date Number Date Number t

Nine Mile Point I

m/1993 9308030002 6/26/199$ 9507030056 I Nine Mile Point 2 [/3 1 92080$0183 $/06/1993 930$1301ll North Anna la2 12/!4/1992 9212210199 4/27/1995 950$0200379 l

Oconce 1,2&3 11/30/1990_ 901206000$ 8/14/1992 9208240190 Oyster Creek 8/24/1992 9208280377 7/02/1993 93071$0084 Palisades 1/29/1993 9302120094 7/22/1994 9407280168 Palo Verde 1,2&3 4/28/1992 920$06002$ 2/25/1993 9303020319 Peach llotte n 2/[ 8'26/1992 9209010209 " .

Peity i 7/l5/1992 9207240153 11/24/1o93 9312060116 Pilgrim I 9/30/1992 921019010$ 12/28/1995 9601020192 Point Ileach l&2 6/30/1993 93( 7020355 9/26/1994 9406030077 Pral. , Island l&2 3/01/1994 9403090295 2/27/1996 9603040214 Quad Cities :&2 12/13/1993 9312210240 8/08/1994 9408120259 12/23/1994 9412290313 River llend 12/01/1993 9302120067 9/22/1995 9509260374 Robinson 2 1/31/1992 9209090l$2 9/27/1993 9309140049 Salem l&2 7/30/1993 9308060186 * .

San Onofre 2&3 4/29/1993 930$040246 t/20/1993 9501260308 Scabrook 3/01/1991 0103060219 7/23/1991 9107310374 Sequoyah 1&2 9/01/1992 9209030210 2/25/1994 9403080390 Sheaton llatris 1 8/20/1993 9309010155 = 1/25/1995 95012$0408 9/18/199$ 9509250023 South Texas I&2 8/28/1992 920911010$ 11/17/94 9411300102

s. St. Lucie l&2 12/09/1993 93121$0124 * .

Suminer 6/18/1993 9306290220 3/20/1996 9603250273 A.3 NUREG 1560 n ~ ,

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App A. IPE References Table A 1 IPE References (of information Used in NUREG.

1560).

Individual plant esaminatio i Licensee responses to request Plant name submittal for additionalinformation Submittel Accession Submittal Accession Date Number Date Number g Surry l&2 II/26/1991 9112060076 $/15/1992 9206010089 Susquehanna 1&2 12/13/1991 9112200133 6/27/1992 9202030122 Three Mile Island i 5/20/1993 9305280148 12/6/1995 9512110427 lurkey Point 3&4 6/25/1991 9106280106 3/11/1992 9203170219 Vermont Yankee 12/21/1993 9401060043 * .

Vogtle l&2 12/23/1992 9212280069 9/13/1995 9509190310 10/02/I995 9510060072 Waterford 3 8/28/1992 9209010231 * .

Watts liar I 9/01/1992 9210030222 12/27/93 9401070397 5/02/1994 9405090112 WNP 2 R'28/1992 9209080185 10/20/1995 9510230409 Wolf Creek 9/28/1992 92100$0289 8/30/1995 9509060171

7. ion l&2 4/24/1992 9204290315 2/22/1993 9302250285 9/01/1995 9509080045

' Informa* ion not provided on time for this report

" Information not requested NUREG.t 560 A4 n . . . .. .

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APPENDIX Il INDIVIDUAL PLANT EXAMINATION UPDATES 1

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App 11. IPE Updates I

1 The perspectives provided in this report aie based on Table 11.1 summarizes the updated pitnt specific die original probabilistic risk analyses (PRAs) information. Plant names are listed in the first 4 performed by the licensect for their Individual Plant column of the table; the CDF of the original IPE l Examinations (IPEs1 In many cases licensees submittal is listed in the second column for those

updated these analyr.es to teficct plant changes and, in plants that an ev.tated CDF was reported; the updated

, some cases, to incorporate st.ficoncerns, as noted in CDF is listed in the third column. Information the staff evaluation report (SER) of the licensee's regarding updated analyses os plant changes is IPE. For some of these PRAs, the results (e.g., core summarized in the fourth column: and corresponding ;

damage frequencies (CDFs) and dominant sequences) references are provided in the fifth column. !

changed. Furtlermore, seserallicensees provided as

part of their comments on Draft NUREO 1560 It is noted that if a licensee has reported an updated l

information regarding revised analyses and plant CDF more than once the most recently reported CDF changes. These changes are not reflected in the body is listed.

of this report; they are provided, however, in this Appendix. ;

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i B1 NUREO 1560

T C Table B-1 U lated plant-speci3e information' q l ;c :n

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G Plant name IPE Updated Comments Reference 5 '

C

$ CDF CDF 1

Arkansas Nuclear One,1 " " " - g ,

Arkansas Nuclear One,2 Beaver Valley I t

Beav r Valley 2 2.1E-4/yr 3.1E-5/yr No additior al information was provided Workshop Presentation by Westinghouse Owners Group (WOG) April 8,1997.

2.7E-4/yr 1.1 E-5/yr Update ir>cludd Commonwealth Edisc i, ~ Byron & Braidwood Braidwood I&2

- modeling and data changes Stations Individual Plant Examinations, Response to I

- complete revision of the human NRC Requests for Additional Information and reliability analysis (IIRA) Modified Byron and B aidwood IPEs," March 27

- cr-dit for several hardware and 1997.

n proceduralimprovement g,

Commonwealth Edison Company," Commonwealth New dominant sequences were reported Edison Company Comments Regarding Draft j

NUREG-1560,* February 14,1997. !

! " " " - l Big Rock Point Browns Ferry 2 i Brunswick I&2 2.7E-5/yr 9.2C-&yr It is stated that the CDF change is the Carolina Power & Light Company " Comments on .

Draft NUREG-1560, Individual Plant Examination !

result of modeling changes and plant improvesnents Prograrn Perspectives en Reactor Safety and Plant Performance (61 FR 65248),* March 14,1997. ,

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Table B-1 Updated p' ant-specific informatiee' ,

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m Plant name ;iE Updated Comments Reference CDF CDF Crystal River 3 " " The licensee provided an update of its IPE Flonda Power Corporation.

Individual Plant

.+.x to address the weaknesses noted Examination - Internal Events

July 11.1997.

in the SER s

Davis-Besse -

DC Cook 1&2 63E-5/yr 7.lE-5/yr No additional information was provided Workshop newoidion by WOG Apcal 8.1997. ,

1 i

Diablo Canyon I&2 8.8E-5/yr 4.5E-5/yr No additional information was provided Workshop Presentation by WOG, April 8.1997.

A sixth diesel generator was installed since Pacific Gas and Electric Company,

Response to the submittal Request for Comments on Draft NUREG-1560.* r e March 10.1997.

s' .

j Dresden 2 1.9E-5/yr 3.4E-6/yr Update included: Commonwealth Edison Company. "Dresden Dresden 3 1.9E-5/yr 5.0E-6/yr - modeling and data changes Individual Plant Examinanon (IPE). Response to

- complete revision of HRA NRC Staff Evaluation Report ami Modifmi Dresden

- several hardware and procedural IPE " June 28.1996.

improvements Comtnaawealth Edison Company. " Commonwealth

) Edison Company Comments Regarding Draft New dominant sequences were reported ,

NUREG-1560." February 14.1997.

Duane Amold 7.8E-6/yr 1.5E-5/yr New dominant sequences were reported IES Utiinies. Inc "Duane Amold Energy Center.

Response to Request for Additional Information on Individual Plant Examination

June 26.1995.

Farley I&2 13E-4/yr 9.2E-5 No additional infarmation ras provided Workshop Presentation by WOG April 8.1997. . 3

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@ Table B-1 Updated plaat4pecific information, p

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u. - m 8 Plant name IPE Updated Comments R~.we C CDF CDF 1E i

- t LaSalle i&2 4E-5/yr iE-5/yr The IPE has been updated and includes: Commonwealth Edison Company. Commonwealth

- modeling and data changes Edison Company Comments Regardmg Draft ;

- revision of the HRA NUREG-1560,* February 14.1997.

- 6a e improvement No additional information was prended Limerick I&2 " " All improvements listed as planned in the PECO Energy Company. " Comments Concerning IPE submittal have been implemented Draft NUREG_1560 Indmdual Plant Examinaten i Program: Pugves on Reactor Safety

, Performance ~ March 14.1997.

J c2 Maine Yankee -

L McGuire 1&2 " "

Direct current power improvements have Duke Power, " Duke 6er Company Comments on been implemented and the PRA is being Draft NUREG-1560, March 3. IM updated i

Millstone I -

e Millstone 2 -

i Millstone 3 5.6E-5/yr 5.tE-5/yr No additional information was provided Workshop P-uG by WOG. April 8,1907.

Monticello -

Nine Mile Pcint I -

i Nine Mile Point 2 - !

, i l North Anna 1&2 7.2E-5/yr 5.6E-5/yr Na additional information was provided Workshop Pwwd.Gon by WOG, April 8,1997. !

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i Table B-1 Updated plant 4pecific informatice~

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Plant masse IPE Updated Comasents Reference CDF CDF Oconce 1,2&3 " " The PRA has been updated Duke Power," Duke Power Company Comments on Draft NUREG-1560, March 3,1997.

Oyster Creek Palisades Palo Verde 1,2&3 -

Peach Bottom 2&3 r Perry I " " The contribution of anticipated transient Centerior Energy, " Perry Nuclear Plar.t Voluntary j w:thout scram (ATWS) to the overall CDF Co .tment on Draft 1560, Indivxlual Plant as has been reduced as a result of procedural Examination Pivpoui. Pugyes on Reactor Safety

" and Plant Performance, Summary Report. February modifications ,

25,1997.

i De . .a impovements inhibit automatic depressuruatim during an ATWS and l passive containment vent, under ;

consideration during the IPE, will not be implemented t i

I Pilgrim I 5.8F 5/yr 2.8E-5/yr .' 'iew dominant sequences were reported Boston Edison," Response to Request or Additional Information Regarding the Pilgrim Individual Plant r Examination (IPE) Submittal," December 28,1995.

Point Beach I&2 -

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Z Prairie Island I&2 5.0E-5/yr 1.7E-5/yr No additional information was provided Workshop Presentation by WOG, April 8,1997. __

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2 y Table B-1 Updated plant 4pecific information.

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9

~ Plant meme IPE Updated Comments Reference y CDF CDF C :

$ 1 1 No additional informaten was provided Workshop T. 44sve by WOG. April 8.1997.

Summer 2.0E-4/yr 9.6E-4'yr {--

7.2E-5/yr No additional information was pnmded Workshop Presentation by WOG, April 8.1997. 1 Suny I&2 1.25E-4/yr Susquehanna l&2 Three Mile Island I Turkey Point 3&4 Vermont Yankee No additional information was provided Workshop thw4 son by WOG. April S.1997.

Vogtic I&2 4.9E-5/yr 4.4E-5/p Waterford 3 I 4.4E-5/yr No additional informatbn was provided Workshop ik aina by WOG. Apii 8.1997.

C2 Watts Bar 1 33FAyr

& I.5E-5/yr New dominant sequences were reported Washington Tv .e- Supply System. *R4,A to Rw.4 WNP-2 2.0E-5/yr for Additional Informatica Related to Washmgton Public Power Supply System (WPPS) Nuclear Prticct No 2 (WNP-2)* October 20.199i 63E-5/yr No additional information was provided Workshcp thwenson by WOG. April 8.1997.

Wolf Creek 4.2E-5/yr f.0F-6/yr 4.8E-6/yr Update included- Commonwealth Edo - Comms Zion Individual Pla: t Zion l&2 Examination (TPE) Response to NRC Staff Evaluation

- modelmg and data changes

- compicte revision of IIRA Report and Modified Zica IPE.* September I,1995

- sever-? hardware and procedural Commonweahh Edison Company. "C .w.m. dth Edison improvements Company Comments Regarding Draft NUREG-1560."

Update resulted in new dominant xy w February 14.1997.

The most rec-ntly reported values are reficcted in this table.

- No new information was provided _ .a

APPENDIX C PUBLIC COMMENTS AND NRC RESPONSES ON DRAFT NUREG-1560

TABLE OF CONTENTS Chapter East C.1 Introduction . . ........................................................C1 C.2 - Chapters 2 and 9: Impact of the IPE Pro 6 ram on Reactor Safety . . . . . . . . . . . . . . . .......C4 C.3 - Chapters 3 and 11: IPE Results Perspectives: Core Camage Frequency . . . . . . . . . . . . . . . . . . . . C 5 C4 Chapters 4 and 12: IPE Results Perspectives: Containment Perfarrr.ance . . . . . . . . . . . . . . . . . C ll C.5 Chapters 5 and 13: IPE Results Perspectives: lluman Performance . . . . . . . . . . . . . . . . . . . . . . C 13 C.6 Chapters 6 and 14: IPE Models and Methods Perspectives . . . . . . . . . . . . . . . . . . . . . . . . . . . C-16 C.7 Section 7.1 and Cl. apter 15: Safety Goal implications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C 17 C.8 Section 7.2 and Chapter 16: Impact of Station Blackout Rule on Core Damage Frequencies . . . . C 19 C.9 Section 7.3 and Chapter 17: Comparl on with NUREO.ll50 Perspectives . . . . . . . . . . . . . . . . C 20 C.10 Chapter 8: Overall Conclusions and Observations . . . . . . , . . . . . . . , . . . . . . . . . . . . . . . . , . C 20 C.Ii Chapt:r 10: Background for Obtaining IPE Perspectives . . . . . . . . . . . . . . . . . . . . . . . . , . . . C 22 LIST OF FIGURES

_ Elst EAtt C.l Comparison of NUREG 1150 and Westinghouse seal LOCA models - old o-ring elastomer. . . . . C 9 LIST OF TABLES Inbic . Eggs C.l ~ Submitted comments on draft NUREO 1560. , . . , . . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . C1 C.2 - Relationship of draft NUREO.1560 to the final NUREO 1560 . . . . . . . . . . . . . . . . . . . . . . . . . C 3 ill NUREO 1560 4 -

ipp C. Comments and Responses Col Introduction Commission (NRC) and their contractors, representative of the owner's grcups, vendors, utilisies and their contractors, coasultants, and l'ederal and NUREG 1560, Volumes I and 2 were initially issued State agencies. A report summarizing the workshop in October and November 1996, respectively as a was prepared and is available for inspection in the draft report for public comment with tFe comment period ending May 9,1997. At that time, notices NRC Public Document Room (Ref. C.3).

The report includes presentation material distributed were published in the Federal Register announcing at the meeting and summarizes the discussien periods the availability of the report and requesting comment during which questions were raised and responses (Hel C.1). Distribution w as made to over provided. In addition, three sets of written comments 500 people and organizations in the United States and atro,d. wcre submitted at tbc meeting. These comments are included in Ar -ndix C or the Workshop Summary R eport. Tb m, hors and organizations submitting To assist readers of the document, a 3 day public these comme as are also listed in Table C.I (Items workshop was held h April 1997 on the contents of #23 26).

draR NUREG.1560. A notice of this workshop was published in the Federal Register (Ref. C.2) In response to the request for comments, the NRC and notification of the workshop was sent to all staff received 23 letters. The authors and persons receiving the dra4 report. The workshop organizations submitting these letters are listed in took place in Austin, Texas and was attended by Tab!c C.I. All letters received are avaihble for representathes of the U. S. Nuclear Regulatory inspection in the NRC Public Document Room.

Table C.I Subrnitted comments on draft NUREG 1560.

Identifiestion Date Author (s)

Organization recelsed by NRC

..r I Commonwealth Edison Company Thomas J. Maiman 21497 Executive Vice President (VP) 2 Niagara hiohawk h1artin hitCormick, Jr. 21497 VP Nuclear Engineering 3 South Carolina Electric and Gas Gary J. Taylor 2-17 97 Company VP Nuclear Operations 4 --

Tony Spurgin 21897 5 Centerior Energy Lew W. hicycrs 22597 yp 6 Duke Power Company h1 S. Tuckman 3 397 Sr. VP Nuclar Generation 7 Nev York Power Authority James Knubec 3497 Chief Nuclear Of0cer 8 Emergy Operations, Inc. Jenand G. Dewcase 3797 VP Operation: , Support C1 NUREG 1560 0 - -

App C. Comments and Responses Table C.! Submitted comments on draft NUREG 1560.

Identification Author (s) Liste

Orgenlaation received by NRC 9 tilinois Power Company Paul J. Telthorst 3797 Director, licensing 10 Pacific Ons and Electric Company Gregory Rueger 3 10-97 Sr VP & General Manager i1 - C.A Kukielka, 31297 i Eric R. Jebsen 12 Carolina Power and Light Company William Orser 31497 (

Ex.VP Energy Supply i 13 PECO Nuclear 0.A. Ilunger 31497 Director, Licens!ng 14 TU Electric C.L. Terry 31497 Group VP l$ 11WR Owner's Group - 31497 16 Westinghouse Owner's Group Louis F. Liberatori, Jr. 32597 Vice Chairman 17 Northeast Utilities Services Company Sunil Weerokkody 4 10-97 Supervisor, PRA l 18 OPU Nuclear, Inc. J.C. I'ornicola 4 29-97 Director, Licensing & Regulatory Affairs ,

19 Italtimore Ons and Electric Company Charles 11. Cruse 32797 VP Nucler.r Energy 5-8-97 20 Public Service Electric and Gas Company D R. Powell 59-97 Manager, Licensing & Regulation 21 IES Utilities,Inc. John F. Franz $ 9 97 VP Nuclear 22 Nuclear Energy Institute Anthony Pietrangelo 5-9 97 Director, Licensing Nuclear Generation 23 Environmental Protect t Agency T. Margulies

24 Virginia Power K. Tuley

25 New York State Department of Ilealth J. Dunkleberger

26- NRC IPE Workshop " '

' Written comments submitted at NRC.IPE workshop.

" Verbal comments discuss (d at NRC IPE workshop, NUREG 1560 C-2 l

In addition to these reviews and comments, as part of comments received on the draft, some of the chapters the normal review process, the staff discussed the were rearranged or renamed in the final report.

approach and results of draft NUREG 1560 with the Table C.2 shows the relationship of the drafi report to Advisory Committee on Reactor Safeguards on the final report on a chapter by chapter basis. The several occasions (Ref. C 4).

comments rectived were reviewed and categorized As discussed in Chapter i of this NUREO, the report acc rding to the various chapters. Comments related is comprised of two volumes, with Volume I as a to the

summary" chapter (from Volume 1) and the summary of the more detailed inintmation contained associated detailed chapter (s)(from Volume 2) are in Volume 2. Ilowever, due to the nature of the grouped together.

Table C.2 Relationship of dra'. 'iUREG 1560 to the final NUREG 1560 Volume I chapters Volume 2 corresponding detailed ciespters Einal report Drart report f inal report draft report

1. Introduction same no corresponding chapter no corresponding chapter

2. Impact of the IPE same ., 9 Plant Vulnerabilitbes same Program on Reactor and Plant Improvements l Sarcty l no correspondag chapter no contspondmg chapter ., 10. llackground for 10. llackground for Obtaining l Obtaining IPE Results Reactor and Containment Penpectius Dengn Perspectives 3 IPE Results 3. Core Damage ., 11. IPE Core Damage 11. Reactor Design Penpectives Perspecthen Core Trcquency Ircquency Penpectives Damage f requency Perspecthes

4. IPE Resulta 4. Containment 12. IPE Containment 12. Coatainment Design Perspectives Performance Performance Perspectives Containment Penpectives Perspectives I Performance

$. IP! Results 5. Iluman Action ., 13 IPE Iluman 13. Operational Perspectives Penpectives: lluman Penpectives Performance Performance Perspectives 6 IPE Models and 6. IPf4 with Respect to 14 Perspectives on PRA 14. Attributes of a Osality PRA Methods Perspectnes Risk.Infv,med Models and Methods Regulation Used in the IPEs 15. Comparison of IPts to a Quality PRA

7. AdditionalIPE same 15. Safety Goal 16 Safety Goalimplications Perspectives implications

16. Impact of Station 17. Impact of Station 111ackout Illackout Rule on Core Rule on Core Damage Damage Frequencies Frequencies

17. Comparison with 18. Comparison with NUREO-NUREO-!!$0 1150 Penpectives Penpectis es 8 Os erall Conclusions same e, no cornsponding chapter no corresponding chapter and Obsen ations C3 NUREG 1560

App C. Comments and Responses All of the written comments sent directly to the NRC C.2 Chapters 2 and 9: Impact of

(?tems 122 in Table C.1) and submitted at the yam on ReacW workshop (Items 23 25 in Table C.1)togetherwith all of the verbal comments provided at the workshop Safety (Item 26 in Table C.1) have been addressed in the final version of NUREO 1560. The somments fell in addition to comments identifying factual errors in into three broad categories: these chapters which were corrected, the following general comments were received. These comments and the NRC respuse are provided below (1) A number of comments either were editorialin nature or address the accuracyof the information provided in draft NUREG 1560. For these I, Comment: Numerous erroneous claims of comments, corrections were made to the text general applicability of vulnerabilities are made where appropriate. These comments are not in the report. Implying gener c applicability of reproduced in this appendix with staff response, vulnerabilities is inconsistent with the Individual plant Examination (IPE) purpose which is to The comments are available in the NRC Public Document Room.

identify plant.specinc vulrcrabilities and cost-effective improvements. (

Reference:

see Table (2) Some comments were observations in nature and did not appear to solicit a response not seek a revision to the text of the repon. These RE "'

comments are also not reproduced in this h is tme Gat de gene.k appucaMty of appendix with staff response. The comments are identined vulnerabilities cannot be ascertained, available in the NRC public Document Room.

in addition, there it no consistent dennition of vulnerability used in the IPEt Further, (3) Other comments address insights, interpretations variability in plant design and operation, as well and perspectives drawn in the dran NUREG. as different modeling assumptions, can make a 1560. In some cases, the commentors were vulnerability unique to a particular plant.

concerned that the conclusions were Therefore, statements regarding generic unsubstantiated. In other cases.commentors were applicability of vulnerabilities have been concerned about policy implications. For these rephrased in the NUREO. The purpose of comments, summaries were developed that presenting the vulnerabilities and associated plant captured the concem and an NRC staff response improsements identined by Se licensees is so to the comment is provided. These comments that all of the licensees may benent from and associated responses are provided in the considering these enhancements as means of following sections. The specine comments are improving the safety at their plant in a cost-available in the NRC Public Document Room. effective manner, Some of the comments discussed in the following 2. Comment: Claims that plant improvements sections are more general in nature and applied to identified by one licensee could be implemented insights, interpretations, etc. discussed in more than by other plants should not be made. Plant one chtpter of the report. Comments of this nature improveme nts sh ould not be implemented without can, therefore, appear in several sections of this a full assessment of induced competing risks and appendix. An attempt is made in each section to the expenditure of tesources required that may far identify those comments thst apply to other parts of outweigh any safety b:ncut gained. (

Reference:

the NUREG. see TaSle C.1, #15)

NUREG-1560 C-4

App C, Comments and Responses Response: I. Comment: The reported core damage frequencies (CDFs) and dominant contributors do All statements about generic application of plant not re0cct updated probabilistic risk assessment

' improvements have been rephrased in the (PRA) results. Many utilities '.aw nedated their NUREG. As with the identification of PRAs one or more times in responst to olant vulnerabilities, the purpose for discussing design and procedure changes. In addition, m.sg identified plant improvements is so that all licensers have provided the NRC with revised licensees can benefit by considering their IPE submittals some with extensive mode;mg potential implementa.tlon at their plant to improve changes and changes in the risk contributors and plant safety. A prudent evaluation by a licensee CDF, To correctly reflect insights from the IF2s of the benent of plant improvements identified requires consideration of supplementary 3 by other plants would involve both cost benefit submittals as well, (

Reference:

see Table C.1, and competing risk considerations. #1,12, l$,22)

3. Comment: Listmg improvementimplementation Response:

by the licensees as of the date of the IPE stbmittal is misleading because many plant Because many plent PRAs are being constantly changes have occurred since the initial IPE updated to re0cci the current plant design and submittals. (

Reference:

see Table C,1,61,16) operation, it is not practical to constantly update NUREG 1560 to incorporate new insights.

Response: NUREG 1560 is, and will remain, a compileion of the ca'culated CDFs and in?ights obtained NUREG 1560 represents 6 snapshot in time as far from the odginal IPE submittals, llowever, as risk and identified vulnerabilities and plart information from updated IPE submittals is linprovements (including their implementationL provided in Appendix B.

It is recognized that many licensees have updated their IPEs and the current status of identified 2. Comment: In comparing the plants, the plant improvements may be different than from categorization of boiling water reactors (BWRs) what was reported in the original submittal. solely by vintage, pressurited water reactors Updated plant improvement status reports are (PWRs) by nuclear steam supply system (NSSS) presented in Appendix B for those licensees who vendor, and Westlaghouse PWRs by the number provided updated status information in response of loops is not appropriate and can lead to to the solicitation of comments on Draft misinterpretation of results. It would be valuable NUREG 1560. to also look at the results based on a categorizationof architect /engineerar,ilor builder C.3 Chapters 3 and 11: IPE and also age of plant to see if variations can be b&P Ves: Core explained within each NSSS category, Further subgrouping of plants accordmg to similar dedgn Damage Frequency characteristics (e g., emergency core cooling system ECCS, designs) could be possible.

Many comments were received concerning the (

Reference:

see Table C.1, #16) accuracy of the information provided in these chapters or the insights that were identified. Corrections were Response:

made to the text where appropriate, in addition, several general comments were provided on the Early in the IPE Insights Program, the plants content of this chapter. These comments and an were grouped by architect / engineer and the IPE associated response are provided below, CDFs within and among these groups were c.5 NUREG 1560

App C. Comments and Responses compared, it was found that comparison of Response:

resul's on this basis was not productive because there is considerable design variability esen Whether plant specific design / operational differences or modeling assumptions are among plants designed by the same dominant factors in explaining the variability is architect / engineer, A decision was made to not always bvi us.11 wever,it is believedthat perform the ant. lysis using plant groups based either or both can play a significant role in the upon the NSSS vendor to account for basic NSSS variability for certain accident types. In many design differences. He BWRs were further sub- cases, a judgment is made in the NUREO on categorized by vintage to account for differences which is the dominant factor for an accident class in ECCS design. The Westinghouse plants were for a plant group. The NUREO identifies that a grouped according to the number of loops since significant amount of variability is due to support t the ECCS and other generalplant features for the system andothei plant specificdesign' operational plants in each of these groups are generally the differences. Many of these design / operational same (sec *able 10.3), it is recognized that the differences are highlighted in the report.

11 wever, it is also clear that modeling balance of plant including support systems for assumpti ns play an imponant part in the plants in each of the designated groups can be different and skew any comparison of the results documentation in the IPE submittals, it is not for a plant group. The NUREO identifies that clear if the modeling assumption really reflects a these plant specific features impact the results design or operational difference For ex mple.

aws the appropriate conclusions on the many licenrees did not credit an alternate coolant resultmg insights. Finally, it is recognized that injection system because they did not perform an further subcategorization of plants according to a analysis of whether or not it would be successful.

selected parameter could be made, llowever, The neglec' of the potential use of this system is variability in other parameters would likely a model assumption until it is shown that, because of plant specific factors, such a system impact that comparison. Because of this fact and could not be used. For other cases, it is clear also due to resource limitations, further that a model assumption is being made. For subcategoritation was not pursued.

exampic, many licensets assumed that the DC bus Icad shedding would always successfully

3. Comment: The degree to which a search for occur during a station blackout, variability associated with plant design differences has been made is questionable. The 4. Comment: The choice of success criteria has a NUREO states that important design features, major impact on the variability of the CDF operator actions, and model assumptions all results in a given category of plants. This is not impact the variability in results. Ilowever, few mentioned in the NUREO. Some utilities model assumptions are identified. As is well working with smaller PRA vendors had more known, substantial differences in PRA results stringent success (i.e., conservative) criteria than others who worked with reactor vendors and had occur because of balance-of plant and support accen t inf rma n 6at aHowed for kst system design differences despite similari*ies in

"E*'

NSSS design. %erefore,it is judged that there utilities had the resources to perform the is no basis to assert that the basis for observed necessary analyses to establish a less conservative variability is anything but dominated by plant success criteria where other utilities did not have differences in design, procedures, and training. .uch resources and chose to use a conservative

(

Reference:

see Table C,1, #15) success criteria. (

Reference:

see Table C.1, #16)

NUREG 1560 C6

App C. Comments and Responses Responset Response:

I ne NUREG identifies whrte success criteria Because of the variability in the IPE modeling, it assumptions impact the variability of the is not possible to always ascertain the impact of calculatedCDFs. As meationed in the response component failure rates and common cause to the previous comment, because of limited failure rates. Ilowever, these factors were documentation 5 the submittals, it was not considered in establishing the parameters alwayd cl:ar if differences in success criteria afTecting the variabiliev in the reported CDFs.

were due to dedn differences or inodeling Selected comparisons were made and, as assumptions. The basis for not crediting a discussed in Chapter 11, these failure rates were system (and in some cases, for crediting a found to be important to the CDF variability.

systeia) or for the operatmg requirements of a ' " """"' '"'I * "'

credited system (including support system Chapter 14 indicatesthat a wide variety of failure i

requirements)were not always documented in the ratn were identified in the IPEs for some components. nls variability applies not only to submittals. The CDF evaluation thus made n plant specific data but also to generic failure rates attempt to valida the difTerences in success identified in the subtrittals, criteria but simply reported its impact on the variability on the results. Also, Chapters 10 and 7, Comment: Care must Le taken when comparing 14 it the NUREG discusses the importance of 3 the CDFs from transient events and from loss of success crit e-in to the results in general terms-coolant accidents (LOCAs). De IPEs approach the modeling of consequential LOCAs (e.g.,

5. Comment: The NUREG should nitress the reactor coolant pump, RCP, seal LOCAs or criteria used to determine what constitutes core stuck oper power operated relief valves or safety damage. Many IPEs use core uncovery while relief valm, SRVs) differently. Sometimes the others use a peak cladding temperatue of CDFs from these events are reported in the 2200'F. This is important in that it impacts what transient contribution and sometimes in the small equipment can be used to avoid core damage, or medium LOCA CDF. It needs to be clearly

(

Reference:

see Table C.1, #11,15) stated how this is handled in NUREG-1560.

(

Reference:

see Table C.I. #16)

Responset

Response

ne impact of the definition of core damage on success criteria ls discussed in general terms in it is true that there was considerable variability Chapters 10 and 14. Specific impacts on the "* "E '#E" E'"I*E variability of the reported CDF definitions were "# 9"* "" #E "E " E "" #

LOCAs. Ilowever, the majority of the submittals not addressed because insufficient information was provided in the IPE submittals. reported sequences initiated by either a rupture or an inadvertent open SRV as LOCAs, and

}

6. sequences with consequential LOCAS occurring

) Comment: A discusCon on how the component after some other initiator as transients. This failure rates and the common cause failure rates format was chosen for categorizingand reporting impact the results is missing from the NUREG-the results. For those IPEs that did not provide Th:s could be particularly important for assessing the results according to this format, an attempt the importance cf station blackout (SBO) since was made to regroup the results to allow for the reliability of on site emergency AC power is comparison with the CDFs for other plants.

critical. (

Reference:

see Table C.1, #16) llowever, in some cases. insufficient information C7 NUREG 1560 s

._.__m

App C. Comments and Responses was proviaed in the IPE submittal to distinguish 10. romment: A basis for the key perspective that the CDFs associatedwith these different accident pWRs with better feed and bleed capability sequences. In those instances, the licensee's gerstally have lower CDFs should be provided.

reponed results for transients and LOCAs were There are many other plant design features and used directly. modeling methods that have a greater impact on CDF. (

Reference:

see Table C.1, #16)

8. Commentt it is not elear where specialinitiators 0t into the CDF irJormation reported in the Responset NUREG. Generally loss of component cooling water and loss of service water can be important The observation is made in the contex*. of all contributors to the CDF for PWRs due to the PWRs. Within the Westinghouse plant groups, potential for an RCP seal LOCA. It would be other factors besides feed-and bleed capabilities advantageous tc report the transient results in are more important for explaining differences in terms of CDF due to loss of decay heat removal transient CDFs and are discussed in the teport, a4d the CDF due to consequential LOCAs. Ilowever, differences in feed and bleed (Re ference: see Table C.l. #16) capabilities are important when comparing across all PWRs because of the Babcock & Wilcox and itesponse: Combustion Engineering plant design differences.

It is agreed that it would be useful to separate the 11. Comment: It is not clear from the inGrmation contributions from loss of decay heat removal presented that the Westinghouse RCP seal LOCA and consequential LOCAs for the transient model provides a lower contribution to CDF than sequences llowever, this information is not the IPEs that used the NUREG il50 model.

available vonsistently from the IPE submittals.

Estimates were maoe from the reported Since this is very imponant to many plants, it is information, whenever possible, and used in the recommended that NUREG-1560 provide a report to identify relevantinsights. t he NUREG detailed comparison of the two approaches. One identines that consequential LOCAs are of the dominating factors in the seal LOCA important contributors to the CDF for many model is the probability of core uncovery llWRs and PWRs. occurring within the Orst hour. IPEs using the Westinghouse RCP seal LOCA model typically

9. Comment: The discussion on LOCAs should be use 0.0283 and the NUREG 1150 model uses 0.0.

directed at the ability of plants to mitigate small The NUREG 1150 model does not ceasider any LOCAs. Overall, hrge LOCAs are not seal leakage for the first 90 minutes. From these signincant contributors to CDF. (

Reference:

see facts it appears that the Westinghouse RCP seal Table C.1, #16) LOCA modelis more conservative. (

Reference:

see Table C.1, #16)

Response

Response:

Signincant contributions were observed from different sizes of LOCAs in different submi:tals. A comparison of the seal LOCA probabihties Therefore,it is not always true that large LOCAs from the two models was not possible due to the are not signincant contributors and that the unavailability of the reports documenting the discussion should focus on only small LOCAs. Westinghouse model (with and without seal The NUREG discussion identifies what sizes of binding and popping open included). Ilowever, LOCAs dominate the LOCA centributions in the Point Beach IPE and the response to each plant group and the reasons why. questions concerning the Farley IPE did provide NUREG-1560 C8 I

App C. Comments and Responses an opportunity to compare the core uncovery while Surry and Farley are three loop plants and thus probability as a function of time for cases the core uncovery time for a given leak rate could be involving RCPs equipped with the old o.rinE different. Ilowever, since the reactor coolant system elastomer with the vessel either depressurized or volumes for the plants are roughly scaled by the not depressurited,and with the RCPs tripped. A number of coolant loops, the core uncovery times for cornparison of the values from these curve fits three plants for the same amount of leakage from with the core uncovery times calculated for identical cases for the Surry plant, as reported in each pump should not be substantially different.

NUREG.l l 50, Volume 3, is provided in Thus, the con uncovery probability comparison in Figure C.I, The curve fit is only valid over the Figures C.! provides a reasonable picture of the time frame of 30 minutes to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . It should differences between the NUREG.ll50 and be r, tad that point Ileach is a two loop plant Westinghouse seal LOCA models.

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00 i

'+ M4 EE1150 RM Z Hi

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t 13 23 23 14 18 41 O lo H 7A 1 13 13 H H H (1 46 u u 7J 1-igure C.1 Comparison of NURtu l150 and Westinghouse seal 1.OCA models - old o-ring elastomer.

Figure C.1 indicates that all three models predict smaller probabilities for core uncovery for time small probabilities of leaks and core r' Tery periods greater than approximately 3 '13.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , for early times (less than about 3 Lours). particularl, for cases where the vessel is Elecause of this, differences between the three depressurized. For scenarios where the vessel is not models do not have a significant impa:t on CDF depressurized, how ever, the probabilities predicted by for this early time period. Ilowev,r, for later the Westinghouse models rise charply at about 8 times, the differences are more significant. The hours, so that the three models give similar Westinghoun models generally predict much probabilities at that time.

C.9 NUREG 1560

App C. Camments cnd Responses The fact that seal LOCAs occur in all three 7.7E 7/yr). Accounting for this error would models does not mean that the impact on the slightly widen the difTerence between the CDP will be the same in both cases. As noted Westinghouse and NUREG ll$0 models.

earlier, none of the models result in a signincant cmtribution to CDF in the first three hours. 12, Comment: The NUREO discusses uncertainty llowever, unlike the Westinghouse models, the lated with the Byrondackson N 9000 seals NUREG il50 model can result in significant and " Infers rhar the / pes (for plants with these contributions to CDP based on core uncovery in

the 3 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months time frame. For example,in this pumps) are susput in their D seal m time frame during a station blackout, the core is conclusfora" Details concerning this technical likely being cooled by auxiliary feedwater, given issue have been provided to the NRC in various that battery power is still available. Therefore, forms in the past. Please modify the NUREO to without a seal LOCA, ccie damag .vould not be reDect the technical information provided and expected during this time frame. For times part remove the inference that the IPEs are sutpect k )

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , all three models predict a high their RCP seal LOCA conclusions. (

Reference:

probability of a seal LOCA leading to core see Table C.1, #8) uncovery. Ilowever, for these longer times, battery depletion would have occurred at most

E""'

Westinghouse plants, leading to loss of heat removal and boiloff. Therefore, if AC power recovery does not occur, core damage will result NUREG l$60 re0ccts the information provided whether or not a seal LOCA is present. In this in the IPE submittals which indicate that the situation, the station blackout CDF is not affected contribution from RCP seal LOCAs is generally by small seal 1.OCAs that would result in core small for plant with Dyrondackson pumps. 'Ihe uncovery at times greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The NUREG reiterates the statements made ..: the precise impact of the model differences is submittals that there is little or no potential for plant yecine, depending on battery depletion seal LOCAs in these plants if the RCPs are times and AC power recovery alternatives.

trippe<l. The submittals cite the design of the Similar impacts occur for non station blackout pumps some limited analyses, test, and actual scenarios (e g., loss of component cooling mter experience as the bas.is for tMs argument eventi) where the seal leakage rate impacts time available for other recovery actions such as providing some references. No judgement is arranging alternate charging pump cooling. made in NUREG 1560 ebout the accuracyof the RCP seal LOCA modeling for these plants based The documented NUREG il50 seal LOCA on the information in the subm'..al. llowever, model indicates no seal failure prior to 90 the potential for RCP seal LOCAs in these plants minutes. Ilowever, after inost of the IPEs were is still being reviewed as part of NRC's Generic completed, an error in the NUREG ll50 model Safety issue 23. The information cited in the was iden ined which indicates that there should submittals as the basis for the RCP seal LOCA be some probsbility of seal failure immediately modeling is being examined as part of the after loss of seal cooling. Thus, the contribution resolution of this issue.

of RCP seal LOCAA in the IPEs that utilized the NUREG-ll50 model is likely underestimated.

An evaluation for the NUREG il50 study for the 13. Comment: The reported CDFs have been Sequoyah plant indicates that the seal LOCA tounded to one significant figure. The NUREG contribution was underestimated by 18 % should report the actual CDFs reported in the (corresponds to a core damat,c frequency of IPEs. (haference: see Table C.1, #8)

NUREG-1560 C 10

App C. Comments and Responses Response: performance in tne same way that isolation or stractural failure of the containment 15.

A decision was made to report the CDFs to one Therefore, the NURFG separates the conditional signl0 cant Ogure (to provide consistency)and are probabilities of containment bypass and based on the actual values reponed in the IPE containment

failure" w hen making comparisons.

submittals. The importance of contain.nent failure frequency is acknowledged in CI' apter 12 of the NUREG C.4 Chapters 4 and 12: IPE where comparisons of containment failure Results Perspectives: '"9"'"*"*"''"9"'""

presented. The NUREG does not draw Containment Performance conclusions or make impli:ationi regarding

verall plant safety based on CCFPs.

1. Comment: Conditional containment failure Containment failure probabilitiei are used only to probability (CCFP) is not a good measure of compare the containment perf(rmance among safety performance. The use of conditional plants w;th the same type of containment and measures implies an independence between the among different containment ty) es. For this systems which prevent core damage and the purpose the CCFP is the best suited parameter.

systems which prevent containment failure w hlch is part of the design of the current generation of 2. Comment: The report utilizen at least Ove light water reactors. Plants with relatively higher different figures of merit in characterizing CCFPs are not necessarily less safe than those containment performance. It is never clear which with relatively lower CCFPs. The measures figure is most appropriate or why. The Ogures whicle impact public safety are related to the include: total conditional containment failure frequency of releases from the containment, probability, conditional probabi!ity of various

(

Reference:

see Table C.1, #11,12,22,23) containmem release types (bypass, early failure, late failure), frequency of bypass and early Response: release, conditional probability of "signincant early release," and frequency of releases with the One of the main objectives of the chapters in potential to cause carly fatalities. (Referei,ce:see NUREG 1560 related to containment Table C.1, #22) performance is to obtain perspectives on the performance of the various containment types Response:

independent from other plant features. For this purpose, the CCFP is a useful parameter since it NUREG-1560 uses various parameters related to decouples containment failure from core damage containment performance in different chapters of frequency. This was also recognized by the the report depending on the purpose of the majority of licensees since CCFPs are reported comparisons to be made and the perspectives to directly in most of the IPE submittals. Ideally, be obtained. There is no single "most the comparison of containment performance appropriate" containment performance figure of among difTerent IPEs would be accomplished by merit for the whole report, nor should there be, cor1 paring CCFPs for individual plant damage Those parameters which best served to illustrate states. Ilowever, such a comparison is rot the points to be made for the issues at hand were possible since the detinition of the plant damage chosen in different sections of the report Total states was left to the individual analyst and thus conditional containment failure probability is not varies from IPE to IPE. NUREG 1560 also used in the NUREG. For purposes of obtaining recognizes that the probability of containment perspectives o., containment performance, bypass is not a measure of containment conditions! probabilities of con,ainment bypass, C-11 NUREG-1560

App C. Comments and Responses the CCFPs for early and late failure are used in Regarding the industry position papers, their Chapters 4 and 12. Conditional probabilities of application in an IPE to qualitatively dismiss a significant or large early release, defined as early number of accident progression phenomena, releasewhere releases of Cs, I and Te exceeded without any sensitivity considerations.or without 0.1 of core inventory, are also compared in these any understanding of the uncertainty associated chapters since this type of release was singled out with the different phenomena,is not in line with in many IPE submittals. Finally, frequencies of the intent of Generic Letter 88 20. This ca ly release from bypass and early containment approach was less helpful in fostering a licensee's failure were used in Chapters 7 and 16 since this appreciation and understanding of severe accident prameter was the one which allowed an indirect behavior than a proper application of MAAP.

comparison of the IPE results with the safety goals. 4. Comment: Results are presented by reactor and containment type and NSSS. It would be 3, Comment: While there have been some mis- Saluable to also look at the architect /engineeror applications of MAAP, any implication that the bcilder to explain the variation in reported MAAP code is inadequate is wrong. It is results. (

Reference:

sec Table C.1, #16) misleading to state that MAAP does not have a comprehensive treatment of severe accident Response phenomena. A more problematic item involves the utilities which did not properly apply MA AP Early in the IPE Insights Program a decision was and'or relied on the industry position papers. made to group th containment performance

(

Reference:

e. Table C.1, #8,11,12,22) results under the five common containment classes ustd in the United States. Containment Response: response to severe accidents has been found to correlate to these five containment clanes as M A AP as well as other system level codes do not illustrated in the NRC's Containment cover the range of postulated severe accident Performance improvement progra n. In phenomena (e g., steam explosions, direct discussin g containm ent perform anc e perspective s, containment heating, shell melt thiough, NUREG 1560 identifies those architect / engineer hydrogen detonation). This is what is meant by specific containment construction features which the statement that the MAAP code does not have play a significant role in the IPE analysis, as a comprehensi e treatment of severe accident reported in the IPE submittals. These features phenomena. The EPRI report on MAAP include th: containment material, layout of acknowledges "one should recognize that MA AP reactor cavity, and location of sumps and drain cannot and does not contain detail,rd modcIsfor lines.

allphenomena " As noted above, other system level codes thare this limitation, and this is one 5, Comment: It is judged that there is no basis in reason why the IPE guidance called for proper NUREG 1560 to assert that the observed sensitivity studies to be conducted as part of the variability in the IPE results is anything but 1.evel 2 analysis. In some cases MAAP was dominated by plant differences in design, applied by the IPE analysts in a way that did not procedure, imd training. (

Reference:

see Table follow industry recommended guidelines. NRC C.1,#15) noted "..the adequacy of the MAAP 3.0B code for use In the IPEs.. " but also stated that Response:

" licensees .. bear Ihe burden of proof Ihat Ihey have applied she code properly, and that they In discussing containment performance meet the intent of the IPE generic letter." perspectives, NUREG 1560 identifies the plant NUREG 1560 c.12

App C. Comments and Responses specinc differences described in the IPE Hesponses submittals uhich lead to some of the variability in the reported results, llowever, it is clear that There exists detailed discussion in the appropriate modeling assumptions also play an important role sections of Chapters 4 and 12 of NUREG 1560 in the observed variability in containment performance. Assumptions regardingthe amount

how venting was grouped to the different and composition of core materialexiting from the containment failure modes.

scactor vessel, the coolability of this debn,s, and the pressure and temperature rise in the

how "carly" and

late" was deOned in the containment due to core debris dispersal are comparison of failure modes and releases.

examples of modeling assumptions which had a signl0 cant innuence on the assessment of

how multiple containment failure modes contahment performance. Other assumptions were treated as they were reponed in the IPE include the likelihood ofin vesselrecoveryof the submittal (i.e. whichever failure mode was accident, including the likelihood of retaining the considered dominant in the submittal base l core debris in the reactor vessel <la external case results was the one used in NUREG-l cooling of the vessel. 1560).

6. Comment: It would seem prudent to avoid The above comment on the treatment of misinterpretations by providing the specinc NRC dynamic failure modes is not clear, and no assumptions used in extrapolating IPE submitted further clari0 cation was provided at the

E; ' "'*9" Y' " '""E**

words to the construction of the comparisons made to NUREG 1560.

among plant results in NUREG 1560. These assumptions would include:

C.5 Chapters 5 and 13: IPE

What the relationship of containment vent llesults Perspectives: liunian treatment is to the CCFP, the early releases, Perforniance and other measures of risk;

l. Comment: It is stated in the report that in most

what the correlation is between each IPE cases there is little evidence thst the human result for early and late releases and their reliability analysis (llRA) quaatincation method dennition of *carly" and " late"; per se has a major impact on the results. This seems to imply that "the impact of HR4 on PX4

how the assignment of multiple containment can best Ac desnecdas indernminate"or "that failure nodes affects the assignment of the the HR4 sums to han hule det on the resuhs o ss e case, wh are tk allocation of failure modes in comparisons llRAs identined as important shortcomings of the (e g., shell melt-through following wetwell failure); and IPEs and why is the quality of the llRAs a concern. (

Reference:

see Table C.1, #8,11,12, 15, 21) e denning the treatment of dynamic failure modes and their associated failure locations Response:

as it relates to inferences about failure locations and timing. (

Reference:

see Table The interpretation that "the impact of HR4 on C.1,N15) PS4 can best be describedas indeterminate"or C 13 NUREG 1560

App C. Comments and Responses that "the llR4 seems to have little effe.t on the HR4s " (

Reference:

see Table C.1, Ml,2, 8,11, results of the PR4

is not what was meant. Ilow 12,22)

I and how well the llRA method is applied and the resulting human error probabilitics (llEPs) clearly itenponset I

have significant impacts on the results of the TRA. Thus, it is for this reason that concern is Confusion arose regarding the implication or raised in the NUREG about the " quality" of the meaning of the signincant variability in llEPs llRAs performed by the different licensees. ne that us identined for selected human actions statement that "in most cases there is litt/c across plants, particularly in terms of the quality evidence l at the HR4 quant (/1 cation methodper of the llRAs. Figures displaying the llEPs for se has a mq/or impact on the results," was meant several events (e.g., manual depressurization to imply that the llRA results from the different during transients) were presented in the report IPEs did not in general appear to vary directly as and discussions of the reasons for the variability a function of the particular or " nominal" llRA were provided. Many of the comments received method used, e g., the Technique for lluman from licensees on this topic attempted to defend Error Rate Prediction versus the Success the variability on the basis of the numerous Likelihood index Methodology versus the liuman reasonable factors that would lead to the variability. That is, the values across plants may Cognitive Reliability model. The variability in nave been developed on entirely different bases.

results appeared to be more a function of how or For examNe, Merent planu have Merent how well the llRA methods were applied or the system charhetedsdes and may W Merent impact of plant-specific characteristics, as

' ' " "" "'D'"#' E" #

oppmed to which nominal llRA method was factors and dependencies will also lead to used. Due the confusion caused by the statement variability in llEPs. Moreover, some plants only and the fact that the direct impact of the nominal g ,, ,, g ,

methcsl per se is dif0 cult to evaluate given the examined events. On the basis of these and other many other relevant factors, the statement was factors, the commentors indicated that such deleted from the Anal NUREG. Additional variability would be expected.

clari0 cation regarding the quality of the llRAs performed for the IPEs is provided below in the This conclusion is, at least in part, one point the response to Comment #2.

staff was trying to make and which was stated in the report. That is, there are "rcasonable

2. Comment: In spite of the assertion in the report explanationsfor much ofthe observed variability that "it appears that there are reasonable un llEPs across plants." in other words, the explanatiomfor much of the variability in llEP3 rather striking degree of variability, in at least and in the results of the llR4s er ss the nominally similar human actions, is t>ased to d#crcnt IPEs." it is also asserted that because some extent on valid differences. From this

" marry of the licemeesfailed to perform high- perspective it can be argued that the licensees quality llRAs. It is possible that the licensees attempted to consider relevant factors in obtained HEP values that are not appropriatefor obtaining the HEPs for operator actions and that their plants " These statements appear to be the results of the llRAs performed by the inconsistent. Moreover, others sections of the different licensees were generally consistent and report indicate that ext all of the variability in therefore useful. In fact, the staff does not in llEPs could be explained. Please provide general disagree with this conclusion.

clarification on what anpears to be inconsistent statements and address the assertion that "many of Ilowever, another conclusion reached by the staff the liceraces failed to perform highquality and documented in this report was that not all of NUREG 1560 C 14 i

App C. Comments and Responses the variability in the examined ilEPs was easily consistency exias, it is not necessarily tha case explained. That is, aner "au:eptable" reasons Sr that all the HEPs calculated by a particular plant variation were considered, there still eppearco to were realistic and vali<l for that plant. As noted be some degree of uner.plained variation the in Chapter: 5 and i l, reasonable consistency can HEPs (see Chapter 13). While sune of this be obtained in HRA without necessarily variation would be expected due to the lack of producing valid ilEPs. An HEP is oniy valid to precision in existing IIRA methods, it is also the extent that a correct and thorough application possib!e that some of the variation was due to of HRA principles has occurred. For example,if factors such as analyst biases, invalid ilRA a licensee shnply assumed (without adequate assumptions made by analyr performing the analysis) that their plant is " average"in terms of IIRAs. or superficial HRA an..ysss that failed to many of the relevant PSFs for a givr., mnt, but adequately examine and model the potential for then does appropriately consider the time human error (e.g., through careful consideration available for the event in a given conta be of plam specific performance shaping factors value obtained may be simil- .4 (PSFs), consideration of dependencies, use of for other plants with simil- m 4 .he simulator exercises, etc). Due to ,e limited event. Yet,the restilting- . ;stic information provided in many submittals on the or pessimistic relative ta a : hat would derivation of particular ilEPs, it is difficult to have been obtained if the licensee had conducted determine the extent to which inaporopriate a detailed vmiaation of the relevant riant-factors actually influenced the derived HEPs. specific - rs. Thus, while the degree of Ilowever, examinations of the submittals during m.sisterm obtahad by the licensees is the poject indicated that not all licensees encouraging regarding the ability to compare the performed quality HRAs. That is, not all results of the IPEs, and while many licensees licensees applied the existina HRA methods as performed excellent HRAs, the fact that somt well as the could have. For example, they did licensees did not perfonn as thorough HRAs as not always consider dependencies, appropriately possible given the state-of-the-an in HRA at the assess the impact of time availability.or carefully time, means that the results are not as good as consider plant-specific PSFs. Some iailed to they might have been it does not mean that model pre-initiator actio .s and others did not individual licensees and the industry in general conduct simulator exercises or peifonn did not obtain important information from walkdowns and timing of operator actions to be perfonni;.g the IPEs.

cot:dected outside the cc:itrol room, etc. The conclusion that not alllice:r ces conducted high- 3. Comment: By questioning the quality of the quality llRAs it fu ther documented in some of HRAs performed for the IPEs, NUREG-1560 the staff evaluation reports (SERs) that have been seems to imply that the licensees should have issued on the submittals. Some submittals attemp'ed to extend the state of-the-art in HRA in indicated u having met the intent of Generic order to obtain quality results. (

Reference:

see Letter 88 20 were found to have various Table C.1, #8,11, 21) weaknesses that could have influenced the HEPs obtained for particular eventa. Response:

y While the degree of consistency in HEPs The staff bclieves that tne state-of-the-artin HRA obtained for similar human actions in similar at the time of the IPEs was adequate for the contexts suggests that in general the HRA results inteia of Generic Letter 88-20. The shortcomings o

from the IPEs wue uscful in terms of meeting related to the HRAs performed for the IPEs were the intent of Generic Letter 88-20, it should N in how the existing methods were applied, rather furtlier noted that even who reasorwale han the methods themselves. Of course, this C-15 NUREG-1560 II

___.a N b

App C. Comments and Responses position does not impl/ that improvements are C.6 Chapters 6 and 14: IPE N needed in HRA, but rather that useful results Models and Methods can be obtained with thoughtful and thorough Perspect.ives -

applications of existing methcds.

Several comments were received expressing technical

4. Comment: he NRC needs to initiate a number disagrcement with sorre of the information ptoWded of policy and research activities to addres' in these chapters. The text was revised where shortcomings both in the NRC's attitudes and appropriate. In addition, several general comments ,

strategies for ensuring that the ticensees maintain were provided on the content of these chapters.

safe plants and in the development and use of These comment and associated responses are provided ,

PRA and HRA methoC and techniques. Rese below.

activities (summarized) include establishing a L

1. Comment: Numerous comiaents were received regulatory attitude that encourages the b.eensees on the description of " quality" PRA in to be pro-active rather than reactive (to the NRC) Chapters 6 and 14 anu _ .ne comparison of the in ensuring plant safety, encouraging more IPEs to a quality PRA in Chapters 6 and 15 of thorough and realistic HRAs, supporting the the draft NUREG Several commentors felt that development of multiple new rpproachesto HRA these chapters w e inappropriate for NUREG-(which inchide more effe:tive use of simulators), 1560 and that they should be deleted from the reevaluation of the real contribution of common t' mal report. This recommendation was largely driven by the assumption that the attributes of a cause to risk, reevaluating the use of Bayesian

"" *

I"I' " * *** **'

updating during " period of rapid changes in vequirements and that all the attributes had to be maintenance," and investigating the impact of met prior to using PRAs in future risk-infortned management and organizational factors on plant regulatory activities. Given that teie safety. (

Reference:

see Table C.1, #4) commentors felt that the PRA attributes were too demanding, overly prescriptive snd beyond the Response: current state-of-the-art, it follows that if they were assumed to be requirements then they could e n erPreted as a sigscant budn on &

The author (of the comments summarized above) mdustry. Several comments emphasized that the acknowledged that the " comments are notjust on scope and attributes of a PRA is be used for risk-the NUREG document itself but are also directed informed regulatory activitles should be towards some overall aspects of PRAs and commensurate with the application. This implies HRAs." However, none of ta comments appear that PRAs with significantly less attributes and of to address the NUREG itself. Nevertheless the more limited scope than the PRA described in NRC does cuirently have programs addressing NUREG-1560 would be acceptable for risk-exb of the i was raised by the author, e.g., inf ied applicatiens. Other commentors stressed t%t any anplications of the PRA development of i~,,, roved HRA methods and

. attributes in NUREG 1360 to the creation of an considerat. ion or the impact of management and indt.stry standard shouM ba viewed as organizational factors on plant safety. Further, developmental in nature. An industry-wide the NRC staff has reviewed the comments and standard for PRA quality should be based on a will consider them in future directions of broader and more deliberate development effort research. tl . involves practitioners from various NUREG-1560 c-16

- m . _ _ _

App C. CommGnts and Responses t,rganiutions. (neference: see Table c.I, #1. 2 C.7 Section 7.1 and Chapter 15:

8,9,15,16,20,22 and 26)

] Safety Goal Implications Responm Several comments were received expressing technical disagreement with some of the information provided Chapters 6,- 14 and 15 of draft NUREG 1560 in these chapters. The text was revised where I have been significantly rev:. sed for the final appropriate in addition, several general comments report. Specifically, Cnapters 14 and 15 have were provided on the content of these chapters.

been replaced with a new Chapter 14, and These comments and associated responses are i pmvided below.

referencesto the use of the IPEs in risk informed regul. tion have been removed, Chapters 6 and I. Comment: The concem is that the results

, 14 hi the final report summarize PRA reported in the original IPE submittals are not

characteristics and state that they

current and could be misleading when compared to the Safety Goals. For example, several plants

'are not " standards" nor do they represent identified in Chapters 7 and 16 in Draft NUREG-

resu'atory guidan: c. 1560 (Chapter 15 iri Final NUREG-1560) as potentially approachlag the early fatality

are included nnly as a benchmark in order to quantitative health objective (QHO) have draw perspectives on the models and suMequently updated their PRAs with significant methods used in the IPEs.

to t ns in CN and large ea@ release frequency, LERF (including- Browns Ferry, l Beaver Valley and Palo Verde). (

Reference:

see e do not define the needed quality or scope of Table C.1, #22, :5 and 26) the ira elements needed for a particular regulatory application. Pesponse:

2. Comment: Several comments vare related to NURECd 560 has been revised to clarify th:t the the following statement in draft NUREG 1560, P"5Pectiver on the safety goal are based on the j " and oueer utility personnel are excludedfrom c % Ins! IPEs/PRAs which may have subsequ 9 , ;

the peer review team." This statement was NUREG 1560 will not be revised. New interpreted by some comm: ntors as iruplying that information obtained by the stafTwill be included no employees of any utility can rerve as a peer in NUREG-1560 (see Appendix B). - In the cast reviewer. (

Reference:

see Table C.1, #1, 8,12, of the safety goal comparisons if any of the 15,16,20 a.nd 22) plants that we-e identified as ape esching the

' early fatality QHO subm'. revised results, this Response: will be noted in Chapter 7 ani 15 and the reader will be directed to the appendix.

i This int rpretation was not intended. The

2. Comment: Inferences tnat a few plants may

. statement was included simply to indicate that it approach the early fatality health objective based would be inap, spriate for utility staft to be part on a comparison of the IPE and NUREG 1150

! of the PRA paer revbv team for plants owned results'may noe be valid. Additional insights j- and operated by their utility. NUREG 1560 has gained from the cor,tainment perfoiw ance l been revised accordingly. evaluations a.nd recent research in the area may l NUREG 1560 C-17 4

App C. Comments and Responses lead to different conclusions than the NUREG- denne absolute risk levels, but rather to identify 1150 analyses. (

Reference:

see Table C.1, #16) plant sescre accident vulnerabilities.

Consequently, the safety goal computations Response: performed by the staff (described in Chapters 7 and 10 of draft NUREG 1560) are not an NUREG 15601 as been revised to clarify that adequate technical basis on which such a NURLG-ll50 containment results were not used conclusion can be drawn, in a related comment, to link the IPE results to the safety goals. For SECY 90-104 was ciuoted, " based on the early fatality risk, a two step process was used. sign {ficant additional r ' sources that would be in the first step, the frequencies of early required to make a mcaningful comparison of the l contuinment failure and bypass were obtained IPE results with the safety goalpolicy statement from the IPEs and plants with low frequencies and the potentialproblems astociated with using

(<l0 4/ry) wsrc screened out f- )m further the as-submitted IPE data, the staf recommends consideration. For the remaining plants, the that no direct comparisons be made unless the

frequencies of source terms with relatively large IPEs are reviewed to a greater level of detail release fractions (>0.03 Cs, I, Te) were obtained. than currently planned" As the commentor The source term frequencies were then adjusted believes that a review of g,reater detail did not for population and compared to the goal. occur, it was recccnmended that the direct comparison of IPEs to the Safety Goals in

3. Commert: There is an implication in the report Chapters 7 and 16 be removed from the Onal that the only way a comparison can be made to NUREG. (

Reference:

see Table C.1, #19,22)

~

the " Safety Goal" is to have a Level 3 PRA.

Such a PRA was never mandated, requested or Response:

suggested by the NRC and there are a number of ways to compare to the Safety Goal other than The final version of NUREG-1560 has been having a Level 3 PRA. The NURhG could revised to clearly describe the limitations of the address how the N (C and industry (there are approach used to compare the IPE results to the several EPRI documents t.nd other papers, safety goals and subsidiary objectives. However, positions and reports) have defined or linked the the use of Level 1 and 2 mdicator (CDF and NRC " Safety Goal" in terms of Level 1 and 2 LEr1F) as surrogates for the safety goals is surrogate indicators. (

Reference:

see Table C.1, consistent with recent industry positions (refer to

15, 21) Comment #3 in Section C.7 above) and consistent with the guidance provided by the Response: NRC for use of PRAs in risk-informed regulatory applications (Ref. C.5). The manner in The approach used by the staffir. Chapters 7 and which the IPE results are compared to the safety 15 of NUREG 1560 was based on using Level 1 goals is consistent with the " Integration Plan for and 2 surrogate udicators to link the IPE results Closure of Severe Accident issues," SECY to the safety goals. The wording in Section 6.4 147 art also consistent with the has, therefore, been changed to make it clear that recommendations of SECY-90-101, nan.ely, a Level 3 FRA is not the only way to make a " .. indirect comparison of the IPEs and other compton to the safety Goals. available PRAs with the Safety Goals, focusing on the insights gained and the adequacy of

4. Comment: One commert stated that conclusions regulation, is planned." Yhe SECY further based on using the IPE results for comparisons to recommends that the " stag evaluate the IPE the QHOs of the safety goals must be carefully results ar a whole and summarise any qualified. The purpose of the IPEs was not to conclusions and recommendations for the NUREG-1560 C 18 a~ ,- -

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

d App C. Comments and Respoases Commission at the completion of the JPE C.8 Section 7.2 and Chapter 16:

P W e88 "

Impact of Station Blackout

s. Comment: severai verbal and written questions Rule on Core Damage were received at the workshop related to the Frequencies appropriateness of the current tafety goals and the manner in which comparisons were made to Several comments were received expressing technical these goals. (

Reference:

see Table C.I, #23,26) disagreement with some of the information provided in these chapters. The text was revised where Response: appropriate, in addition, several general comments were provided on the content of these chapters.

The appropriateness of the current safety goals is These comnient and associatedresponses are provided a policy issue and outside the scope of NUREG- below.

1560. The use of Level I and 2 indicators as surrogates for the safety goals is consistent with I. Comment: Evaluation of the SBO rule would the staff's guidance provided in the recently benefit from a review of the results by Architect / Engineer and not just by reactor type.

published regulatory guides (Ref. C.x).

(

Reference:

see Table C.1, #16)

6. Comment: The definition of an early relca-Response: As is discussed in the response to particuhrly a large early release, and the tinv similar comments on Chapters 3 and 1I (and the as silable for effective evacuation after declaration report in general), early in the IPE Insights of a genercl emergency appears to be arbitrary. Program the plants were grouped by Consideration of the accident timing, tb site, and architect /engineerand the IPE CDFs within and the impacten esaeuation(such as an SBO) need among these groups were compared. No strong to be considered. (

Reference:

see Table C.I. correlation w' h the architect / engineer was found

25) because inere is considerable design variability even among plants designed by the same Response: architect / engineer. A decision was made to perform the analysis using plant grouas based A unique definition of a lcrge early release was upon the NSSS vendor to account for basic NSSS not provided in NUREG 1560. A large early design differences. The BWRs were further sub-release is defined in the staff's regulatory guides categorized by vintage to account for differences in ECCS design. The Westinghouse plants were (Ref. C.x) on the use of PRA in risk-informed egulation. Numerical objectives for the grouped according to the number of loops since the ECCS and other general plant features for the frequency of a large early release are also Pl ants in eacn of these groups are generally the provided in those documents. The frequencies of

, same (see Table 10.3). It is recognized that the early containment failure and bypass were used in balance of plant inclueng support systems for HUREG-1560 to screen out plants with low plants in each of the designated groups can be frequencies. The frequency of source terms with difrerent and skew any comparison of the re:ults relatively large rehase fractions were then for a plant group. The NUREG consistently examined in more detail to estimate the potential identifies that these plant-specific features impact early health effects. The assumption was made the results and draws the appropriate conclusions that these releases occur prior to efrective offsite on the resulting insights. Finally, it is cecognized i evacuation. This assumption could overestimate

hat further subcategorization of plants according the potential for early health effects. to a selected paran.etercould be made. ilowever, C-19 NUREG-1560

App C. Comments and Responses variability in other parameters would likely C.10 Chapter 8: Overall impact that comparisor.. Because of this fact and Conclusions and also due to resource limitations, further Obsen at. ions subcategorization was not pursued.

Some general comments concerning the content of C.9 Section 7.3 and Chapter 17: Chapter 8 were received from several organizations Comparison with NUREG- *"d '"d i*id"* "* *P "' ' 'h**** """*"**""

provided below.

1150 Perspectives

1. Comment; Due to the nature of the IPE process requested in Generic Letter 88-20 (a search for Several comments were tecelud expressing technical vulnerabilitica, not characterization of absolute disagreement with some of the information provided risk), the applicability of the IPE results for e in these chapters. The text was revised where regulatory follow up activity should be limited.

appropriatc. In addition, several general comments Section 8.2.4 states that the NRC staff plans were provided on the content of these chapters. follow-up activities to determine if additional These comment and associated responses are provided regulatory actions are warrantert for plants with below, relatively high CDFs or CCFP; NUREG 1560 does not con.ider revised CDF and CCFP values provid.d to the NRC, which in some cases, are

1. Comment: Chapter 18 in Draft NUREG-1560 substantially different than the original IPE (Chapter 17 in Final NUREG 1560; presents a submittal values. Consequently, use of the IPEs comparison of NUREG 1150 results with IPE for comparison to safety goals, identification of results as a whole. A more interesting " outlier" plants, and for direction of inspection somparison would be between the individual and follow up activities should be minimized.

Such actions have the potential to lead to NUREG-i l 50 results and the corresponding lPEs.

mefrective use of NRC staff ar.d utility resources This would provide a more detailed information in pursuing areas which are known to be on specific modeling issues. (

Reference:

see outdated. The NRC staff should evaluate these Table C.1, #2) changes in the plant CDF and CCFP values before planning follov.-up 'ivities. (

Reference:

see Table C.1, #20,22)

Response

Response:

Section 7.3 indicates that the focus of NUREG-1560 is on comparing global perspectives The IPE results and insights provide a useful discuaed in NUREG-1150 with the overall source of information for identifying areas w here results of the IPEs. A plant specific comparison follow-up activities might be warranted. The information containedin NUREG- 1560, however, between NUREG-ll50 and the applicable IPE is merely a starting point and is by no means the analyses are provided in the individual SERs on

** "E * #' * ' "I the five rPEs. Chapter 17 in the Final NUREG- .

plant-specific actions are taken, the best available 1560 has been revised to clarify the scope of the informatinr. will be considered, including any comparison in NUREG-1560 and to note that revisions to the original IPE submittals, plant-specific comparisons may be found in the recognizing that most of the newer information SERs. has not yet received staff review. Further, any NUREG 1560 C-20

App C. Comments and Responses proposed regulatory actions are subject to the the NRC tends to be more concerned with Backfit Rule as described in 10CFR50.109, eliminating vulnerabilities and reducing risks than with reducing burden. However, the latter

2. Comment: The NRC stafrs approach in looking objective is desirable and the NRC encourages at CDF and CCFP as independent factors is the industry to submit requests for reduced incorrect. It assume.= the existence of either a regulatory burdens in areas where they believe high CDF or higl CCFP is evidence on its own that risks are low and substantial cost savings can of a potential concern. In reality,the two factors be achieved.

should be looked at together. They are each a part of the overall mput to risk, which should be 4. Comment: The discussion of the Maintenance the f.gure of merit (the CDF/CCFP criteria do Rule says it is acceptable to use the IPEs to not have any established technical connection to determine risk significant systems. However, this the QHOs of the Safety Goal). (

Reference:

see is not compatible with the findings abot.t the Table C.1, #8,22) usefulness of the IPEs for risk-informed regulation. Likewise, the NRC implies that for Response: inspection purposes the IPEs a. adequate for tnem to target areas for plant-specific inspections The major objectives of the IPE Insights Program but NUREG 1560 states that the PRAs are only are outlined in both the Forward and introduction adequate to identify dominant accident sequence of NUREG 1560. For at least me of those types ard their relative importance. This seems objectives (i.e., providing perspectives on plant inconsistent. Furthermore, the NRC seems to be feature and assumptions that play a role in the attempting to use PRA information in a selective estimation of CDF, containment performance and manner, where it serves their purposes.

human performance), it is useful to look at CDF (

Reference:

see Table C.1, #22) and CCFP separately. The use of these parameters in NUREG-1560 does not imply that Response:

a high value for either parameter alone is a potential concern er will be the basis for References to the use of the IPEs in risk-regulatory decisions. . 4 stead, the use of these informed regulation have been removed fr.:n the parameters allows the staff to focus individually final version of NUREG 1560. Issues related to on the Level 1 and Level 2 analyses performed the quality and scope cf PRAs needed for risk-for the IPEs, thereby accomplishing the informed regulation are discussed in the staff objectives noted above. regulatory guides, and standard review plans.

The role of the IPEs in risk-informed regulation

3. Comment: Concerning any follow up regulatory will be deterrrined in the context of these activities,it's suggested that the investigation and documents, not NUREG 1560.

regulatory considerstions not be limited just to the high CDF or CCFP issues. Areas where the 5. Comment: The report implies that until risk impact is small and the safety benefit is not " quality" PRA requirements are fully met, PRAs appreciable should also be investigated for cannot be used for any regulate,) purposes. If reducedregulatcry burden. (

Reference:

see Table that is the case,"as is" PRAs are inappropriate to C.1, #6,16) support such areas as the Maintenance Rule and Technical Specification changes. Such an Response; interpretation is counterproductive and is not supportive of the PRA Policy that looks to The primary focus of the NRC is to assure the enhance use of PRA in regulation commensurate safety of the public. Therefore,it is natural that with the state-of-the-art technology. Recognized C-21 NUREG 1560 t

App C, Comments and Responses j weaknesses, and tools to deal with those submittals to the NRC, it is important for the weaknesses delineated in the Standard Review NRC staff to know if the crcdited improvements Plan makes the "as is" PRA applicabla for a wide have been made, variety of applications while " quality" PRA requirements are phased in. Waiting until perfect 7. Comment: The use of NUREG 1560 for a

" quality" of PRA is achieved before utilizing the variety of issues is discussed in Chapter 8.

results is impractical, it is expected that However, most of the disassions are actually

" quality" :nd " standardization " will evolve, r ot related to the use of the IPEs to address these through a priori defmition, but through frequent, issues. NUREG-1560 should not be the source repeated application and peer review of PRAs. of information for applications as discussed in

(

Reference:

sce Table C.1, #15,17) Chapter 8. The IPEs/PRAs are the primary source and should be used. (

Reference:

see Table Response: C,1, #8) l The comment is similar to comments received on Response:

Chapters 6 and 14 (refer to Commer,t #1 in Section C.6). These chapten and Chapter 8 have NUREG 1560 summarizes a great deal of been significantly revised for the final report it i.mportant safety information and provides a was not intended to imply that all the attributes starting point for identifying and addressing a in draft NUREG-1560 have to be met before a number of important safety issues. As such, it is PRA can be used to support risk informed an important document and staff resource.

regulatory applications. 11owever, tne staff recognizes that some of the information is out of date and that the individual

6. Comment: The NUREG states the NRC staff submittals contain more information. For any plans to conduct follow up activities to monitor particular issue, the staff will use the best imple entation of the potential plant available information, including any new improvements identified by the IPEs. The submitMis, recognizing that some of the new improvements were identiced as " potential information may require additional review, improvements" which in most cases were NUREG-1560 also provides comparisons among identified as areas for further rey! w The NRC the IPEs on selected issues, and this information seems to be taking them as having been is usefal to tha staff when evaluating the commitments. These improvements should not treatment of an issue by a panicular plant.

be treated as commitments unless the utility clearly identified them as commitments.

(

Reference:

see Table C.1, #8)

C.11 Chapter 10: Background for Response: Obtaining IPE Perspectives The NRC recognizes that the potential Several comments were received concerning the improvements are not commirments in a accuracy of the information provided in this chapter.

regulatory sense. However, in many cases the Corrections were made to the text where appropriate. 1 improvements were credited in the IPE. No general comments were made concerning this Therefore,if the licensee uses the IPE in future chapter.

NUREG-IS60 C-22 1

App C. Comments and Responns REFERENCES FOR APPENDlX C C. ! Federal Register, " Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Summary Report, Draft," Vol. 61, No. 221, Page 58429, November 14, 1996.

Federal Register, " Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Volume 2, Parts 2 5, Draft," Vol. 61, Ne 239, Page 65248, December iI,1996.

C.2 Federal Register, " Individual Plant Examination Program: Perspectives on Reactor Safety and Plant Performance, Volume 2, Parts 2 5, Draft," Vol. 61, No. 239, Page 65248, December 11,1996.

C.3 NRC Memorandum (From Mary Drouin to M. Wayne Hodges), " Draft NUREG-1560 Public Workshop Summary Report," October 3,1997.

C.4 ACRS meetings on IPE insights.

November 18,1993 January 26,1996 (subcommittee)

December 10,1993 February 8,1996 September 27,1994 May 23,1996 October 7,1994 June 11,1996 (subcommittee)

December 7,1995 June 12,1996 C.5 USNRC, "An Approach for Using Probabilistic Risk Assessment in Risk Informec Decisions on Plant-Specific Changes to the Current Licensing Basis," Draft Regulatory Guide DG-1061, June 1907.

USNRC, "An Approach for Plant Specific, Risk-Informed Decisionmaking: Inservice Testing," Draft Regulatory Guide DG 1062, June 1997.

USNRC, "An Approach for Plant-Specific, Risk informed Decisionmaking: Graded Quality Assurance,"

Draft Regulatory Guide DG 1064, June 1997.

USNRC, "An Approach for Plant-Specific, Risk Informed Decisionmaking:TechnicalSpecifications," Draft Regulatory Guide DG-1065, June 1997.

3 C-23 NUREG-1560

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

4 NRCFciJ1336 u.&. NUCLEAR RE ULATORY cOMMisblON 1. i.EPORT NUMBER b t t02, any[

m :2=

BIBLIOGRAPHIC DATA SHEET

, NUREG 1560 t ? llTLE AND SuBUTLE Volume 3, Part 6 Individual Plan' Examination Program:

P:rspectives on Reactor Safety and Plant Performance 3 DATE REPORT PuBUSHED 4 uosTR vtAR l

Appendices December 1997 4 FIN OR GRANT NUMBER 6 AUTHOR (S) 6. TYPE OF REPORT Finol 7, PERIOD COVERED pncheve Dates)

8. PERFORMING ORGANIZATION . NAME AND ADDRESS frNac, prowne Owman, onico or Region, U S Nuceer Reguinfory cc mirsson, ew meang ed*ess, a conw prowne name eM menne eness)

Division of Systems Technology Office of Nuclear Regulatory Research US Nuclear Regulatory Commission i

W:shington, DC 20555-0001 9 SPONSORING ORGANIZATION NAME AND ADDRESS pf NRL fype 'Same es above*, a contractor, prowde NRc Dwason, omco or Regeon. O S Nucker Reguiefory Comtreason, au manng essess)

I same as above

10. SUPPLEMENTARY NOTES 113.BSTRAcT (200 worts or mas)

This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment i'

performance for accidents initiated by intemal events (including intemal floods, but exciuding mtemal fire). The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, reviewed the IPE submittals with the objective of gaining perspectives in three major areaa: (1) improvernents made to individual pisnts as a result of their IPEs ano the collective results of the IPE program, (2) plant-specific design nnd operational features and modeling assumptions that significantly affect the estimates of CDF and containment ;>er.cmance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance r:sults, includ;ng overall CDF, accident sequences, dominant contributions to the design and operationa! characteristics of the v rious reactor and containment types, and by comparinc the IPEs to probabilistic risk assessment charactriristica Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the difference and similarities observed among the various types of plants.

12. KEY WORDS/DESCRIPTORS (l.sst words or phreses rief wie essst resecchm m beatmg rie report J 13 AVAILABUrY STArEMENT Probabilistic Risk Assessment unlimited Indi>/idual Plant Exemination 14 secuntrycWSlFicAT,0N Severe Accident (7w, enge; Generic Letter 08-20 uncianified l

~(Thns Repet) unclassified

15. NUMBER of PAGES to. price NRc FORM 336(240)

I E

Printed on recycled I Paper l

Federal Recycling Program

C NUKEG-1560 ' s h-Vol3,Part 6 REACTOR SAFETY AND FIANT PERFORMANCE FIRST CEASS MAIL UNITED STATES POSTAGE AND FEES PAfD NUCLEAR REGULATORY COMMISSION u WASH!NGTON, DC 20555-0001 pggg ,

OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 l

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