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Summary of ACRS Subcommittee on Severe (Class 9) Accident 861219 Meeting in Washington,Dc Re Proposed Generic Ltr for Individual Plant Exams as Part of Implementation Plan for Severe Accident Policy Statement.Agenda Encl
	ML20205D326
Person / Time
Issue date:	01/14/1987
From:	; Advisory Committee on Reactor Safeguards
To:	; Advisory Committee on Reactor Safeguards
References
	ACRS-2483, NUDOCS 8703300371
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DATE ISSUED:

1/14/87 g fp7 ACRS Subcommittee Meeting Summary / Minutes For Severe-(Class 9) Accidents December 19, 1986 Washington, DC Purpose The ACRS Subcommittee on Severe (Class 9) Accidents met on December 19, 1986 in Washington, DC. The purpose of this meeting was to discuss the Staff's (NRR) proposed generic letter for Individual Plant Examinations (IPEs) as part of the Implementation Plan for the Severe Accident Policy Statement in regard to the evaluation of existing nuclear power plants.

Included in the discussion were:

(1) Guidelines and Criteria for Five Reference Plants, and (2) IDCOR-Individual Plant Examination Methodology (IPEM). Copies of the agenda and selected slides from the presentations are attached. The meeting began at 8:30 a.m. and adjourned at 4:00 p.m., and was held entirely in open session. The principal attendees were as follows:

Attendees ACRS BNL (NRR Consultant)

W. Kerr, Chairman R. Bari M. Carbon, Member K. Perkins C. Mark, Member W. Lucas P. Shewmon, Member R. Fitzpatrick C. Wylie, Member I. Catton, Consultant IDCOR M. Corradini, Consultant l

P. Davis, Consultant J. Carter (ITC)

J. Lee, Consultant R. Henry (FAI)

D. Houston, Staff M. Kenton (FAI)

J.Gabor(FAI)

NRC/NRR K. Vavrek (W)

R. Brown (Delian)

T. Speis l

Z. Rosztoczy i

R. Landry l

F. Coffman l

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Severe Accidents Meeting Minutes December 19, 1986 Discussion In his opening comments, W. Kerr noted that the subject of the meeting was considered important and difficult. He indicated that perfection in the methodology would most likely not be achieved the first time around and that the Subcommittee had a considerable responsibility in the review of this matter.

T. Speis (NRR) introduced the topics for discussion with a brief over-view of the NRR Implementation Plan for the Severe Accident Policy Statement. In his discussion of accomplishments to date, he indicated that 16 of the 18 NRC/IDCOR technical issues of concern had been re-solved. He briefly reviewed the documentation status for the IPEs and the tentative schedule for the generic letter and the completion of individual plant analysis.

R.Landry(NRR)presentedthebackground,structureandcontentofthe proposed IPE generic letter. He emphasized that the letter would instruct utilities to consider only internal initiators. He discussed the scope of the examination and listed five acceptable methods for application in performing an IPE. These are as follows:

(1) 10COR-IPEM as approved, (2) Level II or III PRA with update, (3) Level I PRA with IPEM source term, (4) Simplified or Smart (Phase I) PRA as approved, or (5) Other systematic evaluation method as approved.

The NRR review of the acceptable methods is tentatively to be completed by January 8, 1987. He discussed the tentative schedule for completion of the IPEs and the regional / quarterly interfaces with the utilities while perfonning the IPEs.

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Y Severe Accidents Meeting Minutes December 19, 1986 Z. Rosztoczy (NRR) discussed, in detail, the scope of the IPE. He listed four expected accomplishments:

(1) identification and assessment of potential severe accidents, (2) plant improvements to prevent severe accidents, (3) improved containment performance, and (4) development and implementation of an accident management program.

He indicated that extremely unlikely events need not be considered, only those giving a core damage frequency (CDF) greater than 2x10-6 per year.

Also, sequences with a contribution to CDF of greater than 5%, irrespec-tive of the frequency limit, should be considered. He used the current analysis for the BWR Mark I reference plant in discussing how one would apply plant specific valuas in the performance of the IPE.

For accident management, he discussed three areas:

(1)approachandorganization, (2) training and procedures, and (3) instrumentation and equipment. He indicated that existing Emergency Operating Procedures would be reviewed and revised as necessary for accident management.

The severe accident guidelines and criteria for the five reference plants were presented by three BNL personnel:

R. Bari - outline and overview, R. Fitzpatrick - guidelines and criteria for RCS integrity, RCS heat removal, RPV depressurization, ATWS response, station blackout response and support system failures, and K. Perkins - guidelines and criteria for containment integrity and control of hydrogen burning. BNL has produced a series of reports for the five reference plants. The reports for BWR Mark Is and IIIs and PWR ice condenser containments had been provided to the Subcommittee. The report for BWR Mark IIs was now available and the one for PWR large dry containments would soon be available. The guidelines highlighted essential functions and the criteria were based on system availability, operating / emergency proce-dures or maintenance surveillances.

r Severe Accidents Meeting Minutes December 19, 1986 The IDCOR-IPEM presentation was given in three parts:

(1) BWR-IPEM by R. Brown (Delian), (2) PWR-IPEM by K. Vavrek (W) and (3) Source Term Methodology by R. Henry (FAI).

J. Carter (ITC)gavetheintroductionto the IDCOR activities. He indicated that the methodology is a screening methodology, not a PRA technique but based on PRA techniques.

IDCOR believes existing plants are safe enough and is searching only for those cases of unusually high core damage frequency or unusually poor contain-ment performance. The IDCOR presentation was planned to inform the Subcommittee of the revisions to the methodology that had been made since the previous presentation on this matter on September 24, 1986.

He stated that IDCOR has some serious disagreements with the NRC/NRR efforts and feels that NRR is mixing other agenda or programs with the guidance in the Severe Accident Policy Statement.

R. Brown (Delian) presented an overview of the BWR-IPEM, some results obtained on various BWR reference plar.ts and some responses to NRC/EPRI comments on the BWR-IPEM.

Four plants were identified for the verifica-tion phase:

PeachBottom(MarkI),Susquehanna(MarkII),Shoreham (MarkII)andGrandGulf(MarkIII). The resources estimated to perform an individual plant analysis ranged from 24 to 48 man-months.

In most cases, the core damage frequencies were higher by using the BWR-IPEM as compared to other PRA (ASEP) results. The dominant accident sequence differed in the two studies: station blackout and transient initiated sequences about equal for IDCOR, station blackout the dominant one for ASEP.

K.Vavrek(}{}presentedanoverviewofthePWR-IPEMandindicatedthat revisions to the methodology were only in.the form of an expansion of the earlier model. He discussed the plant walk-through checklist and the control room / man-machine interfaces system interaction checklist.

He mentioned that the IPEM document would be expanded in the following l

manner:

(1) Section/ Chapter 3.0 would be the User's Guide, (2) Appendix

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4 Severe Accidents Meeting Minutes December 19, 1986 C would address system interaction, and (3) Appendix (X_) would address internal. flooding.

R. Henry (FAI) reviewed the IPE source term methodology. As with the PWR-IPEM, significant revisions to the source term methodology were not apparent. He discussed the severe accident sequences considered for the Zion, Indian Point 2, Peach Bottom and Limerick analyses.

F.Coffman(NRR)presentedtheNRC/NRRcommentsontheIDCOR-IPEM. A major concern to NRR was that the IDCOR analysis failed to identify any vulnerabilities in the reference plants. He indicated that the IDCOR position was derived from a 1983 draft of the safety goals while the NRC position was based on the 1986 final version of the safety goals. He discussed the numerical differences in the two positions and indicated that IDCOR was less conservative than NRC. He further discussed the two positions in terms of consistency with generic resolutions, specifically in details regarding the use of the MAAP code and consideration of uncertainties. The IDCOR results appear to be insensitive to uncertainties, thus IDCOR feels that uncertainties can be ignored. He indicated that NRC was concerned about non-uniform results obtained with the IDCOR methodology when applied to similar plants by the same personnel. There were unexpected differences. NRC feels that the IDCOR-IPEM is not yet complete and in its present form, allows too many options. A schedule for the NRC evaluation of the IPEM was discussed.

The present schedule calls for a final evaluation report on January 30, 1987, contingent upon receipt of three or four key IDCOR documents by early January:

revised source term methodology, PWR-IPEM, Sequoyah IPE and Grand Gulf IPE. He concluded with a listing of documents that formed the basis for the IDCOR-IPEM review. At least three of these are forthcoming and have not been provided to the Subcommittee for review.

Rosztoczy (NRR) requested another meeting on this matter later in January 1987 and W. Kerr indicated that while the

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Severe Accidents Meeting Minutes December 19, 1986 request would be considered, it appeared to be premature based on the schedules for IDCOR documentation submittal and NRC review and approval.

During the meeting, Subcommittee members and consultants expressed concerns and opinions as follows:

(1)

W. Kerr questioned the Staff about the application of the safety goal and their assignment of numerical values to it when the Commission had deliberately chosen not to put in specific values.

He also expressed concerns about the selection process based on an aggregate probability of extremely unlikely events being less than E-6.

If one summed a large number of sequences in the E-8 range, the limit (E-6) would be exceeded and all of the low probability sequences would have to be considered. He asked about the qualifications required of the staff that will perform the IPE study.

(2)

P. Shewmon also questioned the selection process based on the aggregate probability.

In regard to the evaluation of only in-ternal initiators at this time, he expressed concerns about the resources required to perform the first IPE, to be followed in the near term with another IPE which considers external events as well.

He thinks that direct containment heating (DCH) is incredible, and that the review of DCH needs to be performed by someone with an understanding of heat flow and failure mechanisms.

(3)

M. Carbcn questioned the goai of the proposed generic letter and indicated that for low probability sequences, he felt that the letter went beyond the safety goal.

(4)

C. Mark asked if the IPE would provide assurance that the plant was built properly. He also questioned how operator performance was evaluated, e.g., a review of operating procedures, an interview of the operators or other.

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, Severe Accidents Meeting Minutes December 19, 1986 (5)

I. Catton asked how Bernero's proposed Mark I containment require-ments are related to the Implementation Plan. He also stated that systems parameter display should be a line item on the PWR human factors checklist.

(6). M. Corradini questioned the Staff about comparison of the IPE results. There seems to be a disconnect between the BNL guidelines and criteria and the IPE results.

i (7)

P. Davis expressed a concern related to on-going programs at NRC, e.g., resolution of station blackout, decay heat removal and ATWS, and a possible conflict between new requirements from these pro-grams and the guidelines and criteria. He also indicated that the IDCOR-IPEMs do not seem to be directed toward giving CDF and containment failure probability values in a manner desired by NRC.

He doubts that NRC will get an approved IPEM from IDCOR. He predicts that the IPEM will need to be modified by NRC for-acceptability.

(8)

J. Lee questioned the Staff's reason for not providing a list of instrumentation and operating procedures for severe accident management. He asked if there were hardware modifications that could be made to relieve operator stress. Also, he asked if BNL f

had compared their guidelines and criteria with the IDCOR IPE for l

any reference plant.

In regard to the systems interaction checklist for PWRs, he asked if or how the IPEM was set up to use the six decision factors.

NOTE:

Additional meeting details can be obtained from a transcript 4

of this meeting available in the NRC Public Document Room, 1717 H Street, N.W., Washington, D.C., or can be purchased from ACE-Federal Reporters, 444 North Capitol Street, Wash-ington, DC 20001, (202) 347-3700.

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v REVISED:

12/11/86 ACRS Severe (Class 9) Accidents Subcommittee December 19, 1986 Washington, DC Individual Plant Examination Generic Letter A.

Subcommittee Chairman's Remarks W. Kerr 8:30am B.

Introductory Remarks T. Speis, NRR 8:40am C.

Generic Letter 1.

Structure and Content R. Landry, NRR 8:55am D.

Scope of the Individual Plant

2. Rosztoczy, NRR 9:35am Examination Performed by Licensees

- - Break ***

10:35 - 10:45am E.

Guidelines and Criteria T. Pratt, BNL 10:45am

BWR Mark I

EWR Mark II

BWR Mark III

PWR Large Dry

PWR Ice Condenser

- - Lunch ***

12:30 - 1:30pm F.

IDCOR-IPEM Revisions and J. Carter, IDCOR 1:30pm i

Responses to ACRS Comments

BWR Methodology E. Burns. Delian

PWR Methodology M. Hitchler, W

- - Break ***

3:00 - 3:10pm

Source Term R. Henry, FAI Methodology G.

NRC Comments on IDCOR-IPEM F. Coffman, NRR 4:15pm l

H.

Concluding Remarks W. Kerr 4:30pm

- - Adjourn ***

4:45pm

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NRR STAFF PRESENTATION TO THE ACRS

SUBJECT:

INTRODUCTORY REMARKS DATE:

DECEMBER 19, 1986 PRESENTER:

THEMIS P. SPEIS PRESENTER'S TITLE / BRANCH /DIV:

DIRECTOR DIVISION OF SAFETY REVIEW & OVERSIGHT OFFICE OF NUCLEAR REACTOR REGULATION PRESENTER'S NRC TEL. NO.:

492-7517 SUBCOMMITTEE:

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DISCUSSION TOPICS GENERIC LETTER FOR EXISTING PLANTS SCOPE OF INDIVIDUAL PLANT EXAMINATIONS GUIDELINES AND CRITERIA IDCOR-IPEM REVISIONS AND RESOLUTION OF CONCERNS f

NRC COMMENTS ON THE IDCOR-IPEM l

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e-II EXAMINATION OF EXISTING PLANTS FOR SEVERE ACCIDENT VULNERABILITIES ACCOMPLISHMENTS GENERIC LETTER SCOPE OF INDIVIDUAL PLANT EXAMINATION GUIDELINES AND CRITERIA

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BOTH IDCOP AND NRC UPDATED AVAILABLE PRA STUDIES FOR FOUR REFERENCE PLANTS, o

NRC QUANTIFIED UNCERTAINTIES ASSOCIATED WITH THE RISK ASSESSMENTS, o

NRC AND IDCOR IDENTIFIED 18 TECHNICAL ISSUES OF CONCERN, 16 0F THESE ISSUES HAVE BEEN RESOLVED, o

APPROXIMATELY 20 IDCOR/NRC TECHNICAL EXCHANGE MEETINGS, o

IDCOR DEVELOPED METHODOLOGY FOR SYSTEMATIC PLANT EXAMINATION, i

o NRC DEVELOPED GUIDELINES AND CRITERIA FOR THE PLANT

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DOCUMENTATION STATUS RES 18/20 REPORTS RECEIVED 5 IN FINAL FORM IDCOR IPEM 15 REPORTS RECEIVED 13 IN DRAFT FORM TEST APPLICATION PLANT CASES RECEIVED

ACRS 3 MEMOS DEALING WITH SEVERE ACCIDENT IMPLEMENTATION CONSULTANTS" REVIEW 0F IDCOR-IPEM

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ASSESSMFT OF ACCIDENT PREVENTION - DESIGN AND O ERATION ACCIDENT MITIGATION - DESIGN AND EERGENCY ACTIONS:

RESul.fS WILL BE EASURED AGAINST GUIDELINES AND CRITERIA SAFETY GOAL POLICY STATEMENT Wi.NERABILITY THE FAILURE TO FULFILL ANY NECESSARY PREVENTIVE OR 5

MITIGATIVE FUNCTION (HARDWARE, HUMAN ACTION OR PROCEDURE)

SPECIFIED IN THE PROPOSED CRITERIA, OP EQUIVALENT CRITERIA x

ACCEPTABLE ETHODS IDCOR IPEM AS APPROVED LEVEL II OR III PRA WITH UPDATE LEVEL I PRA TOGETHER WITH SOURCE TERM IPEM, OR EQUIVALENT SIMPLIFIED, OR PHASE I, PRA WITH NRC APPROVAL OTHER SYSTEMATIC EVALUATION ETHOD WITH NRC APPROVAL

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PLANTS WITH LEVEL II OR III PRA AND IPEM TEST-APPLICATION PLANTS-7 TO 14 M0hTHS AFTER GENERIC LETTER

OTHER PLANTS 14 TO 20 MONTHS AFTER GENERIC LETTER

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' LICENSEES SLEMIT ACTUAL SCHEDULES 60 DAYS AFTER GENERIC LETTER NRC-LICENSEE INTEPFACES DURING PERFORMANCE OF I W

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SHORTLY AFTER ISSUANCE OF GENERIC LEITER EGI0r AL MEETINGS WITH LICENSEES TO DISCUSS REQUEST

QUARTERLY MEETINGS WITH LICENSES IN BETHESDA TO DISCUSS QUESTIONS AND CLARIFICATIONS.

EETINGS WITH UTILITIES OR GROUPS OF UTILITIES AS REQUIRED l

NRR STAFF PRESENTATION TO THE ACRS

SUBJECT:

5 COPE OF THE INDIVIDUAL PLANT EXAMINATION DATE:

DECEMBER 19, 1986 PRESENTER:

Z0LTAN R'.

ROSZTOCZY PRESENTER'S TITLE / BRANCH /DIV:

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DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.:

492-8016 SUBCOMMITTEE: CLASS 9 ACCIDENTS e

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o DEVELOP AN UNDERSTANDING OF WHAT COULD POSSIBLY G0 WRONG IN THE PLANT i

o BE PREPAkED TO MANDLE THESE EVENTS l

EXAMINATION PROCESS o

A THOROUGH, SYSTEMATIC EXAMINATION OF PLANT DESIGN, OPERATION, MAINTENANCE AND EMERGENCY OPERATION o

IDENTIFICATION OF DESIRED PLANT hTTRIBUTES AND ACCIDENT MANAGEMENT MEASURES o

IDENTIFICATION OF POTENTIAL IMPROVEMENTS IN AREAS WHERE DESIRED FEATURES ARE NOT IN PLACE o

DECISION ON POTENTIAL IMPROVEMENTS

CRITERIA TO IDENTIFY DESIRED ATTRIBUTES o

SAFETY G0AL POLICY STATEMENT

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LICENSEE'S.PR,0 POSED CRITERIA o

SCREENING CRITERIA DECISION CRITERIA o

LICENSEE'S JUDGEMENT o

BACKFIT RULE EXPECTED ACCOMPLISHMENTS o

IDENTIFICATION AND ASSESSMENT OF POTENTIAL SEVERE ACCIDENTS o

PLANT IMPROVEMENTS TO PREVENT SEVERE ACCIDENTS o

IMPROVED CONTAINMENT PERFORMANCE o

DEVELOPMENT AND IMPLEMENTATION OF AN ACCIDENT MANAGEMENT PROGRAM

IDENTIFICATION OF P0TENTIAL SEVERE ACCIDENTS SELECTION OF DOMINANT SEQUENCES o

DEFINITION OF A SEVERE ACCIDENT:

AN ACCIDENT THAT RESULTS IN SEVERE CORE DAMAGE (SUBSTANTIAL CORE MELTING, SIGNIFICANT FRACTION OF; FISSION PRODUCTS' RELEASED FROM FUEL) 0 EXTREMELY UNLIKELY EVENTS NEED NOT BE CONSIDERED (AGGREGATE PROBABILITY OF EXTREMELY UNLIKELY EVENTS SHOULD BE LESS THAN 10-6 )

o IDENTIFY SEQUENCES THAT COULD LEAD TO SEVERE CORE DAMAGE o

REVIEW EACH SYSTEM RELATED SAFETY TO DETERMINE UNAVAILABILITY PERFORM WALKDOWNS AS NEEDED o

CALCULATE SEQUENCE FREQUENCY AND CORE DESIGN FREQUENCY o

PREDICT CONTAINMENT LOADINGS AND CONTAINMENT PERFORMANCE FOR EACH OF THE SEVERE CORE DAMAGE SEQUENCES q

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USE SCREENING CRITERIA TO IDENTIFY DOMINANT SEVERE ACCIDENTS o

BASE ASSESSMENT OF PREVENTION, MITIGATION AND ACCIDENT MANAGEMENT ON DOMINANT SEQUENCES

NRC SCREENING CRITERIA FOR DOMINANT SEQUENCES o

CONSIDERATION OF CORE DAMAGE FREQUENCY CDFJS2.10-6 PER YEAR

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CONTRIBUTION TO CDF IS GREATER THAN 5%

0 CONSIDERATION OF CONTAINMENT PERFORMANCE CONTAINMENT FAILS IN LESS THAN 1 DAY CONTAINMENT BYPASS PROBABILITY IS GREATER THAN 10-7 PER YEAR CONTAINMENT FAILS IN LESS THAN 10 DAYS AND RELEASE IS NOT FILTERED i

o ENGINEERING JUDGEMENT SEQUENCES IMPORTANT FOR PLANT DESIGN SEQUENCES IMPORTANT FOR ACCIDENT MANAGEMENT

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DOMINANT SEVERE ACCIDENT SEQUENCES BWR MARK I REFERENCE PLANT o

STATION BLACK 0UT CDF = J 28,1,10-6 PER YEAR 88% OF CDF e

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LOSS OF DECAY HEAT REMOVAL DOMINANT IN PREVIOUS STUDIES l

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DETERMINISTIC: APPROACH TO ACCIDENT PREVENTION

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POTENTIAL PLANT IMPROVEMENTS - BWR MARK I REFERENCE PLANT

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NRC ESTIMATE:

9,9 x 10-6 PER YEAR o

IDCOR ESTIMATE:

8,1 x 10-6 PER YEAR f

o UNCERTAINTY BAND AS DETERMINED BY NRC 2.9 x 10-5 TO 1,6 x 10-6 PER YEAR 4

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CONTAINMENT PERFORMANCE ARE DESIRABLE LARGE RELEASE FREQUENCY <,10-6 PER YEAR, CONTAINMENT PERFOR-0 MANCE IS SATISFACTORY, CHECK MAJOR CONTRIBUTORS TO LARGE RELEASE l

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PREVENTION OF CONTAINMENT FAILURE IS DESIRABLE 0

LIKELIHOOD OF MAJOR CONTAINMENT BYPASS EVENTS SHOULD BE o

KEPT AS LOW AS PRACTICAL 1

RELEASE THROUGH CONTAINMENT LEAKAGE AND CONTAINMENT VEN o

SHOULD NOT ENDANGER PUBLIC HEALTH AND SAFETY - PART l

I

SEVERE ACCIDENT GUIDELINES AND CRITERIA FOR THE FIVE REFERENCE PLANTS DEPARTMENT OF NUCLEAR ENERGY BROOKHAVEN NATIONAL LABORATORY UPTON, NY 11973 PRESENTED AT ACRS SUBCOMMITTEE MEETING l

ON CLASS 9 ACCIDENTS DECEMBER 19, 1986 I

BROOKHAVEN NATIONAL LABORATORY l)l)

ASSOCIATED UNIVERSITIES, INC.(llll

j OUTLINE BACKGROUND AND OBJECTIVES

SUMMARY

GUIDELINES AND CRITERIA FOR FIVE REFERENCE PLANTS:

PREVENTION MITIGATION ADDITIONAL INFORMATION:

SPECIFIC GUIDELINES AND CRITERIA FOR EACH REFERENCE PLANT BROOKHAVEN Nail 0NAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(1lll

GUIDELINES AND CRITERIA FOR THE INDIVIDUAL PLANT EXAMINATIONS l

OBJECTIVES:

1 i

TO IDENTIFY FEATURES OF THE PLANTS THAT INFLUENCE SEVERE ACCIDENTS AND TO PROVIDE FOR THE DETERMINISTIC AND VERIFIABLE BASES AGAINST WHICH TO JUDGE POTENTIAL VULNERABILITIES TO SEVERE ACCIDENTS BY PARTICULAR PLANT TYPES.

APPROACH:

TO DEVELOP THE GUIDELINES AND CRITERIA FROM INSIGHTS i

DERIVED FROM PAST PRAs AND OTHER AVAILABLE SEVERE ACCIDENT INFORMATION TO MAINTAIN A BALANCE BETWEEN BOTH SEVERE ACCIDENT PREVENTION AND CONSEQUENCE MITIGATION WITH THE APPROPRIATE WEIGHT BEING GIVEN TO CONTAINMENT PERFORMANCE BROOKHAVEN Nail 0NAL LABORATORY l}l)l AS500ATED UNIVERSITIES, INC.(llll

EVALUATION PROCESS SCREENING OBJECTIVE:

TO SEPARATE THE POTENTIALLY IMPORTANT SEQUENCES FROM UNIMPORTANT SEQUENCES CRITERIA:

1 SEQUENCE CDF GREATER THAN 1E-6/RY.

2 SEQUENCE CDF GREATER THAN 510F THE TOTAL CDF.

(NOTE:

INTERNAL EVENTS ONLY) 3 CONDITIONAL FAILURE PROBABILITY FOR CONTAINMENT WITHIN ONE DAY GIVEN VESSEL PENETRATION GREATER THAN 0 1 4

PROBABILITY OF CONTAINMENT BYPASS GREATER THAN 1E-7/RY.

5 SEQUENCES JUDGED TO BE UNIQUELY IMPORTANT, E.G.,

VERY SEVERE CONSEQUENCES.

i i

BROOKHAVEN Nail 0NAL LABORATORY l} g)l AS50 GATED UNIVERSITIES, INC.(Itil

+

SUMMARY

OF GUIDELINES AND CRITERIA GUIDELINES GROUPED ACCORDING TO FUNCTION 1

DETERMINISTIC CRITERIA SUMMARIZED TO ADDRESS GENERAL AREAS OF APPLICATION PLANT TO PLANT COMPARISONS PROVIDED TO HIGHLIGHT SIMILARITIES AND DIFFERENCES i

i BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(1 til

GUIDELINES AND CRITERIA 1

MAINTAIN RCS INTEGRITY GUIDELINES A.

PREVENT OVERPRESSURE n.

PREVENT STEAM c.

PREVENT PUMP (LOW PRESSURE SYSTEMS)

GENERATOR TUBE SEAL LOCA RUPTURE RELATED CRITERIA

- TESTING AND MAINTENANCE

- TECH. SPECS.

- CCW/ESW AVAILABILITY

- RELIEF. CAPABILITY

- EMERGENCY PROCEDURES - EMER6ENCY PROCEDURES

- OPERATOR TRAINING

- TECH. SPECS.

~

- SEAL INJECTION PLANT APPLICATION BWR MARK I YES NOT APPLICABLE NOT APPLICABLE

{

BWR MARK 11 YES NOT APPLICABLE NOT APPLICABLE BNR MARK III YES NOT APPLICABLE NOT APPLICABLE PWR ICE CONDENSER YES YES YES PWR LARGE DRY YES YES YES BR00DiAVW NADONAL UBORATORYl}gyl A5500ATED UNNERSmES,INC.(EEll

~

j s

I GUIDELINES AND CRITERIA

)

2.

MAINTAIN RCS HEAT REMOVAL

~

l j

GUIDELINES A.

AC INDEPENDENT INJECTION s.

HIGN PRESSURE c.

ECCS EQUIPMENT i

(STATION BLACK 0UT MITIGATION)

INJECTION FLOODING AVAILABILITY RELATED CRITERIA

- EQUIPMENT AVAILABILITY

- RECIRC. PROCS.

- SEPARATION

- EMERGENCY PROCEDURES

- RECIRC. COOLING

- ELECTRICAL EVAL.

j

- OPERATOR TRAINING

- CONTAINMENT HEAT - EMERGENCY PROCS.

i REMOVAL 4

1 PLANT APPLICATION f

j MARK I YES N/A (To BE ADDED) i MARK 11 YES N/A YES 1

l MARK 111 YES N/A NOT IDENTIFIED AS

\\

RISK SIGNIFICANT 1

PWR ICE CONDENSER NO YES FOR OTHER PLANTS, (REFER TO 2E)

BUT FLOODING PWR LARGE DRY NO YES LINE WILL BE ADDED.

(REFER TO 2E) j s

BROOKHAVEN NATIONAL LABORATORY l)gyl l

1 E500ATED UNIVERSITIES, INC.(EIll i

e

)

GUIDELINES AND CRITERIA 2

MAINTAIN RCS HEAT REMOVAL (CONT'D)

GUIDELINES.

n.

LOW PRESSURE INJECTION E.

AUXILIARY FEEDWATER F.

FEED & BLEED AVAILABILITY COOLING RELATED CRITERIA

- DEPRESSURIZATION

- DIVERSITY

- EQUIP. AVAILABILITY

- RECIRC. COOLING

- REDUNDANCY

- TRAINING

- EMERGENCY PROCEDURES

- WATER SUPPLY

- PROCEDURES'

- CONTAINMENT HEAT REMOVAL - TRAINING

- WATER SUPPLY

- VENTING

- PROCEDURES PLANT APPLICATION MARK I YES N/A N/A MARK 11 YES N/A N/A MARK 111 YES N/A N/A PWR ICE CONDENSER N/A YES YES PWR LARGE DRY N/A YES YES BR0010MVEN NATIONAL LABORATORY l} gg l A5500ATED UNIVERSITIES, INC.(IEll J

e

i

. GUIDELINES AND CRITERIA j

3 RPV DEPRESSURIZATION 4

]

GUIDELINES A.

AUTOMATIC ADS s.

SECONDARY BLOWDOWN j

(ELIMINATE PRESSURE j

PERMISSIVE, ETC.)

j RELATED CRITERIA

- TECH. SPECS.

- OPERATOR TRAINING l

- EMERGENCY PROCEDURES

- TRAINING

- EQUIP. AVAILABILITY PLANT APPLICATION 1

BWR MARK I YES NOT APPLICABLE BWR MARK 11 YES NOT APPLICABLE BWR MARK III YES NOT APPLICABLE l

PWR ICE CONDENSER NOT APPLICABLE YES J

PWR LARGE DRY NOT APPLICABLE YES i

I l

BROOKHAVEN WOW LABORATORY)gd j

A550CNED UNMR5mES, INC.(IIll I

, GUIDELINES AND CRITERIA 4

ATWS RESPONSE I

J GUIDELINES A.

OPERATOR RESPONSE AND EQUIPMENT RELATED CRITERIA

- EMERGENCY PROCEDURES

- OPERATOR TRAINING i

- ADS DEFEAT j

PLANT APPLICATION l

BWR MARK I YES i

j BWR MARK 11 YES 1

j BWR MARK 111 YES

\\

PWR ICE CONDENSER YES i

(EXCEPT ADS) i PWR LARGE DRY YES (EXCEPT ADS) t t

BROOKHAVEN NATIONAL LABORATORY l} gy l A5500ATED UNIVERSITIS, INC(1 Ell l

i

l GUIDELIVES AND CRITERIA

~

5 STATION BLACK 0UT RESPONSE l

GUIDELINES A.

OPERATOR RESPONSE AND s.

VENTING EQUIPMENT

)

RELATED CRITERIA

- EMERGENCY PROCEDURES

- CPERATOR TRAINING

- OPERATOR TRAINING l

- EQUIPMENT AVAILABILITY

- EQUIPMENT CAPACITY l

- CONTROL LOCATION

]

PLANT APPLICATION j

BWR MARK I YES YES BWR MARK 11 YES YES BWR MARK Ill YES YES PWR ICE CONDENSER YES STATION BLACK 0UT NOT AS DOMINANT FOR PWRs.

CON-PWR LARGE DRY YES TAINMENT THREAT NOT AS RAPID i

BR000tAVEN NATIONAL LABORATORY l} g)l l

AS500ATED UNIVERSWES, INC.(IIll l

l i

.=

~

GUIDELINES AND CRITERIA 6

EVALUATE SUPPORT SYSTEM FAILURES GUIDELINES A.

EXAMINE SYSTEM INTERDEPENDENCIES RELATED CRITERIA

- ANALYSIS OF SUPPORT SYSTEM FAIL-URE EFFECTS PLANT APPLICATION BNR MARK I YES BWR MARK II YES BWR MARK III YES PWR ICE CONDENSER YES PNR LARGE DRY YES BROOKHAVEN NATIONAL LABORATORY l} gy l A5500ATED UNIVERSmES,INC.(EEll

i l

GUIDELINES AND CRITERIA 7

MAINTAIN CONTAI M ENT INTEGRITY GUIDELINES A.

VENTING s.

PREVENT POOL c.

ASSESS DIRECT n.

CONTAIMENT

{

BYPASS HEATING SPRAY RELATED CRITERIA

- EMERGENCY PROCS. - DEBRIS CONTROL - CONTAIMENT

- DIVE!;SE POWER

- TRAINING

- CONTAINMENT CAPACITY

- EMERGENCY'PROCS.

i

- EQUIP. CAPACITY ISOLATION

- CAVITY GECMETRY

- TRAINING j

- EQUIP. FUNCTION

- DOWNCOMERS

- CAVITY Fl.00 DING

- LONG TERM i

PLANT APPLICAT10N

~

BWR 8. ARK I YES YES*

FCUND NOT IMPORTANT YES

'1 TO RISK BWR MARK 11 YES YES YES l

j BWR MARK III YES ONLY ADDRESSES YES

'm-J.

I

-,.m, CONTAIMENT ISO-

~,

A i

LATION m

PWR ICE CONDENSER MAY BE NEEDED FDR COMTAIMENT ISO-YES -

YES i

LONG TERM CONTAIN-LATION -

HEAT REMOVAL T,

c l

PWR LARGE DRY CONTAliinENTISO-YEd2 ADDiEIONAL SPRAY 1

LATION

' CAPABILITY FOUND

,NOT TO BE iMPORJ TANT TO RIS".

)

ADDRESSES DEBRIS CONTROL AND CONTAIMENT ISOLATION.

p s

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D g

x GUIDELINES AND-CRITERIA i

s m

1 i

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8 CONTROL HYDROGEN BURNING

~

GUIDELINES A.

PREVENT DEINERTING Y s.

IGNITER AVAILABILITY

~.

~

RELATED CRITER1A

- EMER5ENCY PROCEDURES

- DIVERSE POWER

'J

- EQUIPMENT OPERABILITY

- EMERGENC.Y PROCEDURES PLANT APPLICATION BWR MARK 1 YES NO (INERT CONTAINMENT)

BWR MARK 11 YES NO (INERT CONTAINMENT) l BWR MARK III NOT APPLICABLE YES 1

PWR ICE CONDENSER NOT APPLICABLE YES l

PWR LARGE DRY NOT APPLICABLE IGNITERS ARE NOT IMPORTANT jl TO RISK UNLESS SPECIFIC PLANT IS VULNERABLE TO OVERPRESSURE FAILURE BY HYDROGEN BURNING (PENDING RESULTS OF GENERIC ISSUE 121).

)l BROOKHAVEN Nail 0NAL tABORATORYl}g}l A5500ATED UNIVERSITIES, llK.(IIll l

l i

f

~

j GUIDELINES AND CRITERIA

~

J 9

MAINTAIN CONTAINMENT HEAT REMOVAL 1

GUIDELINES A.

ALTERNATE RPV a.

ICE CONDENSER & FANS l

INJECT 10N/ VENTING j

RELATEG CRITERIA

- DIVERSE POWER

- EMERGENCY PROCEDURES i

- LONG TERM WATER SUPPLY

- TRAINING

~

- EMERGENCY PROCEDURES 1

- TRAINING l

{

PLANT APPLICATION BWR MARK I YES N/A g

BWR MARK II YES

-N/A j

BWR MARK 111 YES N/A i

PWR ICE: CONDENSER' AC DEPENDENCE NOT DOMINANT YES CONTRIBUTOR.

ALTERNATIVE PWR LARGE DRY APPROACH SUGGESTED:

(AC FAILURE OF CONTAINMENT FAN INDEPENDENT FEEDWATER)

COOLERS NOT DOMINANT CON-L TRIBUTOR TO RISK o

(

BROOKHAVEN NATIONAL 1A80RATORfl} g)l 1

ASSOCIATED UNIVERSITIES, INC.(E tll i

o s

i i

BWR INDIVIDUAL PLANT 1

EVALUATION

~

l METHODOLOGY e

4 P

l~

d OBJECTIVES OF PRESENTATION:

BRIEF REVIEW OF IPE

. DISCUSSION OF THE PHILOSOPHY OF THE TECHN!QE KEY EVENTS IN IPE DEVELOPMENT PROCESS IPE APPLICATIONS REStA.TS IDENTIFY COPEENTS AND NSOLUTION ON THE BWR METHOD INSIGHTS FROM THE TEST PLANT APPLICATIONS IDENTIFY WHERE SUPPLEMENTAL PROGRAMS MAY ENHANCE THE METHODOLOGY

SUMMARY

l O

W e

\\

I e

T

e e

IDCDR INDIVIDUAL PLANT EVALUATION METHODOLOGY SEQUENCE EVENT SOURCE TERM EVALUATJDN TREE EVALUATION f

\\ f PLANT SPECIFIC EVALLIATI'ON TO IDENTIFY THAT THE PLANT RISK IS IN TEE SAME RANGE AS THAT ESTIMATED IN THE IDCOR CONCLllSIONS.

l

_.----_,,_--__,_m_._..

_-_---.--.,,..----y

SCOPE lNTERNAL" EVENTS:

e TRANSIENTS, ATWS, LOCAs, RARE INITIATORS INTERNAL FLOODS INTERFACING LOCA COMMON MODE FAILURES SUPPORT SYSTEM DEPENDENCIES AND INITIATORS APPR0X1 MATE fETHOD CAPABLE OF EXPAMSION TO LEVEL 1 e

e' DETAlLED EVENT TREES e

ALL 6E BWRs SELECT SUPPORT SYSTEMS ARE EXPLICITLY ADDRESSED:

e ROOM COOLING SERVICE WATER AC POWER DC POWER INSTRIFENT AIR /Kj PROVISION IS MADE FOR ADDITIONAL SUPPORT SYSTEM D e

DEPENDENCIES ADDRESSED FUNCTIONAL l

HUMAN INTERSYSTEM e

PLANT WALKDOWN e

SYSTEM NOTEBOOKS i

OPERATING EXPERIENCE DATA e

e r

y me-yr-i--v---mg-w-- -,,

,,,--m-

-.w.--

.m-,.y e

-w---w.w i, --

m

'i PROCESS INCLUDES e

SYSTEM NOTEBOOKS:

REFERENCE SOURCE OF INFORMATION ABOUT THE PLANT e

EVENT TREES:

FRAMEWORK FOR IDENTIFYING AND EVALUATING ACCIDENT SEQUENCES e

FAULT TREES:

FOCAL POINT FOR PLANT SPECIFIC DESIGN, OPERATION, MAINTENANCE, AND TEST INFORMATION e

SUPPORT SYSTEM DEPENDENCY MATRICES:-

IDENTIFY AND MODEL IMPORTANT INTERACTIONS e

AVAILABLE DATA:

00ANTIFY THE E0DELS (GENERIC DATA, AND REFERENCE PLANT DATA IS PROVIDED IF PLANT SPECIFIC DATA IS NOT AVAILABLE) e ENGINEERING INSIGHTS:

~ PROBE FOR PLANT UNIQUE FEATURES OR P0TENTIALLY VULNERABLE AREAS

u

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et I

31 ll t

1I p,li Ili Il3 Il

=

1

e l

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INPUT TO CONTAINMENT EVENT TREE EVALUATION ACCIDENT SEQUENCE END STATES e

CORE MELT e

SUPPORT SYSTEM CONFIGURATION e

CONTAINMENT STATUS TEMPERATURE PRESSURE INTEGRITY e

RPV STATUS REACTIVITY CONTROL PRESSURE 4

PRELIMINARY RESULTS OF IPE VERIFICATI.0N PHASE i

IPE COMMENTS IFDATES

1. NEED EXPLICIT

1. PROVIDE:

ROAD MAP OF THE DIRECTIONS ON PROVIDE NUMERICAL MECHANICS OF PROCESS EXAMPLES (COMPLETE) l

2. NEED PRA EXPERTISE

2. MODIFY RECOMMENDED TEAM ON TEAM MAKE UP "O IDENTIFY PRA PERSON, l.E. DELETE ~

"0PTIONAL" (COMPLETE)

3. CLARlFY THE USE OF

3. INCORPORATE SOME QUANTITATIVE (IN PROGRESS)

ESTIMATES

4. CLARIFY SELECTED

4. UPDATE THE METHODOLOGY ENGINEERING INSIGHTS (COMPLETE)

TO MAKE THE ISSUES OBVIOUS AND THE ACTIONS CLEAR

5. PROVIDE EXAMPLES OF

5. UPDATE THE ETHODOLOGY THE DEPENDENCY (COMPLETE)

MATRICES

6. PROVIDE ADDITIONAL

6. IIPDATE THE METHODOLOGY EXAMPLES OF SERVICE

(. COMPLETE) i WATER SYSTEMS FOR l

QUANTIFICATION

7. CORRECT ERRORS

7. IIP 9 ATE THE METHODOLOGY (COMPLETE)

8. MAY REQUIRE MORE

8. INCLUDE DISCUSSION IN TIME THAN IDENTIFIED METHODOLOGY (COMPLETE) e

\\

e f

ERSGDTICH W 1GC 030EH1S W

BWR DE (sarnmBER ises)

COe4ENT SYSTBEi MES m

-1.

SAFELY GOAL COEISTDCT O

E CDeG SERSURE FOR CDEWtISW O

m21R XIXMPLISH SDWAR SE!CTIVE 2.

MATEING CRITERIA FOR O

IPE IDENTIFIES mmTT.TTIES VARICIJS LEVIIS AFFICTDG OVERALL EEVEL W EAFEIY O

GLY FtBCTICMAL IEVEL MEEDG IS M39tDGIL O

INSIGf1S F50VIIED IN AFFEICIX D l

3.

CEARACTERIZE TENTDG FCR O

IFT21rsn 3 mPDOIX D MARK I AND II O

ENGOReG EBUS1EE FKERADEi MILL PROGDE EDETEterL GUIDAM2 4.

EXAMPLES T FLANT SPICIFIC O

E *GHLIE3tS* IDENTIFIED vuLNERABILITIIS O

EIGH-IEVEL CEE3 LIST C

APPDOIX D

. d~

\\

4 RI!SCLUTIN & IOC COGENIS W

IMR IPE (SEPTE!MBER 1986)

C09ENT SYSTIBE AIN2 SIS m 5.

EQUIRENT SURVIVhBILITY O

MEINTIAL EQUIMENT EENTIFIED IN suuNT TREES AfD APPBCII D 0

SEDMCE IICLIEED 3R AWHBOIX D 6.

VISML DEPBCTICE O

DISCUSSIN ENENCBD PRCCIDURES 7.

DCCGENTATION 3GQUIREMENIS O

ElEIT3CIN WIDMCE PEWIDEID TO EFTOItT DETUDEL UTILIW OSE Alm REVIEN l

8.

SCIHCE TEIM ME'HiCDrvro O

SEQUDCES BI50ED INTERFACE O

5ERESIILD 0

Etc SMTE 00tOITRalS O

GREEDMENT EVENT TitEE a

.E

WERIFICATION PHASE o

PURPOSE:

TEST THE BWR IPE ETH000 LOGY AND VERIFY ITS USABILITY i

)

o BWR PLANTS IDENTIFIED FOR VERIFICATION PHASE 1

PEACH BOTTOM (BWR/4) M RK I SUSQUEHANNA (BWR/4) M RK II SHOREHAM (BWR/4) M5RK II GRAND GULF (BWR/6) MRK III o

RESULTS SUBMITTED TO MRC AS PACKAGE WITH THE UPDATED IPE METHOD IN MAY 1986 (GRAND GULF DECEMBER 1986)

O e

,.yg

-,_,,,_m.

-_,.y%.

_y.

,,,,,,,,,%,,,,,,,.,,c m,,-

,,,,y.,

y,.,

,,,c

,-n,,

BWR PE ACCOEPR SEQUENCE EVALLRTION ESTIP4TED >WPOWER M I

CALENDAR PLANT UTILITY CONSULTANT MOIENS SHOREHAM 20 4

4 PEACH BOTTOM 16

'12 B

SUSQUEHANt%

24 6

GRAND GULF 45*

3' 5

GRAND GULF EFFORT INCLLDED TASKS NOT SPECIFIED IN IPEM.

e

9 a

BWR IPE ACCIDENT SEQUENCE EVALUATION CORE t<LT FREQUENCY (PER YR.)

PRA IPE PLANT AVAILABLE PRA RESULTS SHOREHAM YES 5E-5 BE-5 PEACH BOTTOM YES (WASH 1400) 3E-5 4E-5 SUSQUEHANNA

-YES 1E-5 3E-5 (1) NOT PUBLISHED

=Y g

e PEACH BOTTott ATOMIC 700fER STATI0tt CDF COMPARISON SEQU5NCE ASEP IPE TYPE

_CIASS PROGRAM STUDY TQUV &

IA & ID 6.8E-8/YR 1.4E-5/YR TQUX TB IB 8.7E-6/YR 4.6E-6/YR TW II

-1.0E-8/YR 7.4E-7/YR AE &

III

-1.1E-7/YR

~1.5E-6/YR SE 1

TC IV 1.9E-6/YR 4.1E-6/YR+

TOTAL 9.9E-4

-2.0E-5

+

TC:

CIASS IC 5.9E-7/YR CIASS III 1.1E-9/YR CIASS IV 3.5E-6/YR TOTAL 4.iE-6/YR u

J

e G

e 9

i.

C05tE DAMAGE FREQUENCY DOMINANT SEQUENCE CONTRIBUTION (IN PERCENT)

ASEP PEACH BOTTOM PROGRAM +

IPE STUDY ++

STATION 88%

39%

BLACKOUT ATWS 10%

19%

42%+++

l RENAINING 2%

SEQUENCES l

+

CDF = 9.9E-6/RX TR

++

CDF = 2.2E-5/RX TR MAJOR CONTRIBUTICet FROM TRANSIENT INITIATED SEQUENCES

+++

2 l

4 d

t 530thRY &

SIE asem O.

DElVEIiPED 'IO CMGIATE A RElhLISTIC MAIIT WE3FK: CIEE NET FR100DCY O

BASED W INSIGEIS FROI PAST PRhs AIC E(IR O

IS NOT A PRA O

IS USABLE FOR CCDMUNICATIN 'to MMEGBENT O

PMNIDE UTILIH A RISK PERS1CTIVE W 'EHEIR INVES' DENT O

PMNIDES REAL IEIGEEIS IMIO DEIOPBGES, EKNEVER EBIALL, SEAT CAN BE DCBE: IKN BY 'IBE UTILIR 50 m M PIANT SAFER O

IDENTIFT 70TENTIJIL MPEIIR O

QUANTDT 'HE: (IRE IGLT FRIQUEECT BY 'Ef7E W SENESCE O

EXPAMMEEE TO 12 VEL 1 PRA O

IS EASILY G M TED IF IMOEMATIN BICOtES AVAIIABIE IN 'IBE FUIURE, E.G. MANT SPECIFIC Dh33L i

i 9

9 m

O 4

..,.__,r_,%-

rr

,,rv-,, -.

.o a

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D 9

9 4

PWR INDIVIDUAL PLANT EVALUATION STATUS DECEMBER 18/19 1986 WESTINGHOUSE ELECTRIC CORPORATION KEN J. VAVREK i

(412) 374 4028 E

o-4 l

e

r THE METHODOLOGY APPROACH INCLUDES THE FOLLOWING COMPONENTS:

0 INITIATING EVENTS 0

EVENT TREES / ACCIDENT SEQUENCES 0

SUCCESS CRITERIA 0

SUPPORT SYSTEMS 0

FAULT TREES / SYSTEMS ANALYSIS 0

FAILURE DATA l

e e

7

PLANT SPECIFIC IMPLEMENTATION FLOW DIAGRAM l

SYS B AM LYSIS S WiTIFICATIM M TA

)

FAllWE MTES TEST MINIDWEE

'WES F CIDENT SERHCES INITIATES SIFPETSTATEET SIFPWT F BIT M SYSIBl DAINE SIAIES i

1HANSIENTS SIATE A

E i

LOCAs s

s B

At.

B E

SPECIAL INITIATORS C

E pggggy 0F CEE IEl.T l

0 9-4 IPE METHODOLOGY REVISIONS AREAS OF NEW IPE DOCUMENTATION METHODS MANAGEMENT SYSTEM INTERACTION EVALUATION INTERNAL FLOODING EVALUATION AREAS OF EXPANDED IPE DOCUMENTATION COMMON CAUSE FAILURE TREATMENT LOCA INITIATING EVENT FREQUENCIES EVENT TREE TEMPLATES ACCIDENT SEQUENCE NOTEBOOK l

u

?

l l

,,~.,

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

f

~

METHODS MANAGEMENT 9

COORDINATION PROJECT LEADER & RESPONSIBILITIES PLANNING SELECTION OF PROJECT PARTICIPANTS ALLOCATION OF RESOURCES PER TASK SCHEDULING SCHEDULING OF PROJECT TASKS TECHNICAL QUALITY CONSISTENT DOCUMENTATION OF IPE ANALYSIS 9

Y e

o n

v

SYSTEM INTERACTION EVALUATION PLANT WALKTHROUGH ENVIRONMENTAL CONDUCTORS CHECKLIST CONTROL ROOM MAN-MACHINE INTERFACES CHECKLIST PLANT TALKTHROUGH NORMAL, EMERGENCY, TEST & MAINTENANCE PROCEDURES CHECKLIST L

1 l

l'

INTERNAL FLOODING EVALUATION OUALITATIVE EVALUATION DETERMINATION OF POTENTIAL FLOODING SOURCES COMPONENTS AFFECTED BY FLOODING SOURCES DESCRIPTION OF SOURCES & POSSIBLE MITIGATING ACTIONS MAPPING OF CONSEQUENCES OF FLOODING POTENTIAL PLANT AREAS IDENTIFIED QUANTITATIVE EVALUATION INITIATING EVENT FREQUENCIES MITIGATING ACTION PROBABILITIES ACCIDENT SEQUENCE QUANTIFICATION IDENTIFICATION OF DOMINANT ACCIDENT SEQUENCES e

l-l I;.

~

TABLE D.4-4 INTERNAL FLOODING INITIATING EVENT DATA BASE Component Failure Mode Failure Rate Source Air Operated Valve Rupture 2.0E-7/hr WUREG-1363 g.

Manual Valve Rupture 1.3E-8/hr NUREG-1363 S.

Motor-Operated Valve Rupture 1.7E-7/hr NUREG-1363 Check Valves Rupture 5.2E-8/hr NUREG-1363 Tank Rupture 8.6E-10/hr WASH-1400 Piping (> 3" Diameter)

Rupture 8.6E-10/section-hr WASH-1400

(< 3" Diameter) 8.6E-9/section-hr WASH-1400 7

t Expansion Joints Rupture 2.5E-4/ expansion Oconee 3 PRA jointyear e

(

4 a:54mionaosas D-16

SOURCE TERM METHODOLOGY R. E. Henry J. R. Gabor M. A. Kenton Fauske & Associates, Inc.

Burr Ridge, Illinois Presentation to:

ACRS Severe '(Diass 9)

Accident Subcommittee December 19,1986 Washington, D.C.

4 IPE SOURCE TERM METHODOLOGY

.v e

Designed to search for potential outliers for containment behavior (fission product retention) under severe accident conditions.

e Focus is on major mechanisms for fission product retention.

p:

s Quenching of debris and containment heat removal.

?

1.

Containment ' sprays.

2.

Containment fan coolers where applicable.

i Wetwell venting where applicable.

Deposition in containment.

Deposition in adjacent buildings.

s f

.V

q x

s~

34

' x i

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'i g

IPE SOURCE TERM METHODOLOGY

.i e

Uses streamlined containment event trees.

e Uses likelihood of occurrence for each decision t

point.

High - Written procedures with equipment that can be implemented on a timely basis.

Medium - Demonstrated capability with equipment that could be used on a timely basis.

,i Low - No written procedures or no demonstrated capability.

p 1

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l' IPE SOURCE TERM METHODOLOGY.

t

.+

l' t

~

l e IdentJfies the same. controlling features as.. full

,. scope PRAs.

g e 1PE approximate source terms can be developed on a seduence specific ba' sis.

g e.lPE approximate source terms are in agreement with those developed in full scope PRAs.

D e' IPE ' methodology is $ufficient for searching for-outlier conditions or"c'o'nfigurations.

l y.

,1 -

r t'

/r 9

1 i

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[

7, r A h

a Table H.1 1

G'.Q 1

.( y SEVERE ACCIDENT SEQUENCES CONSIDERED IN THE IDCOR ZION CONTAINMENT ANALYSES

~

I Environmental IPE Sequence Sequence Source Term Approximate Designation Description Calculated by Source the MAAP Code Term Station Recovery of one No core damage.

Insignificant Blackout vital bus at (containment With 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

Recovery heat removal available)

Station Station blackout No core damage.

Insignificant Blackout with a seal LOCA (containment With with recovery of heat removal Recovery one vital bus at I hour.

available)

Station Station blackout Core damage without Insignificant Blackout with a seal LOCA containment failure.

(containment With with recovery of heat removal Recovery one vital bus at 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

available)

Station Station blackout Core damage, vessel Insignificant '

Blackout with a seal LOCA failure but no con-(containment With with recovery of tainment failure.

heat removal Recovery one vital bus at available) 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Station Station blackout Core damage, vessel Insignificant Blackout with a seal LOCA failure but to con-(containment With with recovery of tainment failure, heat removal Recovery one vital bus at 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months .

available)

Small LOCA Small LOCA with Core damage, vessel Insignificant With initial failure failure but no con-(containment Recovery to achieve recir-tainment failure.

heat ramoval culation for available) injection.

Re-circulation capa-bilities recovered at 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

TableH.1(Continued)

SEVERE ACCIDENT SEQUENCES CONSIDERED IN THE IDCOR ZION CONTAINMENT ANALYSES Environmental IPE Sequence Sequence Source Ters Approximate i

Designation Description Calculated by Source the MAAP Code Term Large LOCA A large break No core damage.

Insignificant with only one (containment charging pump heat removal operational and

'available)

~

limited RWST refill capability.

Station Loss of all AC Csl - 0.002 CsI < 0.01 Blackout power and aux-Te - 2 x 10-5 Te - 0.004*

With a Seal iliary feedwater LOCA with a 50 gpm per pump seal LOCA at 45 minutes.

. Station Loss of all AC CsI - 0.002 Cs! < 0.01 jBlackout power and aux-Te - 2 x 10-5 Te - 0.004*

j iliary feedwater without a seal LOCA.

Small LOCA Small (2 inch)

Core damage, vessel Insignificant cold leg break, failure but no con-(containment failure of ECCS tainment failure.

heat removal recirculation but available) fan coolers and l

containment sprays are available.

Large Break Large cold leg Core damage, vessel Insignificant LOCA break, failure of failure but no con-(containment ECCS recirculation tainment failure.

heat removal but fan coolers available) and containment sprays are avail-able.

V Sequence'

'Assiimed failure of Noble gases.

Noble gases.

the isolation valve discs and consequential failure of the RHR pump seals.

5r and Ba assumed equal.

Table L-1 SEVERE ACCIDENT SEQUENCES CONSIDERED IN THE IDCOR PEACH BOTTOM CONTAINMENT ANALYSES Environmental IPE Sequence Sequence Source Term Approximate Designation Description Calculated by Source the MAAP Code Term TW With Loss of suppres-No core Insignificant Recovery sion pool cooling degradation.

(containment with refilling of heat removal the condensate available) storage tank and containment venting.

TC With ATWS with injec-No core damage.

Insignificant Recovery tion flow throt-(containment tied to TAF and heat removal containment available)

Small LOCA Small LOCA with Core damage without Insignificant With failure of injec-containment failure.

(containment Failure of tion.

Crywell heat removal Injection sprays initiated available) at 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months .

venting.

Transient Transient with No core damage.

Insignificant With Loss loss of injection (containment-of Low and containment heat removal Pressure heat removal with available)

Injection on-site power re-and Con-stored at 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

tainment Heat Re-moval With Recovery Transient Transient with NG - 1.0 NG - 1.0 With Fail-failure of con-I - 0.2 I - 0.1 ure of Heat tainment heat Cs - 0.2 Cs - 0.1 Removal removal and no recovery.

ATWS(TC)

ATWS with no oper-NG - 1.0 NG - 1.0 (Case 1) ator actions taken.

I - 0.1 I - 0.1 Cs - 0.1 Cs - 0.1

.,.._._,.-,.__....-.,,_.-,,._m___

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

.m..

l Table L-1 (Continued)

SEVERE ACCIDENT SEQUENCES CONSIDERED IN THE IDCOR PEACH BOTTOM CONTAINMENT ANALYSES Environmental IPE Sequence Sequence Source Term Approximate Designation Description Calculated by Source the MAAP Code Tem ATWS(TC)

ATWS with wetwell NG - 1.0 NG - 1.0 (Case 2) venting at 115 I - 0.03 I < 0.01 psia.

Cs - 0.03 Cs < 0.01 ATWS(TC)

ATWS with refill NG - 1.0 NG - 1.0 (Case 3) of the CST.

I - 0.03 I < 0.01 Cs - 0.03 Cs < 0.01 ATWS(TC)

ATWS with wetwell NG - 1.0 NG 1.0 (Case 4) venting and refill I - 6 x 10-4 I < 0.01 of the CST.

Cs - 6 x 10-4 Cs < 0.01 Station Station blackout NG - 1.0 NG - 1.0 Blackout without recovery.

I - 0.05

.I - 0.02 Cs - 0.05 Cs - 0.02 Small LOCA Small LOCA with t'G - 1.0 NG - 1.0 Without failure of all I - 0.04 I - 0.02 Injection injection and no Cs - 0.04 Cs - 0.02 recovery.

i l

.Y l

NRR STAFF PRESENTATION TO THE ACRS

SUBJECT:

NRC COMMENTS ON IDCOR IPEM DATE:

DECEMBER 19, 1986 PRESENTER:

FRANKLIN C0FFMAN PRESENTER'S TITLE / BRANCH /DIV:

SECTION LEADER /

REGULATORY IMPROVEMENTS BRANCH /

DSRO PRESENTER'S NRC TEL. NO.: 492-4609 SUBCOMMITTEE:

SEVERE (CLASS 9) ACCIDENTS e

e

4 EVALUATION STANDAPDS FOR SEVERE ACCIDENT METHODS 1.

CAPABILITY TO FIND VULNERABILITIES 2.

CONSISTENCY WITH GENERIC RESOLUTIONS 3.

SYSTEMATIC EXAMINATIONS A,

INTEGRATE CONSIDERATIONS OF SAFETY CONCERNS B.

BALANCE PREVENTIONS AND MITIGATIONS C,

ACHIEVE UNIFORM EXAMINATIONS D.

PROVIDE REASONABLE COMPLETENESS 4.

LIMITATIONS AND CAUTIONS

. 5.

COVERAGE OF GUIDELINES AND CRITERIA 6,

DOCUMENTATION AFD PRESENTATION 7.

APPLICABILITY BETWEEN EXISTING PLANT AND REFERENCE PLANT 8.

ROLE OF VISUAL INSPECTION 9.

COVERAGE OF CURRENT INSIGHTS

=Y

L 1.

CAPABILITY TO FIND VULNERABILITIES A.

IDCOR (POSITION DERIVED FROM 1983 DRAFT SAFETY GOALS)

OUTLIERS

.1, CDF> 3E-4/RY (INTERNAL EVENTS) 2.

SEQUENCE > 30 TIMES SIMILAR IDCOR SEQUENCE 3.

" CHECKLIST OF EXAMPLES" (TABLE 3.2-3) 4.

TWO NEGATIVE ANSWERS ON SIMPLIFIED EVENT TREE POTENTIAL AREAS FOR IMPROVEMENT 1.

KEY SYSTEMS UNAVAILABILITY >> IE-2 FOR INTERNAL INITI ATORS >-

0.1/YR 2.

SEQUENCE t> 0.2 EQUIVALENT IODINE AT > IE-5/RY INTERNAL UTILITY RESTRICTIVE CRITERIA POSSIBLE B.

NRC VULNERABILITIES (UNDER CONSIDERATION) 1.

MEAN CDF POSITION BEING DEVELOPED FROM '86 SAFETY G0AL (BOTH INTERNAL AND EXTERNAL) 2, MEAN PROBABILITY OF LARGE RELEASE 2> IE-6/RY POTENTIAL VULNERABILITIES (UNDER CONSIDERATION)

I.

SEQUENCE CDF >" 2E-6/RY 2.

SEQUENCE > 5% TOTAL CDF 3.

PROB, OF CONT, FAIL, GIVEN CORE DAMAGE 0,01 4.

CONTAINMENT BYPASS SEQUENCES 1> IE-7/RY 5.

UNIQUELY IMPORTANT SEQUENCES C.

TEST APPLICATIONS BY UTILITY AND IDCOR

~

1.

NO " OUTLIERS" WERE DISCOVERED 2,

MANY PLANT-SPECIFIC " INSIGHTS" WERE HIGHLIGHTED d

2.

CONSISTENCY WITH GENERIC RESOLUTIONS-A.

IDCOR USE OF MAAP CODE 1.

REVIEW OF MAAP BEING DISCUSSED WITH NRC MANAGEMENT 2.

NO REVISIONS TO MAAP PLANNED CONSIDERATION OF UNCERTAINTIES 1.

IDCOR IPEM RESULTS ARE INSENSITIVE TO UNCERTAINTIES 2.

UNCERTAINTY STUDY UNDERWAY EVENT "V" CHECKLIST 1.

REVISIONS PLANNED SIMPLIFIED CET REVISIONS 1.

REVISIONS PLANNED B.

NRC USE OF MAAP CODE 1.

MAAP IS NOT A REVIEWED CODE 2.

MAAP REVIEW IS SEPARABLE FROM APPROVAL OF IDCOR IPEM CONSIDERATION OF UNCERTAINTIES 1.

IPE CRITERIA MUST ADDRESS PHENOMEN0 LOGICAL UNCERTAINTIES 1

l l

l

9 3.

SYSTEMATIC EXAMINATIONS A.

IDCOR ACHIEVE UNIFORM EXAMINATIONS 1.

BWR AND PWR METHODS ARE EQUIVALENT i

B.

NRC ACHIEVE UNIFORM EXAMINATIONS 1.

TEST APPLICATIONS REPORTS ARE DIVERSIFIED 2.

TEST APPLICATIONS APPEAR TO HAVE SIGNIFICANT VARIATIONS AMONG UTILITIES EVEN THOUGH IDCOR IPEM CONSULTANTS WERE USED.

3.

NO TEST APPLICATION WITHOUT BEING SUPPLEMENTED BY PRA-TYPE ANALYSES 4.

IDCOR IPEM STILL APPEARS TO BE UNDER DEVELOPMENT 5.

IDCOR IPE METHODS MANAGEMENT ALLOWS MANY OPTIONS WITH THE POTENTIAL TO INTRODUCE DEVIATIONS 6.

SOURCE TERM METHODS ARE SIGNIFICANTLY SIMPLER THAN ACCIDENT SEQUENCE METHODS e %*

DOCUMENTS FOR REVIEW 0F THE IDCOR IPEM:

1. DRAFT IDCOR PROGRAM REPORT, TECHNICAL REPORT 85.3-A1; PWR ACCIDENT SEQUENCE - INDIVIDUAL PLANT EVALUATION METHODOLOGY, APRIL 1986

2. DRAFT IDCOR PROGRAM REPORT, TECHNICAL REPORT 85.3-A2; PWR SOURCE TERM - INDIVIDUAL PLANT EVALUATION METHODOLOGY, APRIL 1986

3. DRAFT IDCOR PROGRAM REPORT, TECHNICAL REPORT FAI/85-58, APPROXIMATE SOURCE TERM METHODOLOGY FOR PRESSURIZED WATER REACT 0PS, FAUSKE 8 ASSOCIATES, DECEMBER 1986

4. DRAFT IDCOR PROGRAM REPORT, TECHNICAL REPORT 85.3-B1; BWR ACCIDENT SEQUENCE - INDIVIDUAL PLANT EVALUATION METHODOLOGY, APRIL 1986

5. DRAFT IDCOR PROGRAM REPORT, TECHNICAL REPORT 85.3-B2; BWR SOURCE TERM - INDIVIDUAL PLANT EVALUATION METHODOLOGY, APRIL 1986 6.

DRAFT IDCOR PROGRAM REPORT, TECHNICAL REPORT FAI/86-1, l

APPR0XIMATE SOURCE TERM METHODOLOGY FOR BOILING WATER REACTORS, FAUSKE 8 ASSOCIATES, DECEMBER 1986 l

7. DRAFT IDCOR PROGRAM REPORT, BWR IPE PLANT SPECIFIC ACCIDENT SEQUENCE EVALUATION METHODOLOGY, USER's GUIDE. REV. 1, DEC 1986

8. DRAFT IDCOR PROGRAM REPORT, INDIVIDUAL PLANT EVALUATION, PEACH BOTTOM' ATOMIC POWER STATION, PHILADELPHIA ELECTRIC COMPANY, MAY 1986

9. DRAFT IDCOR PROGRAM REPORT, INDIVIDUAL PLANT EVALUATION FOR SUSQUEHANNA STEAM ELECTRIC STATION, P.R. HILL, C.A. KUKIELKA, AND C.A. BOSCHETTI, SUBMITTED TO IDCOR APRIL 1986

I e C

' hl 10.

DRAFT IDCOR PROGRAM REPORT, SHOREHAM NUCLEAR POWER STATION, IDCOR INDIVIDUAL PLANT EVALUATION, LONG ISLAND LIGHTING COMPANY, APRIL 1986 11, DRAFT IDCOR PROGRAM REPORT, INDIVIDUAL PLANT EVALUATION METHODOLOGY APPLIED TO THE OCONEE NUCLEAR GENERATING STATION.

SUBMITTED T0:

AIF/IDCOR PROGRAM, MAY 1986 12, DRAFT IDCOR PROGRAM REPORT, INDIVIDUAL PLANT EVALUATION METHODOLOGY APPLIED TO THE ZION NUCLEAR GENERATING STATION, SUBMITTED T0:

AIF/IDCOR PROGRAM, FEBRUARY 1986 13.

DRAFT IDCOR PROGRAM REPORT, IDCOR/IPE REPORT, CALVERT CLIFFS NUCLEAR POWER PLANT UNIT 1, SUBMITTED TO IT CORP BY BG8E WITH LETTER DATED OCTOBER 20, 1986.

14.

LETTER FROM T. P. SPEIS, NRC, TO A. BUHL, IDCOR WITH PRELIMINARY EVALUATION OF THE IDCOR IPEM, SEPTEMBER 9, 1986 15.

LETTER FROM A. BL'4L, IDCOR, TO T,P. SPEIS, NRC

SUBJECT:

IDCOR RESPONSE TO NRC COMMENTS ON THE INDIVIDUAL PLANT EVALUATION METHODOLOGY, DEC 1986 16.

LETTER FROM J.W. HICKMAN, SNL, TO M.D. HOUSTON, ACRS STAFF, TRANSMITTING COMMENTS FROM A REVIEW 0F THE IDOCR IPEM, SEPTEMBER 22, J986 17.

LETTER FROM A. BUHL, IDCOR, TO W. KERR, ACRS,

SUBJECT:

'IDCOR RESPONSES TO SNL COMMENTS ON THE IDCOR IPEM, OCTOBER 30, 1986 l

- -

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