ML20213A545
| ML20213A545 | |
| Person / Time | |
|---|---|
| Issue date: | 10/29/1986 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2462, NUDOCS 8702030340 | |
| Download: ML20213A545 (76) | |
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' o y; ; h~M k DATE ISSUED: 10/9 and 0 2 /86 ACRS Combined Subcommittee Meeting Sumary/ Minutes for Severe (Class 9) Accidents / Nuclear Plant Chemistry September 24, 1986 Washington, DC Purpose The Combined ACRS Subcommittees on Severe (Class 9) Accidents and Nuclear Plant Chemistry met on September 24, 1986 in Washington, DC.
The purpose of this meeting was to discuss portions of the NRR Implemen-tation Plan for the Severe Accident Policy, including proposed regula-tory changes in regard to PWR spray additives and BWR suppression pool scrubbing, and the IDCOR Methodology for Individual Plant Examination.
The Subcommittees heard presentations from members of NRC/NRR and IDCOR.
Copies of the agenda and selected slides from the presentations are attached. The meeting began at 8:30 a.m. and adjourned at 5:40 p.m.,
and was held entirely in open session. The principal attendees were as follows:
Attendees:
ACRS NRC/NRR W. Kerr, Co-Chairman Z. Rosztoczy D. Moeller, Co-Chairman F. Coffman M. Carbon, Member L. Soffer J. Ebersole, Member R. Palla C. Mark, Member F. Eltawila D. Okrent, Member J. Read P. Shewmon, Member T. Pratt, Consultant C. Siess, Member D. Ward, Member IDCOR C. Wylie, Member J. Carter (ITC)
M. Bender, Consultant R. Henry (FAI)
- 1. Catton, Consultant E. Burns (Delian)
M. Corradini, Consultant K. Vavrek (W)
P. Davis, Consultant M. Plesset, Consultant D. Houston, Staff DEsrCNATED ORIGINAL 20 0 861029 Certified By [//g ACRS-2462 PDR --
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t 3 Severe Accidents / Nuclear Plant Chemistry Meeting Minutes September 24, 1986 Discussion The combined Subcommittees on Severe (Class 9) Accidents and Nuclear Plant Chemistry heard presentations by NRC/NRR and IDCOR that were all related to the NRR Implementation Plan for the Severe Accident Policy.
Z. Rosztoczy (NRR) began by reviewing the implementation elements and giving a status report on the progress of these elements. Four of the elements were discussed in detail during the course of the meeting: (1)
Disposition of the NRC/IDCOR Technical Issues, (2) Review of the IDCOR Individual Plant Examination Methods (IPEM), (3) Development of Guidelines and Criteria for the Individual Plant Examinations (IPE), and (4) Source Term Related Changes. Rosztoczy indicated that the NRR program was close to schedule and their findings should be forwarded to the Consnission by the end of the year. Guidance for IPEs should be issued in 2-3 months and containment performance criteria should be out by the end of November 1986. The limits in 10 CFR 100 will be reviewed and Regulatory Guides will be revised using realistic meteorological models.
R. Palla (NRR) presented a brief overview and status report on the NRC/IDCOR technical issues. These issues pertain to major differences in the models which were identified in the Spring of 1985. In all other areas, there was agreement or only minor differences. The issues have not been revisited to determine if the situation has changed due to later research results and model development. He described the categories of issue resolution and discussed in detail the background and status of one issue - No. 15 Containment Performance. He discussed briefly the documentation for issue resolution - (1) summary / minutes of meetings held between NRC/IDCOR and (2) IDCOR Technical Report 85.2,
" Technical Support for Issue Resolution."
Z. Rosztoczy (NRR) discussed the preliminary evaluation of individual plant examinations (IPE) and the application of these results in the l
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e Severe Accidents / Nuclear Plant Chemistry Meeting Minutes September 24, 1986 content of a generic letter for IPEs. T. Pratt (BNL) discussed in detail an evaluation of IDCOR/NRC plant studies (mostly BWR Mark I).
The objective of the BNL evaluation was to identify dominant and potentially important sequences, assess containment performance and develop deterministic guidelines and criteria to assure that severe accident goals are satisfied. He discussed four proposed guidelines for Mark Is - Wetwell Venting, Deinerting, Suppression Pool Bypass and Drywell Sprays. F. Eltawila (NRR) further discussed a Mark I reference plant evaluation, more related to a severe accident evaluation than to part of the IPE review.
Various IDCOR representatives presented the details of the proposed IDCOR IPE Methodology. This methodology was made up of four distinct parts: (1) BWR IPEM, (2) PWR IPEM, (3) BWR Source Tenn, and (4) PWR Source Term. R. Henry (FAI) discussed the way in which the IDCOR program meets the requirements of the NRC Severe Accident Policy Statement, and presented some fundamental conclusions. He discussed the key elements of the PWR and BWR Source Term Methodology and indicated that the uncertainty studies associated with source terms would probably not be completed prior to the IPE search for outliers. E. Burns (Delian) presented the details of the BWR IPE Methodology and K. Vavrek (W) discussed the methodology for PWRs. The IPEMs address the following: internal events, detailed event trees, support systems, system dependencies, accident sequences and failure data. The method-ology was said to be capable of expansion to a Level 1 PRA. R. Henry (FAI) completed the discussion by briefly presenting the source term methodology for BWRs and PWRs. The source tenn methodology used specific PWR and BWR plant characteristics and applied a fault tree analysis to determine releases. The key elements of each model were discussed. F. Coffman (NRR) discussed the NRC comments on the IDCOR IPEM; these had been transmitted by letter to IDCOR on September 9, 1986. The preliminary comments were grouped into 8 categories. The IDCOR response to these conrnents is expected on October 31, 1986 and the
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l Severe Accidents / Nuclear Plant Chemistry Meeting Minutes September 24, 1986 final NRR evaluation report is expected to be issued on December 19, 1986.
L. Soffer and J. Read (NRR) discussed proposed source term related changes to Regulatory Rules and Practice. L. Soffer discussed SECY 86-228, " Introduction of Realistic Source Terms Into Licensing." 1 Basically, the Staff proposes to replace TID-14844, making use of NUREG-0956 to provide tables specifying quantity and rates of release into containment for groups of accidents and differing containment types, to realistically compute or model removal processes, ESF per-formance and leakage and to use the latest models for environmental transport and dosimetry. J. Read discussed proposed revisions to SRP Sections 6.5.2(PWRcontainmentsprayadditives)and6.5.3(BWRsuppres-sion pool scrubbing). The revisions can be summarized as follows:
SRP 6.5.2 - deletes reference to TID-14844 source terms, deletes requirement for additive during injection and relaxes sump solution pH from 8.5 to 7.0.
SRP 6.5.3 - gives credit for suppression pool scrubbing, suggests that pool bypass leak rates be tested periodically, and recommends that standby gas treatment systems be upgraded to ESF level.
The Staff indicated that the SRP revisions would be in final form by November and associated Regulatory Guides (1.3 and 1.4) would be r'evised by December 1986. The Subcommittee discussed the assignment of these items for review and concluded that the Nuclear Plant Chemistry Subcom-mittee would handle the SRPs and the Severe (Class 9) Accident Subcom-mittee should review the Regulatory Guides. These reviews will be performed after the draft items were issued for public comment. For the SRP review, the Subcommittee approved the review of the revisions by
r -e Severe Accidents / Nuclear Plant-Chemistry Meeting Minutes September 24, 1986 suitable consultants and resolution of issues by mail rather than in another meeting.
During the meeting, Subcommittee members and consultants expressed concerns and opinions as follows:
(1) W. Kerr indicated that this was.the first of several expected meetings on the IDCOR IPEM. He asked about the general availabil-ity of the Source Term Code Package and was informed that it was contained in NUREG/CR-4587. For each study, he inquired whether core damage frequency meant the onset of fuel damage or core on the floor. The answers on this matter were not consistent.
(2) D. Moeller inquired about the Staff's position on using new dose limits and a move toward ICRP weighting factors. He stated that the three options in the revision for SRP 6.5.2 lacked clarity and
'should be improved. He questioned whether practices at foreign reactors had been reviewed in regard to spray additives. He also, questioned why the Proceedings of the Nuclear Air Cleaning Confer-ences had not been more extensively referenced in the SRP.
(3) C. Mark requested that the relationship between the presentations and the Commission Safety Goal be established. He asked what the most prominent species of fission product releases would be if iodine is to be downgraded.
(4) D. Okrent emphasized the importance of considering external events and indicated it is a fundamental mistake to postpone their consideration. He asked about the availability of IDCOR reports, ingeneral,andtheModularAccidentAnalysisProgram(MAAP) code, specifically. On NRC/IDCOR technical issues, he indicated that the Staff is deficient to push the resolution of some of these off on NUREG-1150. He asked how the steam explosion issue was resolved.
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O 5evere Accidents / Nuclear Plant Chemistry Meeting Minutes September 24, 1986 He believes that the resolution obtained is inadequate and it should be pursued further. He recommended that the Staff provide a written approval / evaluation of the revised IDCOR IPEM before the Subcommittee and Full Comittee reviews the matter. He asked that:
IDCOR and NRC respond in writing to the current and near term comments on the IFEM provided by the ACRS consultants. He recom-mended that the IPEM be tested on a plant or plants that have not been thoroughly evaluated in prior PRA studies. He also questioned the failure modes for ice condenser plants, the " poor man's" filter for venting, and the characterization of Mark III releases as being essentially noble gases.
(5) J. Ebersole asked about the performance criteria for vent valves and felt that most would fail to perform. He indicated that a matrix should be developed for operator action during a severe accident. He suggested that requirements pertaining to microleakage also be deleted as a source term related change.
(6) D. Ward asked about the application of new meteorological models and inquired about the need for further research funds to resolve the NRC/IDCOR technical issues.
(7) P. Davis asked whether the IPE might capture a weakness that a Level 1 PRA would miss.
(8) I. Catton asked if NRR/DSR0 had reviewed the coments on SECY 86-228 by R. Bernero and what their course of action would be. He requested a copy of their planned response to Bernero.
(9) C. Siess asked about the application of the new realistic source term and the consideration of any changes to containment perfor-mance, especially leak rates. He also discussed the concepts of what constituted a large release.
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Severe Accidents / Nuclear Plant Chemistry Meeting Minutes September 24, 1986 (10) M. Corradini emphasized the need for documentation on the resolu-tion of NRC/IDCOR technical issues. He asked whether MAAP described hydrogen generation due to a steam explosion.
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NOTE: Additional meeting details can be obtained from a transcript 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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, .c ACRSSevere(Class 9)Accidentsand Nuclear Plant Chemistry Subcomittees Meeting September 24, 1986 .
Washington, DC 9 - Tentative Presentation Schedule -
Regulatory Implementation Plan for Severe Accident Policy t
W. Kerr 8:30am
. Subcomittee Chairman Remarks Introductory Remarks T. Speis, NRR 8:40am B.
R. Palla NRR 8:55am C. Resolution of IDCOR/NRC Technical Issues D. Evaluation of Reference Plants Z. Rosztoczy, NRR 9:40am (a) Evaluation Process / Individual PlantExamination(IPE)
- Lunch ***
IDCOR IPE Methodology J. Carter, IOCOR* 1:30pm E.
F. Coffman, NRR 3:00pm F. NRC Coments on IPEM 3:30 - 3:40pm
- Break ***
L. Soffer, NRR 3:40pm G. Introduction of Realistic Source Term Estimates Into Licensing H. Source Term Related Changes J. Read, NRR 4:10pm (a)PWRSprayAdditives J. Read, NRR 4:40pm I (b)BWRSuppressionPoolScrubbing l W. Kerr/D. Moeller 5:10pm I. Concluding Remarks 5:20pm
- Adjourn ***
- - IDCOR Speakers were 1 R. Henry (FAI), E. Burns (Delian)
Contact:
Dean Houston and K. Vavrek (}I) 634-3267
NRR STAFF PRESENTATION TO THE ACRS
SUBJECT:
SEVERE ACCIDENT AND SOURCE TERM IMPLEMENTATION
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DATE: SEPTEMBER 24, 1986 PRESENTER: THEMIS P. SPEIS PRESENTER'S TITLE / BRANCH /DIV: DIRECTOR -
DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.: 492-7517 SUBCOMMITTEE: CLASS 9 ACCIDENTS AND NUCLEAR PLANT CHEMIS RY' S
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9 *S IWLEE NTATION ELEFENTS (SECY-86-76)
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- 1. EXISTING PLANT EXAMINATION DISPOSITION OF THE NRC/IDCOR TECHNICAL ISSUES REVIEW 0F THE IDCOR INDIVIDUAL PLANT EXAMINATION ETHODS (IPEM)
DEVELOPENT OF GUIDELINES AND CRITERIA FOR THE
- INDIVIDUAL PLANT EXAMINATIONS EXTERNAL EVENTS (SECY-86-162)
- 2. THE ROLE OF PRAS FOR NEW PLANT APPLICATIONS
- ACCEPTABLE CONTENT OF PRAS
- CRITERIA FOR THE REGULATORY REVIEW AND INTERPRETATION OF THE PRA RESULTS
- 3. CHANGES IN RULES AND REGULATORY PRACTICE
- SOURCE TERM RELATED CHANGES SEVERE ACCIDENT RELATED CHANGES
1, NPC/IDCOR TECHNICAL ISSUES
. MAJOR DIFFERENCE IN NRC AND IDCOR MODELS IDENTIFIED DURING TECHNICAL EXCHANGE E ETINGS MAINLY RELATED TO SEVERE ACCIDBff PHENOWNOLOGY
- BELIEVED TO INFLUENCE PLANT ANALYSES CONSOLIDATED INTO 18 NRC/IDCOR ISSUES d
- 11. EVIEW 0F THE IDCOR INDIVIDUAL PLANT ETHODS (IPEM)
IPEM TO PROVIDE FOR AN INTEGRATED AND SIPPLIFIED ETHOD TO SYSTEMATICALLY EXAMINE EXISTING PLANTS FOR VAJOR VULNERABILITIES TO SEVERE ACCIDENTS SEPARATE ETHODS FOR BWRS AND PWRS TESTED BY APPLICATIONS AT 8 PLANTS
NRR STAFF PRESENTATION TO THE ACRS t
SUBJECT:
RESOLUTION OF IDCOR/NRC TECHNICAL ISSUES DATE: SEPTEMBER 24, 1986 PRESENTER: ROBERT L. PALLA, JR.
PRESENTER'S TITLE / BRANCH /DIV: MECHANICAL ENGINEER REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.: 492-4563 SUBCOMMITTEE: SEVERE (CLASS 9) ACCIDENTS SUBCOMMITTEE ND NUCLEAR. PLANT CHEMISTRY SUBCOMMITTEE
1 NRC/IDCOR TECHNICAL ISSUES o MAJOR DdFFERENCES IN NRC AND IDCOP V0DELS IDENTIFIED DUPING TECHNICAL EXCHANGE MEETINGS ,
- PAINLY RELATED TO SEVERE ACCIDENT PPENOMEN0 LOGY
- BELIEVED TO INFLUENCE PLANT ANALYSES o CONSOLIDATEDINTO18NRd/IDCORISSUES O ISSUE RESOLUTION PROCESS -- INITIATED MARCH 1985
- PATHS TO PESOLUTION IDENTIFIED, E.G.,
CODE MODIFICATIONS ADDITIONAL COMPAPISONS OF CODES WITH DATA
- STANDARD PROBLEM CALCULATIONS IDCOR EFFORTS TOWARDS ISSL'E RESOLUTION DOCUMENTED IN T85.2 JULY 1985 ,
STAFF ASSESSMENT PROVIDED IN NRR POSITION PAPEP FOR EACH ISSUE o PROCESS NEARING COMPLET10h
2 IDCOP ISSUE MONITOPS t
R. PALLA L. CHAN
- 1. IN-VESSEL FP RELEASE V. LEUNG J. HAN
- 2. PECIRCL'LATION IN-VESSEL R. BAPPETT L. CPAN/T, WALKEF
- 3. PELEASE OF CONT, PPD MATLS.
J. PEAD L. CPAN
- 4. PCS AEROSOL DEPOSITIOP '
R. PALLA J. HAN
- 5. IN-VESSEL HpGENEPATION P. PALLA P. WRIGHT
- 6. COPE SLlFP t0PFL C. ALLEP J. TELFOPD
- 7. STEAM EXPIOS10NS
- 8. DIRECT HEATING F. ELTAWILA T. LEE B. HAPDIN E. FURSON
- 9. EX-VESSEL HEAT TPAFSFF.R P. HARDIP B. BURSON Jr. EX-VESSEL FP RELEASE P. S/ MONS L. CPAN
- 31. PEVAPOP!ZATION J. PEAD J. TELFORD
- 12. C0bTAINVENT AEPOSOL DEPOSITIOP J. PEAD J. MITCHELL/D. PYATT
]3A. SL'PPPESSION POOL BYPASS J. PEAD J. MITCHE!L 13E FP REMOVAL IN ICF COND.
- 14. EMEPGENCY PESPONSE MDDELINE J. PE/S __
F. ELTAWilA J. COSTELLO ,
- 15. C0hTAIMBT PERFORMANCE T. VALKER j L6. SECONDARY CONTAINMEhT PEPFOPEANCE P. BAPPETT
! R. PALLt P. WOPTHIFGTON
- 17. H IGNITION 2 /ND BUPNING S. SAFPS W. FAPER
- 18. E0. SURVIVABILITY l
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4 ISSUE RESOLUTION .
DOES N0,T NECESSARILY MEAN
- TOTAL AGPEEMENT EXISTS WITH IDCOR APPROACH, MODELS, RESULTS OF .
NO FURTHEF RESEARCH IS REQUIRED MEANS ,
- DIFFERENCES DO NCT,FAVE A SIGNIFICANT EFFECT ON RESULTS
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- DIFFERENCES ARE SIGNIFICANT, BUT THE PHENOMENA IS SUFFICIENTLY WELL UNDEPST00D TO REACH A POSITION ON HOW TO DEAL WITH THE ISSUE IN PLANT ANALYSES
- ISSUE RESOLUTION CAN BE CONSIDERED ISSUE DISPOSITION, I.E.,
"WHAT TO DO ABOUT IT" e
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- - . , . - e r . ,,. __ _ . .-. , _. . . . _ _ . - . , - . - , , . - . . . ,
5 ISSUE RESOLUTION .
o ISSUE RESOLUTION GROUPED INTO 5 CATEGORIES
- 1. RESOLVED VIA IMPLEMENTED REVISIONS TO MAAP
, REVISIONS DESCRIBED IN T85.2 OP SUESEQUENT C0FFUNICATIONS REVISIONS PP0 DUCE PEASONABLE AGREEMENT WITH STAFF MODELS
- 2. PESOLVEE THROUGH FUPTHER STAFF FEVIEW
, . ADDITIONAL CODE / DATA COMPARISONS IMPPOVED UNDERSTANDING OF PHENOMENA NEED FOR ADDITIONAL IDCOR MODEL CHANGES NOT APPARENT
- 3. RESOLVED THROUGH FUTURE COMPLIANCE WITF STAFF POSITIO WHICH DESCRIBE:
LIMITATIONS ON USE OF CERTAIN MODELS ACCEPTABLE ASSUMPTIONS TO BE USED IN ANALYSES REVISIONS TO CERTAIN MODELS REQUIRED
- 4. RESOLVED THROUGH TREATMENT AS AN UNCEPTAINTY IN REFERENCE PLANT ANALYSIS, E.G., H2 GENERATION
- 5. RESOLVED THROUGH TREATMENT IN THE IPEM ISSUES WHICH MAY BE AFFECTED BY PLANT SPECIFIC DIFFERENCES c RESOLUT10M OF SOME ISSUES CONTINGENT UFON RESULTS OF ON RESEARCH
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ISSUE 15 - CONTAINMENT PERFORMANCE o KEYCON[ERN l
ADt00ACY OF IDCOR CHARACTERIZATION OF. CONTAINMENT FAILURE MODE LEAK-BEFORE-FAILURE MODEL VERSUS THRESHOLD MODEL (GPCSS CONTAINMENT FAILLIEE) e ASSESSMENT .
- PRESSURE CAPAPILITY CALCULATIONS CAN PLACE A LOWER
" BOUND ON CONTAINMENT FAILUPE PRESSURE P
REllABLE PREDICTIONS 0F LEAK AREA AS FUNCTION OF AND T BEYOND STATE-OF-THE-ART -- RESEARCH ONGOING o STATUS
- UNTIL ADEQUATE UNDERSTANDING OF LEAKAGE DEVELOPS THE THRESHOLD MODEL SHOULD BE USED NO CREDIT F0P P- OR T-INDUCED LEAKAGE GROSS CONTAINMENT FAILURE AT THE CAPABILITY PRESSURE LEAKAGE MODELS MAY BE ACCEPTABLE PROVIDCD THEY ARE SUESTANTIATED BY EXPERIMENTS, E.G., PENETRATION TESTS AT SNL
WHERE DO WE G0 FROM HERE E
e IPEP. RECOMMENDS PLANT-SPECIFIC ANALYSES BE PERFORF.ED U CEPTAIN CIRCUMSTANCES ,
CANDIDATE CODES -- MAAP, STCP c ADDITIONAL VCPK REQUIPED BEFORE MAI.P ACCEPTAFLE FOR THIS PURPOSE -
MODIFICATIONS TO ACCOMMODATE ISSUE RESCLUTION USAGE GUIDELINES TO ENSURE COMPLIANCE WITH STAFF POSITIONS, E.G., UNCERTAINTY STUDIES BENCHMARKING AGAINST STCP FULL DOCUMENTATION CONTROLS ON FUTURE CODE CHANGES e REBASELINING OF REFERENCE PLANTS, AND CONFIRMATION OF IPEM CONTAINPEFT EVENT TREES WITH REVISED MAAP ALSO DESIRABLE .
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i NRR STAFF PRESENTATION TO.THE ACRS
SUBJECT:
EVALUATION PROCESS INDIVIDUAL PLANT EXAMINATION DATE: SEPTEFSER 24, 1986 PRESENTER: ZOLTAN R. ROSZTOCZY PRESENTER'S TITLE / BRANCH /DIV: CHIEF .
REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW & OVERSIGHT PRESENTER'S NRC TEL. NO.: 492-8016 1
I IMPLEMENTATION STATUS ,
EXISTING PLANTS ~ ~
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6 18'0VT OF THE 19 NRC/IDCOR TECHNICAL ISSUES ARE CLOSE TO RESOLUTION, o A PRELIMINARY EVALUATION WAS COMPLETED OF THE METHODS DEVELOPED BY IDCOR FOR THE INDIVIDUAL PLANT EXAMINATIONS, o FIVE' TEST APPLICATIONS BEING REVIEW, o METHODS LOOK PROMISING, WILL NEED SOME CHANGES, 9
o PROPOSED CRITERIA TO IDENTIFY AND JUDGE VULNERABILITIES WERE DEVELOPED FOR MARK I AND MARK Il BWRs, NEW PLANT APPLICATIONS .
o GUIDANCE DOCUMENT ON THE ROLE AND MINIMUM CONTENT OF PRAs IS UNDER PREPARATION l
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O EXTERNAL EVENTS o EXTERNhlEVENTSWILLBECONSIDEREDBUTNOTONSAMESCH'EDUL O PHASE 1 - ASSESS MARGIN OF PLANTS BEYOND DESIGN BASIS TO B COMPLETED IN ABOUT 1 YEAR
- DETERMINE EXTENT OF PRESENT REVIEWS, E.G., USI A-46, CONCERNING SEVERE ACCIDENTS.
- ESTIMATE PROTECTION AFFORDED BY PRESENT REVIEWS BEYOND DESIGN BASIS.
SPECIFY EXTERNAL EVENTS TO BE INCLUDED.
- DETERMINE NEEDS BEYOND PRESENT PROGRAMS.
O PHASE 2 - DEVELOP PROGRAM FOR PLANT SPECIFIC EXAMINATION
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o WORK HAS BEEN INITIATED ON NON-SEISMIC EXTERNAL EVENTS.
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CONTENT OF GENERIC LETTERS ,
FOR INDIVIDUAL PLANT EXAMINATIONS o REQUESISUTILITIESTOCONDUCTSYSTEMATICEXAMINATIONFOR SEVERE ACCIDENT VULNERABILITIES, o SPECIFIES SCOPE OF EXAMINATION, DEFINES VULNERABILITIES, o IDENTIFIES ACCEPTABLE METHDOS FOR THE PLANT EXAMINATIONS, o SETS SCHEDULES FOR THE EXAMINATION AND UTILITY /NRC INTER-FACES DURIN'G THE EXAMINATION, o SPECIFIES DOCUMENTATION REQUIREMENTS, a
o ESTABLISHES PROCEDURES TO BE FOLLOWED WHEN VULNERABILITIES ARE FOUND, ,
o DESCRIBES NRC'S REVIEW, AUDIT AND APPROVAL OF THE EXAMINATION'S RESULTS AND PROPOSED FIXES, ,
o DESCRIBES INTERFACES WITH ONG0ING USI's AND GSI's ,
WHICH ARE RELATED TO SEVERE ACCIDENTS, ATTACHMENTS o GUIDELINES AND CRITERIA, o EVALUATION OF APPLICABLE REFERENCE PLANT, l
I o APPROVAL OF IDCOR IPEM, 1
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PREVENTION AND MITIGATION OF SEVERE ACCIDENTS IN A BWR WITH A MARK I CONTAINMENT DEPARTMENT OF NUCLEAR ENERGY l
BROOKHAVEN NATIONAL LABORATORY UPTON, NY 11973 PRESENTED AT ACRS SUBCOMMITTEE MEETING -
ON CLASS 9 ACCIDENTS SEPTEMBER 24, 1986 BROOKHAVEN NATIONAL LABORATORY l)l)l A5500ATED UNIVERSITIES, INC.(llll
l G0ALS f
- THREE BASIC OBJECTIVES WERE IDENTIFIED FOR THIS SEVERE ACCIDENT PROGRAM, WHICH SHOULD APPLY EQUALLY TO ALL PLANT TYPES, NAMELY:
- MITIGATION OF FISSION PRODUCT RELEASE
- PREVENTION OF HIGH CONSEQUENCE SEQUENCES
- REDUCTION OF CORE DAMAGE FREQUENCY l
l BROOKHAVEN NATIONAL LABORATORY l)l)l j A5500ATED UNIVERSITIES, INC.(llll
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OBJECTIVES IDENTIFY DOMINANT AND POTENTIALLY IMPORTANT CORE DAMAGE SEQUENCES ASSESS CONTAINMENT PERFORMANCE AND SOURCE TERM RESULTS FOR DOMINANT AND POTENTIALLY IMPORTANT~
SEQUENCES PROVIDE DETERMINISTIC GUIDELINES AND CRITERIA TO ENSURE THAT SEVERE ACCIDENT G0ALS CAN BE MET FOR EACH MARK I PLANT BROOKHAVEN NATIONAL LABORATORY l} g)l AS500ATED UNIVERSITIES, INC.(Illl
BWR ANALYSES: CDFS OF LEADING SEQUENCES (WITH DOMINANT DIFFERENCES) -
l l RSS ASEP IDCOR LIMERICK SHOREHAM
- PEACH PEACH PEACH PRA LIMERICK PRA ,,,SHOREHAM SEQUENCE BOTTOM BOTTOM BOTTOM REVIEW WOULD-BE REVIEW WOULD-BE
! TW 1.7-5 8 1.5-7 3.2-6 -3.2-8 9.0-6 -9 0-8 (A) (A) (A) (A)
TC 1.3-5 1.0-6 7.3-6 3.7-6 3.7-6 4.5-5 -5.0-6 (B)(C) (B)(C) (C)
TQUV & TOUX 8.4-7 6.8-8 4.1-8 6.0-5 -2.0-7 5 0-5 6.0-7 (D) (D) (D) (E)
TB 1 0-7 8.7-6 4.5-7 3.1-5 4.9-6 1.3-5 2.3-6 (F) (F) (G) (G)(H)
TPQI --- --- ---
9 0-7 e 1.7-7 c (A) (A)
TOTALS 3.1-5 9.8-6 7.9-6 9.9-5 8.8-6 1-2-4 8 0-6 (A) WETWELL VENTING ,
(B) HARDWARE IMPROVEMENTS: ARI AND 86 GPM EQUIVALENT SLCS (C) IMPROVED HUMAN RELIABILITY AND CREDIT FOR CONTROLLED LOW PRESSURE INJECTION (D) MODIFIED ADS (E) ELIMINATE NOVEL TRANSIENTS, IMPROVE HPI UNAVAILABILITY (F) RECOGNITION OF THE TB DEPENDENT FAILURES OF HPCI AND RCIC (E.G., BATTERY DEPLETION, PUMP SEALS, ROOM COOLING)
(G) LOWER INITIATOR FREQUENCY -
(H) RECOVERY PROBABILITIES BROOKHAVEN NATIONAL LABORATORY l} gj l A5500ATED UNIVERSITIES, INC(1til
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RISK ASSESSMENT RESULTS FOR BWR-4 MARK I PLANTS
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CDF i 10-5/RX-YR DOMINANT CONTRIBUTORS _
STATION BLACK 0UT ATWS POTENTIAL CONTRIBUTORS TW TQUX INTERFACING SYSTEM LOCA SUPPORT SYSTEM INTERDEPENDENCIES NEED SPECIFIC GUIDELINES TO ENSURE ,
ACCEPTABLE CDF GUIDELINE 5 MITIGATION OF STATION BLACK 0UT GUIDELINE 6 MITIGATION OF LOSS OF CONTAINMENT HEAT REMOVAL SUIDELINE 7 REACTOR PRESSURE VESSEL DEPRESSURIZATION GUIDELINE 8 SUPPORT SYSTEM INTERDEPENDENCIES BROOKHAVEN NATIONAL LABORATORY l} l)l A5500ATED UNIVERSITIES, INC.(llll
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MARK I CONTAINMENT VULNERABILITIES SMALL CONTAINMENT VOLUME (POTENTIAL FOR EARLY CONTAINMENT FAILURE)
- NONCONDENSIBLE GASES
- HIGH TEMPERATURES
- DEBRIS MELT-THROUGH
- REVAPORIZATION OF IN-VESSEL FISSION PRODUCTS DRY REACTOR CAVITY AND HIGH ZIRCONIUM CONTENT
- HIGH TEtiPERATURE DEBRIS
- AGGRESSIVE CORE CONCRETE ATTACK
- NON-VOLATILE FISSION PRODUCT RELEASE VACUUM BREAKERS
- P0OL BYPASS BROOKHAVEN NATIONAL LABORATORY l} gj l A5500ATED UNIVERSITIES, INC.(IIll
o .
t MAR'K I CONTAINMENT MITIGATING FEATURES
- SUPPRESSION POOL
- CONDENSES STEAM
- SCRUBS FISSION PRODUCTS
- DRYWELL INERTING
- PREVENTS H 2 BURN
- DRYWELL SPRAYS
- REDUCES TEMPERATURE
- REDUCES PRESSURE
- ENHANCES FISSION PRODUCT AEROSOL DEPOSITION
- COOLS DEBRIS
- WETWELL VENTS
- ALLOWS DEPRESSURIZATION
- MAINTAINS POOL SCRUBBING
- REACTOR BUILDING
- ENHANCES AEROSOL DEPOSITION BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(Illl
I GUIDELINE 1 A. PROVIDE WETWELL VENTING BASIS:. IMPLEMENTATION WILL SIGNIFICANTLY REDUCE POTENTIAL l FOR LOSS OF CONTAINMENT INTEGRITY DUE TO OVERPRESSURIZATION EVENTS CRITERIA:
VENTING COMMENCE (EXCEPT FOR STATION BLACK 0UT)
WHEN CONTAINMENT PRESSURE (CP) IS AT LOWEST OF FOLLOWING SET POINTS:
A. 75% OF ULTIMATE CAPABILITY, OR B.
SAFETY RELIEF VALVES PREVENTED FROM PERFORMING THEIR FUNCTION, OR C. VENT VALVES PREVENTED FROM PERFORMING THEIR FUNCTION
, DURING STATION BLACK 0UT, WETWELL VENTING COMMENCES THREE HOURS FOLLOWING ONSET OF TRANSIENT, BUT BEFORE DEPLETION 0F STATION BATTERIES.
- IF MANUAL INITIATION REQUIRED, TRAINING AND EMERGENCY PROCEDURES (TsEP) NEEDED TO ENSURE PERSONNEL ARE NOT EXPOSED TO HARSH ENVIRONMENT. OTHERWISE, VALVES POWERED FROM AN INDEPENDENT SOURCE.
BROOKHAVEN NATIONAL LABORATORY l} g)l
- A5500ATED UNIVERSITIES, INC.(1lll
. . _ _ . . - - . , . _ . - . ._ _ ~ , . . . . - . .-_ _ . - - _ . _ - _ .
6UIDELINE 1.A. PROVIDE WETWELL VENTING (CONTINUED)
VENTING TERMINATED BEFORE CP DECREASES TO 15 PSIG TREP SPECIFY:
A. PLANT PARAMETERS TO PROMPT OPERATORS TO MAKE PREPARATION, COMMENCE AND TERMINATE VENTING.
B. FLOW PATH (S) AVAILABLE, SPECIFIC COMPONENTS ALIGNED, AND REQUIRED POSITIONS / STATES FOR THESE COMPONENTS.
C. HOW TO PROCEED IF TERMINATION IS NOT POSSIBLE 4
VENTING CAPACITY GREATER THAN PREDICTED RATE OF INCREASE OF CONTAINMENT PRESSURE (CP).
RADIOLOGICAL RELEASE REDUCED BY ORDER OF MAGNITUDE COMPARED TO NO FILTERING. FLOW PATH SO ALL MATERIALS TO BE VENTED PASS THROUGH SUPPRESSION POOL (SP). IF BYPASS OF SP, ACTIVE FILTERING (SUCH AS DRYWELL SPRAYING) TO BE PROVIDED.
EQUIPMENT NEEDED TO SUPPORT VENTING FOR 211 HOURS AFTER VENTING INITIATION UNDER PREDICTED ENVIRONMENTAL AND FLUID LOADS.
BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC. (Illl
l GUIDELINE 1.B. PREVENT DEINERTING BASIS: DURING WETWELL VENTING AND DRYWELL SPRAY, POSSIBILITY EXISTS TO CAUSE VACUUM AND DRAWN NORMAL REACTOR BUILDING AIR INTO CONTAINMENT. THIS COULD FORM COMBUSTIBLE MIXTURE IN CONTAINMENT CRITERIA:
OPERATOR T8EP SPECIFY WETWELL VENTING BE TERMINATED BEFORE CP DECREASES TO 15 PSIG T8EP SPECIFY A. METHODS FOR MONITORING AND CONTROLLING CP AND OXYGEN CONTENT ,
B. INSTRUMENTATION AND CONDITIONS FOR OPERATOR ACTION BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(Illl
GUIDELINE 2.A. PREVENT SUPPRESSION POOL BYPASS BASIS: IMPLEMENTATION WILL SIGNIFICANTLY REDUCE POTENTIAL FOR BYPASS CRITERIA:
ENSURE THAT CORE DEBRIS (CD) LEAVING THE RPV DOES NOT COME INTO CONTACT WITH STEEL CONTAINMENT SHELL
- CD CONTROL ACCOMPLISHED FOR LOW PRESSURE SEQUENCES
- CD CONTROL METHOD CAPABLE OF CONFINING CORE, RPV BOTTOM HEAD AND CORE SUPPORT STRUCTURE l
GUIDANCE: SUITABLE METHOD-CONCRETE OR MAGNESIUM OXIDE CURBING TO PREVENT DEBRIS FROM REACHING CONTAINMENT SHELL PROCEDURES AND TRAINING SPECIFY ACTIONS TO ENSURE CONTAINMENT ISOLATION VALVES AND VACUUM BREAKERS CAPABLE OF CLOSING (UPON INITIATION SIGNAL) AND REMAINING CLOSED WITH LEAK TIGHT INTEGRITY REACTOR COOLANT SYSTEM LEAKAGE DIRECTED THROUGH THE SUPPRESSION l POOL UNLESS FILTERING PROVIDED 1
i BROOKHAVEN NATIONAL LABORATORY l} g)l l A5500ATED UNIVERSITIES, INC.(Illl
s GUIDELINE 2.B. PROVIDE DRYWELL SPRAY BASIS: IMPLEMENTATION WILL AID IN FISSION PRODUCT DRYWELL ATMOSPHERE DECONTAMINATION AND WILL HELP CONTROL CP RISE l
CRITERIA:
SPRAY COMMENCES WHEN CP REACHES DESIGN PRESSURE OR BEFORE DRYWELL TEMPERATURE REACHES ADS QUALIFIED VALUE.
'~
SPRAY TERMINATED WHEN CP DECREASES TO 5 TO 10 PSIG.
PROVIDED BACKUP SPRAY SYSTEM WITH DIVERSE WATER SUPPLY AND INDEPENDENT POWER SOURCE.
GUIDANCE: DIESEL-DRIVEN FIRE PUMP WITH A SUFFICIENT FLOW HEAD COULD BE SUITABLE ,
EQUIPMENT FOR SPRhY CAPABLE OF PERFORMING UNDER PREDICTED CONTAINMENT CONDITIONS WITH SUFFICIENT CAPACITY AND HEAD FOR 10 HOURS.
1 i
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(1lll
-+
GUIDELINE 2-B. PROVIDE DRYWELL SPRAY (CONTINUED)
OPERATOR TREP SPECIFY FLOW PATHS AND SPECIFIC COMPONENTS TO BE ALIGNED INCLUDING ANY TEMPORARY SYSTEM CROSS CONNECTIONS IF BACKUP SYSTEM USED OPERATOR TREP SPECIFY PLANT PARAMETERS TO PROMPT OPERATORS TO INITIATE AND TERMINATE SPRAY CONSISTENT WITH TIME REQUIRED TO ALIGN SYSTEM AND COMP 0NENTS SPRAY HEAT REMOVAL RATE PROVIDED BY COMPONENTS SUFFICIENT TO REMOVE 1% DECAY HEAT l
l l
l BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llll
NRR STAFF PRESENTATION TO THE ACRS
SUBJECT:
MARK I REFERENCE PLANT EVALUA 10N DATE: SEPTEPBER 24, 1986 f
PRESENTER: FAROUK ELTAWILA PRESENTER'S TITLE / BRANCH /DIV: SR. REACTOR SYSTEMS ENGINEER l REGULATORY IMPROVEMENTS BRANCH l DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.: 49-24570 SUBCOMMITTEE: CLASS 9 I
t CONTAINME_NT PERFORMANCE
- ESTABLISH THE ULTIMATE PRESSURE CAPABILITY OF THE CONTAINMENT STRUCTURE WHEN SUBJECTED TO THE ACCIDENT
. CONDITIONS
- IDEN,TIFY THE AREAS OF THE CONTAINMENT SYSTEM WITH THE HIGHEST POTENTIAL FOR FAILING
- THE PEACH BOTTOM CONTAINMENT IS CONSTRUCTED OF GA 516 GRADE 70 STEEL, A CONCRETE SHIELD BUILDING ENCLOSES THE STEEL CONTAINMENT. A NOMINAL GAP OF 1-3/4 INCHES SEPARATES THE OUTSIDE SURFACE OF THE ,
CONTAINMENT SHELL FROM THE INNERMOST SURFACE OF THE SHIELD BUILDING.
- THE CONTAINMENT DESIGN PRESSURE IS 55 PSIG.
- IDCOR AND SARRP ASSUMED THE ULTIMATE PRESSURE CAPABILITY OF PEACH BOTTOM TO BE 117 PSIG BASED ON ANALYSIS OF THE BROWNS FERRY CONTAINMENT.
l __ _ . _ _ . _ __ _ _
i
_5_ 1 c
- THE 3 CRITERIA FOR FAILURE IS 2%
STRAIN. MORE REALISTIC STRAIN BASED FAILURE CRITERIA IS NEEDED.
- SUFFICIENT ATTENTION WAS NOT GIVEN TO THIS IMPORTANT ISSUE.
CHANGES IN CONTAINMENT FAILURE PRESSURE MIGHT INCREASE SUCCESS OF CERTAIN OPERATOR ACTION THAT COULD LEAD TO SUBSTANTIAL REDUCTION IN RISK DEPENDING ON THE S E Q'UENCE.
t
- CBI HAS INVESTIGATED FAILURE MODES OF THE PEACH BOTTOM CONTAINMENT AND CONCLUDED THAT UNDER THE HYPOTHESIZED TEMPERATURE CONDITIONS THE CONTAINMENT WILL NOT FAIL CATASTROPHICALLY.
- ADDITIONAL INVESTIGATION ARE BEING PERFORMED BY CBI TO ESTABLISH THE ULTIMATE PRESSURE CAPABILITY OF THE CONTAINMENT WHEN SUBJECTED TO ACCIDENT CONDITION.
1 l .
_a_ l i
CONT'AINMENT VENTING
- IDENTIFIED BY IDCOR AS OPERATOR ACTION THAT WOULD PREVENT AN OVER PRESSURE FAILURE OF THE PEACH BOTTOM, CONTAINMENT AND THUS PRESERVE SUPPRESSION POOL SCRUBBING FUNCTION
- INEL' PERFORMED ANALYSES FOR AN ATWS SEQUENCE AND FOR A STATION BLACKOUT SEQUENCE AND CONCLUDED THAT THE LIKELIHOOD OF SUCCESSFUL CONTAINMENT VENTING IS 2.6 FOR ATWS' AND ZERO FOR STATION BLACKOUT
- THE STAFF BELIEVES THAT CONTAINMENT VENTING WILL MITIGATE THE CONSEQUENCES OF A-BROAD SPECTRUM OF ACCIDENT
- SEQUENCES.
- SOME MODIFICATION TO THE PLANT OPERATING PROCEDURES MIGHT BE NEEDED TO IMPROVE THE LIKELIHOOD OF SUCCESSFUL VENTING.
i l
l--.--
l l
l
~7-t l
CONTAINMENT FAILURE MOQES
- IDCOR ANALYZED FOUR ACCIDENT SEQUENCES TO DETERMINE THE PEACH BOTTOM CONTAINMENT RESPONSE TO SEVERE ACCIDENT AND. CONCLUDED THAT THE MAIN FAILURE IS DUE TO
~
OVER TEMPERATURE.
- SARRP ANALYSTS HAVE POSTULATED SEVERAL CONTAINMENT FAILURE LOCATIONS y THE DRYWELL HEAD, THE DRYWELL KNUCKLE BETWEEN THE CYLINDRICAL AND THE SPHERICAL SECTIONS, AND ,THE WETWELL.
INFORMATION ARE PRESENTED TO A GROUP OF EXPERTS IN ORDER TO OBTAIN EXPERT OPINION ON THE POSSIBLE PHENOMENON AND THE RANGE OF OUTCOMES THAT MIGHT BE ASSOCIATED WITH THESE ISSUES. ,
- THE EXPERT PANEL WAS ASKED TO EXPRESS THEIR VIEWS ON THE POTENTIAL FOR MOLTEN CORE MATERIAL FLOWS ON THE DRYWELL FLOOR AND ATTACK THE STEEL CONTAINMENT SHELL AND FAILING IT BY MELT-THROUGH.
-e--,, _
l
/ \
l I
t IPE METHODOL OG Y OBJECTIVES Robert E. Henry Fauske & Associates, Inc.
Presented To:
Advisory Committee on Reactor Safeguards September 24,1986 .
Washington, D.C.
i
l t l
\
FUNDAMENTAL IDCOR CONCLUSIONS 1 1
a e Severe accidents have a very low probability of occurrence.
4
)
e Given a severe accident, the releases to the environment are much less than those stated
- in WA SH- 1400.
l .
e The risks and consequences are much small_er than those incorporated in the NRC Interim safety goals.
e Major design or operational changes are not.
warranted.
l k
1 3
, , , , , , , . , . . - . . . . - - - . . , ,_---.,,,,,_--.,-,----.n. . , - . , _ . . , - - . . - , , - - . - - - ~ - - , , - - , - - - , - - - - - - , - -
l 1
SCOPE OF THE METHODOLOGY e Addresses Internal Events e Not a PRA, But the Systems Analysis Can be Expanded to a Level 1 PRA e System Models are Assembled at a High Level e includes Critical Safety Functions and Front Line Systems Which Fulfill These Functions e Front Line Sy. stems include Both Safety i and Non-Safety Related Systems e Support Systems and Their Dependencies are Explicitly Addressed .
- AC Power
- DC Power k
7
IPE SOURCE TERM METHODOLOGY l
! PWR j Ice Condensers - Esssentially the Same as Sequoyah Large Dry - Independence of Primary System and Containment and Previous MAAP Analyses Allow a Simpilfled Modeling Approach. Key Elements .Are:
e Time for Debris Dryout in Containment e Time of Containment Failure e inert Aerosol Generation Rate e Aerosol Settling Rates e Tellerium Release
~
e Molybdenum Release 9
IPE SOURCE TERM METHODOLOGY
,! BWR Mark I & il Containments Dependence Between Primary System and Containment Requires the Use of Containment Event Trees for Evaluating Releases Key Elements:
e Containment Failure / Vessel Failure e Wetwell Venting e Drywell Sprays e Debris Transport to the Suppression Pool Mark 111 Containment Releases for All Accident Sequences are Essentially Noble Gasec I
- . . - - - - - , - - - -- _,n_mn- , , - , , --.--n- , - - - - . , - - -, -
l l
IDCOR l c i
, l 0
BWR INDIVIDUAL PLANT EVALUATION METHODOLOGY
SUMMARY
OVERVIEW E.T. BURNS DELIAN CORPORATION SEPTEMBER 24, 1986 PRESENTATION TO i
! ACRS
"- WASHINGTON, D.C..
~
I 2-T-123-033
SCOPE o "lNTERNAL" EVENTS:
- TRANSIENTS, ATWS, LOCAs, RARE INITI ATORS INTERNAL FLOODS INTERFACING LOCA COMMON MODE FAILURES
- SUPPORT SYSTEM DEPENDENCIES AND INITIATORS e APPROXIMATE METHOD CAPABLE OF EXPANSION TO LEVEL 1 PRA e DETAILED EVENT TREES e ALL GE BWRs e SELECT SUPPORT SYSTEMS ARE EXPLICITLY ADDRESSED:
- ROOM COOLING SERVICE WATER AC POWER
- DC POWER INSTRUMENT AIR /N2 ,
i AND PROVISION IS MADE FOR ADDITIONAL SUPPORT SYSTEM e DEPENDENCIES ADDRESSED
- FUNCTIONAL l - HUMAN INTERSYSTEM e PLANT WALKDOWN e SYSTEM NOTEBOOKS i
e OPERATING EXPERIENCE DATA 2-T-123-011
i PROCESS INCLUDES e SYSTEM NOTEBOOKS: REFERENCE SOURCE OF INFORMATION ABOUT THE PLANT I
EVENT TREES: FRAMEWORK FOR IDENTIFYING AND EVALUATING l e ACCIDENT SEQUENCES I
FOCAL POINT FOR PLANT SPECIFIC DESIGN, e FAULT TREES:
I OP RAT 0 MAINTENANCE, AND TEST w
e SUPPORT SYSTEM DEPENDENCY MATRICES: IDENTIFY AND MODEL IMPORTANT INTERACTIONS QUANTIFY THE MODELS (GENERIC DATA, AND e AVAILABLE DATA:
"' REFERENCE PLANT DATA IS PROVIDED IF PLANT l
SPECIFIC DATA IS NOT AVAILABLE) e ENGINEERING INSIGHTS: PROBE FOR PLANT UNIQUE FEATURES OR
' POTENTIALLY VULNERABLE AREAS 2-T-123-033 l_ --- --- - - ____
- CORE ET INITIATORS '
7 EVB4T TREES FREQUB4CY AL FUNCTI0tML LEVEL FAULT TEES J k l 6
! ENGIEERING l INSIGHTS 1
I I
ELEENTS OF If0lVIDUAL PLANT EVALlMTION ETFDD R
- ACCIDENT SEQUENCE EVALUATION METHOD ALTERNATIVES FOR IMPLEMENTATION PRINCIPAL MANPOWER CHARACTERISTIC (MM)
TYPE EXAMINE DOMINANT SEQUENCES 6 HIGH LEVEL PREVIOUSLY IDENTIFIED EXAMINE ADDITIONAL DOMINANT 12 ENGINEERING LEVEL SEQUENCES EXAMINE ALL IDENTIFIED 18 DETAILED ENGINEERING SEQUENCES IN METHOD LEVEL DETAILED FAULT TREE MODELS84-144 PRA: LEVEL 1 AND PLANT SPECIFIC EVENT TREE DEVELOPMENT
{
f 2-T-60-002
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SUMMARY
OF ,
THE METHOD 1
e DEVELOPED TO CALCULATE A REALISTIC PLANT SPECIFIC CORE MELT. FREQUENCY e BASED ON INSIGHTS FROM PAST PRAs AND IDCOR e IS NOT A PRA e IS USABLE FOR COMMUNICATION TO MANAGEMENT-l e IDENTIFIES POTENTIAL OUTLIERS e EXPANDABLE TO A LEVEL 1 PRA e IS EASILY UPDATED IF INFORMATION BECOMES AVAILABLE IN THE FUTURE, E.G. PLANT SPECIFIC DATA e
I l
I I
i i
ll 2'==
I J
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/- .:.* y E
1 IDCOR T1ETHODOLOGY FOR PERFORMING PWR INDIVIDUAL PLANT EVALUATIONS l
i .
. I i
1 a Kermeth J Vevrek Sene Eng m '
vYAaYe eo Tectwoogy Group ectre Corporate Westegs P[Bos355 tEY2bf4 71
, WIN 275 71$$
i
_ _ _ _ . - - . _ _ , . . , _ _ _ _ . - _ . _ _ _ _ _ _ _ _ , , , _ - _ _ , . _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ , , , _ _ . _ . . . _ _ . . , . _ _ _ - _ _ _ . . . , , . . . , , , ,m. _ . , . _ . , . . _ _ .
THE METHODOLOGY APPROACH INCLUDES THE FOLLOWING COMPONENTS:
O INITIATING EVENTS 0 EVENT TREES / ACCIDENT SEQUENCES 0 SUCCESS CRITERIA 0 SUPPORT SYSTEMS 0 FAULT TREES / SYSTEMS ANALYSIS l 0 FAILURE DATA
REQUIRED INITIATING EVENT ANALYSIS Initiatina Event. Frecuency/ Reactor Year Generic Plant Specific Large LOCA 1E-4 1E-4 Medium LOCA 3E-4 3E-4 Small LOCA 7E-3 7E-3
(
Interfacing Systems LOCA Section B.1.2 Type Calculation (
Transient with Main Feedwater Available 7.4 Table 8.1-1 Transien't with Main feedwater Unavailable 1.3 Table B.1-2 Loss of Offsite Power NSAC 80 m..-.-= .i
o .
TYPICAL INITIATING EVENT ASSESSMENTS t
Initiatine Event Freauenev/ Reactor Year Loss of CCW 3nd/or Service Water Plant Specific Analysis Required Loss of DC Vital Bus Plant Specific Analysis Required Major Feedline Rupture 4E-4 Major Steamline Ruptures 4E-4 Loss of AC Vital Bus Plant Specific Analysis Required Steam Generator Tube Rupture Plant Specific Analysis Required Reactor Pressure Vessel Rupture 3E-7 i
l l
I
l IDENTIFICATION OF EVENT TREE SYSTEMS Safety Function Plant Specific Systems ReactorTripI Reactor Protection System High Pressure Safety Injection Chemical Volume and Control System i
RCS Inventory Control l
High Pressure Safety Injection i
Low Pressure Safety Injection High Pressure Recirculation Low Pressure Recirculation Residual Heat Removal System Chemical Volume and Control System Accumulators / Core Flood Tanks
. Decay Heat Removal Residual Heat Removal System i Chemical Volume and Control System Accumulators / Core Flood Tanks Power Operated Relief Valves i Condensate System Auxiliary / Emergency Feedwater Nain Steam System i (Alternate feedwater supply systems such asservicewater, firewater,etc.)
Containment Cooling Containment Fan Coolers Containment Spray Quench Spray System
. Recirculation Spray System l Ice Condensers Radioactivity Scrubbing Containment Fan Coolers Containment Spray Quench Spray System Recirculation Spray System
- Ice Condensers j
- p.
- - - - - - - - . - - - - , , _ - - -. - , . - - -- -- -- ~ .,- - - - --- -.--.. - . - - -
,' TABLE 2.2-8 t SYSTEM DEPENDENCE NATRIX 1
v 3 g W W is - *3 . . u m - ". ". *2 I 5
B 6 6
- b. 2 . . .
F m EA i
Example Matrix j 5 gggjig (ReadAcross) ] ]
', ', a hT f I*g I. I. I. _5 I& m 3 a e w 32 E Engineered Safety Features X X X X Integrated Control System X X X X Electric Power - Offsite AC 1 Electric Power - Onsite AC X X X X
( Electric Power - Onsite DC Component Cooling Water X X X
X X X X
Service Water System X X X Instrument Air System X X HVAC System X X X Auxiliary / Emergency Feedwater X X X X X High Pressure Safety Injection X X X X X X Low Pressure Safety Injection X X X X X X High Pressure Recirculation X X X X X X Low Pressure Recirculation X X X X X X Chemical Volume & Control System X X X X X Main Feedwater System X X X X X Main Steam System X X Condensate System X X X X Accumulators / Core Flood Tanks Containment Fan / Coolers X X X X X X Containment Spray System X X X X X X Recirculation Spray System X X X X X Ouench Spray System X X X X X X Ice Condenser. System X X X Turbine Bypass System X X Reactor Protection System X X X X PZR Power Operated Relief Valves X
"""8'" 2-108
- - x; : .- - af APPROXIMATE SOURCE TERM METHODOLOGY FOR BOILING WATER REACTORS Robert E. Henry Fauske & Associates, Inc.
Presented To:
Advisory Committee on Reactor Safeguards September 24,1986 -
I Washington, D.C.
t
l . .
IPE SOURCE TERM METHODOLOGY l BWR Mark I & il Containments Dependence Between Primary System and Containment Requires the Use of Containment Event Trees for Evaluating Releases Key Elements:
e Containment Failure / Vessel Failure e Wetwell Venting e Drywell Sprays l
e Debris Transport to the Suppression Pool Mark Ill Containment Releases for All Accident Sequences are Essentially Noble Gases
- (
3
7&: e s,-:.f APPROX'I MATE SOURCE TERM METHODOLOG Y FOR PRESSURIZED WATER REA CTORS Marc A. Kenton Fauske & Associates, Inc.
Presented To: -
Advisory Committee on Reactor Safeguards September 24, 1986 l
Washington, D.C.
l
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r" 'N g
I ,
IPE PWR SOURCE TERM METHODOLOGY IS A SCREENING PROCEDURE
- 1. Ice condenser plants are essentially identical from a severe accident perspective.
- 2. Large, dry containments differ but most will have:
- a. Long times to containment failure (unless containment bypassed or impaired),
- b. No primary system revolatilization,
- 3. Large, dry containments which are exceptional are identified via checklists and hand calculations. More detailed analysis is recommended for such plants, a
Simple hand-calculational procedures suffice for unexceptional large, drys.
l l
. . _ _ _ _ - B_ .. . _ _ _ . _ _ . _ _ __
[
LARGE DRY SEQUENCES CAN BE GROUPED INTO FIVE CATEGORIES
- 1. V-sequences (bypassed containment).
- 2. Failure-to-isolate sequences (impaired containment),
- 3. Cases with intact containments and AC power.
I4. Blackouts (and phenomenologically similar sequences with no water injection to containment).
~
- 5. Others - certain sequences have usually not been found to be risk-significant in the past and are more appropriately addressed on a generic basis since tney donotinvolveindividualplantvulnerabilities(e.g.
i tube ruptures),
3
~
GENERAL CHARACTERIZATION OF PWR SEVERE ACCIDENT SEQUENCES Blackout with Intact
! Isolation V With AC Blackout Failure Sequence Power Probability of 6 x 10~0 3 x 10-8 1 x 10~7 s 10-5 Sequence per Reactor Year * -
Debris Remains No No No Yes Water-Covered Throughout Containment Unsaturated Unsaturated Unsaturated Saturated
! Humidity Time to Containment Long 0 N/A Varies, i
Failure Usually Long Auxiliary Building Depends on Usually Yes Yes Usually No Characteristics Containment Important? Failure Mode; Conservative to Neglect .
- From the Zion Probabilistic Safety Study (14] and the IDCOR Technical Sumary Report [15].
[
l j
SOURCE TERM METHODOLOGY
- 1. Sufficient to focus on station blackout (and phenomenologically-similar sequences) and V-sequences unless:
- a. Systems analysis reveals unusually high probabilities for failure-to-isolate sequences with no containment heat removal,
- b. Certain plant geometry criteria are not met,
- c. Debris-covered sequences have very early containment failure times,
- 2. V-sequence is not an issue for most plants which will have high retention in auxiliary building - detect outliers with checklist. .
3- A mechanistic hand-calculational procedure is provided for large, dry containments to compute blackout source terms.
- 4. A mechanistic hand-calculational procedure is provided to verify that source terms from debris-covered sequences are 10w.
- 5. Ice condenser designs are very similar and the TMLB l
releases are bounded by Sequoyah.
k s
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i ~
NRR STAFF PRESENTATION TO THE !
ACRS -
SUBJECT:
NRC COMMENTS ON THE IDCOR INDIVIDUAL PLANT EXAMINATION METHODS DATE: SEPTEMBER 24, 1986 PRESENTER: FRANKLIN C0FFMAN PRESENTER'S TITLE / BRANCH /DIV: SECTION A LEADER REGULATORY IMPROVEMENTS BRANC'H DIVISION OF SAFETY REVIEW & OVERSIGHT PRESENTER'S NRC TEL. NO.: 492-4609 SUBCOMMITTEE: CLASS 9 ACCIDENTS AND NUCLEAR PLANT CHEMISTRY S
STEPS IN THE PRELIMINARY EVALUATION OF THE IDCOR IPEM I. STANDARDS FOR THE REVIEW:
- 1. CAfABILITYTOFINDVULNERABILITIES ,
THE
- 2. APPLICABILITY BETWEEN THE EXISTING PLANT AND REFERENCE PLANT
- 3. ROLE OF VISUAL INSPECTIONS
- 4. COVERAGE OF CURRENT INSIGHTS
- 5. " SYSTEMATIC EXAMINATIONS"
- 6. CONSISTENCY WITH GENERIC RESOLUTIONS 7, COVERAGE OF THE GUIDELINES AND CRITERIA
- 8. DOCUMENTATION AND PRESENTATION
- 9. LIMITATIONS AND CAUTIONS II. THE COMMENT SHOULD BE RELEVANT TO AT LEAST ONE CRITERION III. THE COMMENT SHOULD BE TRANSLATED INTO A C0HERENT ACTION POSITION IV. THE COMMENT SHOULD HAVE CLEAR DISPOSITION FOR THE PREL.IM EVALUATION 2
I
PRELIMINARY COMMENTS ON THE IDCOR IPEM
- 1. PROPOSED GROUNDRULES -
CORE DAMAGE FREQUENCY PROPOSED GROUNDRULES SHOULD BE CONSISTENT WITH THE SAFETY G0ALS .
- UNCERTAINTY SHOULD BE QUANTIFIED FOR THE IPE'S -
-l THE USE OF UNCERTAINTY SHOULD BE SPECIFIED
~
RELEASE FROM CONTAINMENT PROPOSED FORM SHOULD BE MORE PRACTICAL
- 2. INDIVIDUAL PLANTS ARE MATCHED AGAINST THE REFERENCE PLANT, BUT THERE NEEDS TO BE A DEFINITION OF WHAT CONSTITUTES A MATCH,
- 3. ADEQUATE CHARACTERIZATION OF SUCCESSFUL VENTING NEEDED
- 4. VULNERABILITY NEEDS TO BE DEFINED IDENTIFY POTENTIAL VULNERABILITIES FOUND IN REFERENCE PLANTS IDENTIFY POTENTIAL VULNERABILITY BASED ON DEVIATION FROM REFERENCE PLANT (SHOULD BE DONE AT EACH LEVEL OF SYSTEMS MATCHING) 5, IDCOR IPEM DOES NOT CONSIDER OPERABILITY OF ESSENTIAL EQUIPMENT
- 6. SIMPLIFIED TREATMENT OF PHENOMEN0 LOGICAL ISSUES WITHO EXPLICIT TREATMENT OF UNCERTAINTIES .
- 7. POTENTIAL BENEFITS OF VISUAL INSPECTIONS WERE ACKNOWLEDGED, HOWEVER VISUAL INSPECTION PROCEDURES ARE NEEDED
- 8. IDCOR IPEM SPECIFIES SOME DOCUMENTATION, HOWEVER DOCU-MENTATION SHOULD BE CALLED FOR WHERE IT IS NEEDED TO ASSURE TECHNICAL QUALITY i
3
SCHEDULE FOR THE
' IDCOR IPEM EVALVATION .
THE NEXT STEPS BEGIN AT 9
- 1. NRC/IDCOR MEETING ON REVIEW STANDARDS - APRIL 3, 1986
~
- 2. RECEIP.I0FTHEIDCORIPEMREPORTS-MAY2,1986
- 3. lDCOR WORKSHOP ON THEIR IPEM - MAY 14 8 15, 1986
- 4. C0 ORDINATION MEETING FOR THE REVIEWS - MAY 19, 1986
- 5. MEETING TO CONSOLIDATE REVIEW COMMENTS - JUNE 19, 1986
- 6. INDIVIDUAL REVIEWER'S PRELIMINARY EVALUATIONS - JUNE 25, 1986
- 7. RECEIPT OF 5 IDCOR IPEM APPLICATIONS REPORTS - JUNE 26, 1986
- 8. PRELIMINARY EVALUATION TO IDCOR - SEPTEMBER 9, 1986
- 9. RECEIPT OF CALVERT CLIFFS APPLICATIONS REPORT - SEPTEMBE
- 10. RECEIPT OF IDCOR RESPONSES T0 IPEM COMMENTS - OCTOBER 31, 1986
- 11. RECEIPT OF SEQUOYAH APPLICATIONS REPORT - MID NOVEMBER 19
- 12. RECEIPT OF MODIFIED IDCOR IPEM - DECEMBER 1, 1986
- 13. FINAL EVALUATION REPORT - DECEMBER 19, 1986 .
- 14. RECEIPT OF GRAND GULF APPLICATIONS REPORT - DECEMBER 1986 4
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NRR STAFF PRESENTATION TO THE ACRS
SUBJECT:
INTRODUCTION OF, REALISTIC SOURCE TERM ESTIMATES INTO LICENSING DATE: SEPTEMBER 24, 1986 PRESENTER: LEONARD SOFFER PRESENTER'S TITLE / BRANCH /DIV: SECTION LEADER .
REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.: 49-27976 SUBCOMMITTEE: SEVERE (CLASS 9) ACCIDENTS SUBCOMMITTEE AND NUCLEAR PLANT CHEMISTRY SUBCOMMITTEE
INTRODUCTION OF REAllSTIC SOURCE TERMS c
3 INTOLICENSING (SECY 86-228)
STAFF WILL REPLACE TID-14844 WITH REALISTIC SOURCE TERM
- INTENDED ONLY FOR FlfiURE PLANTS, OPTIONAL FOR EXISTING PLAKfS SOURCE TERM CODE PACKAGE FROM NUREG-0956 TO BE USED FOR COMPLITATION OF RELEASES FROM A GROUP 0F SEVERE ACCIDENTS REPLACE TID-14844 WITH TABLE (S) SPECIFYING OVANTITY AND RATES OF RELEASE INTO CONTAINEKi FOR GROUP 0F ACCIDENTS.
. AND DIFFERING CONTAlttENT 1YPES
- REMOVAL PROCESSES, ESF PERF0,7ANCE AND LEAKAGE WILL BE
- COMPUTED REALISTICALLY ENVIR0ffENTAL TRANSPORT AND D0SIETRY TO USE MOST N0DERN MODELS ll EXPLICIT MARGIN OF SAFETY TO BE APPLIED l
SOURCE TERM RELATED CHANGES _ ,
6 DRAFT, REVISED SRPs WERE PREPARED FOR THE TWO SHORT-TERM CHANGES.
o PROPOSED POSITION:
- SPRAY ADDITIVES (PWR) ARE NOT NEEDED DURING INJECTION, BUT SUMP PH CONTROL REQUIREMENT
' REMAINS,
- CREDIT WILL BE GIVEN FOR SUPPRESSION POOL FISSION PRODUCT CLEANUP (BWR) TO THE EXTENT JUSTIFIED BY EXISTING TESTS AND CALCULATIONS, o CHANGES WILL BE DISCUSSED WITH CRGR SHORTLY.
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