ML20195B330

Summary of ACRS Subcommittee on Class 9 Accidents 860224 Meeting in Washington,Dc to Review NRR Implementation Plan for Severe Accident Policy Statement & Regulatory Use of New Source Term Info.Viewgraphs Encl

ML20195B330
Person / Time
Issue date:	03/10/1986
From:	Advisory Committee on Reactor Safeguards
To:	Advisory Committee on Reactor Safeguards
References
ACRS-2400, NUDOCS 8605290218
Download: ML20195B330 (55)

Text

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DATE ISSUED: 3/10/86 3lg,lgg ACRS Subcommittee Meeting Summary / Minutes For Class 9 Accidents February 24, 1986 Washington, DC Purpose The ACRS Subcommittee on Class 9 Accidents met on February 24, 1986 at Washington, DC. The purpose of this meeting was to review the Draft NRR Implementation Plan for Severe Accident Policy Statement and The Regu-latory Use of New Source Term Information (hereafter designated at the Plan). The Subcommittee heard presentations from NRR/DSR0 staff mem-bers. Copies of the agenda and selected slides from the presentation are attached. The meeting began at 8:30 a.m. and adjourned at 3:30 p.m.

and was held entirely in open sessio. The principal attendees were as follows:

Attendees:

ACRS NRC/DSR0 W. Kerr, Chairman T. Speis M. Carbon, Member Z. Rosztoczy P. Davis, Consultant F. Coffman M. Plesset, Consultant L. Sofer D. Houston, Staff R. Barrett R. Palla 1 W. Pratt, Consultant (BNL)

Discussion In the introductory remarks, W. Kerr expressed appreciation to the Staff for providing a draft copy of the Plan for the Subcommittee's review prior to the meeting. He asked the Staff to clarify the following items in their presentation: (a) Interrelationship of NUREG-0956, NUREG-1150, IDCOR and the Plan, (b) the ability to determine the conditional probabi-lity that given a WASH-1400 core melt, a Denton core melt (i.e., core on 060529021a 060310 PDR ACHG 2400 YDR

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f Class 9 Accidents Minutes February 24, 1986 containment floor) will occur, (c) the consideration of safety goals, and (d) the treatment of external initiators. He commended the Staff for their effort to begin the selection of a regulatory principle.

T. Speis (NRR) briefly described the topics for the Staff's presenta-tion. The Plan will be issued as a SECY document when it is discussed with the Commission on March 26, 1986. Z. Rosztoczy (NRR) presented an overview of the Plan as derived from the Severe Accident Policy State-ment. The Plan addresses three general areas: (1) Systematic Exami-nation of Individual Existing Plants, (2) PRA Guidance and Containment Performance Criteria for Future Plants, and (3) Changes in Rules and Regulatory Practices based on New Source Term Information. The studies in the first area (Individual Plant Examination) would be performed using IDCOR methodology which will be submitted to NRC for review in March 1986. Schedules, relationships with other regulatory programs (e.g., safety goals, NUREG-0956 and NUREG-1150) and potential issues were discussed. W. Kerr inquired about (a) guidance given to IDCOR for initial plant analysis, (b) the establishment of appropriate models (uncertaintyanalysisvs.sensitivitystudies),and(c) containment performance guidelines to use with NUREG-0956 for computing source terms. M. Carbon inquired about the establishment of containment performance objectives if safety goals were not given quantitative values. The Staff acknowledged that many of these issues were still under development. In regard to external events, the Staff does not currently plan to evaluate their contribution but they will issue recommendations on the subject in a Commission paper in the near future.

Z. Rosztoczy concluded this portion of the presentation with a dis-cussion of a process to be used in the selection of severe accidents for regulatory considerations. The screening processes consider both accident prevention and accident mitigation. P. Davis connented that some very low probability sequences with major contribution to core melt frequency would be missed if the cut-off limits proposed by the Staff are adopted.

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Class 9 Accidents Minutes February 24, 1986 R. Barrett (NRR) discussed the status of individual plant examinations (IPE). The general areas covered were IDCOR methodology and IPEs, resolution of NRC/IDCOR technical issues, evaluation of reference plant risks (NUREG-1150),andstrawmanguidelinesandcriteria. Most of the presentation dealt with general approaches and anticipated future activities since none of the planned activities had been completed.

Preliminary insights for a Mark I BWR (Peach Bottom) were presented in terms of dominant accident sequences, plant vulnerabilities and mitigat-ing features. Items to be considered for the development of guidelines and criteria were also discussed. IDCOR methodology will be submitted for review in March 1986 and the Staff expects to complete their eval-uation by October 1986. The resolution of NRC/IDCOR technical issues is expected by July 1986. Resolution was defined in terms of the Staff having a definite position on each issue, not in terms of agreement between the two parties.

F. Coffman (NRR) discussed the status for future plant activities, mostly the role of PRAs. Three tasks were described which are tenta-tively scheduled for comoletion in December 1986: (1) deterministic requirements,(2)guidanceonminimumcontentofPRAs,and(3) guidance on criteria for regulatory use. General considerations of safety issues, PRA content and criteria were treated in a preliminary manner.

L. Sofer (NRR) discussed the source term related regulatory changes.

Potential changes were grouped in three time frame categories: near-term (within6 months), intermediate (upto1 year)andlongterm(2-5 years). The near term items included: revised treatment of accidents in Environmental mpact Statements (EIS), removal of additives in PWR containmentsprays(fissionproductremoval)andcreditforfission product scrubbing in BWR suppression pools. M. Carbon recomended that the Staff use a more realistic approach with a safety factor for their source term calculations rather than treat each input in a conservative manner.

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'l Class 9 Accidents Minutes February 24, 1986 At the conclusion of the meeting T. Speis requested a review of the Plan by the Committee during the March meeting. W. Kerr indicated that NRR should plan for a one hour presentation. He also indicated that the general direction of the NRR Plan seems appropriate. He expressed concerns that the development of the Plan was being driven by the schedule and cautioned the Staff that more importantly, the Plan should be developed correctly.

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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2/21/86 ACRS CLASS 9 ACCIDENTS SUBCOMMITTEE MEETING FEBRUARY 24, 1986 WASHINGTON, DC

- Tentative Presentation Schedule -

Regulatory Implementation Plan for Severe Accidents Policy A. Subcommittee Chairman Remarks W. Kerr 10 min 8:30am B. Introductory Remarks T. Speis 10 min 8:40am C. Regulatory Implementation Plan Z. Rosztoczy 1 hr 8:50am

• • • Break *** 10 min 9:50am D. Individual Plant Examination R. Barrett 2 hrs 10:00am (a) Status and Progress (b) Assessment of Peach Bottom

• • • Lunch *** , 12;00-1:00pm E. Status and. Progress for Future F. Coffman 30 min 1:00pm ..

Plants F. Status and Progress for Source L. Sofer 1 hr 1:30pm Term Application G. Subcommittee Remarks _ ', W. Kerr 15 min 2:30pm

'*** Adjourn **A

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ACRS

Contact:

Dean Houston

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7 NRR STAFF PRESENTATION TO THE ACRS i

SUBJECT:

AN;D THE REGULATION USE OF NEW SOURCE TERM INF DATE: FEBRUARY 24, 1986 PRESENTER: Z0LTAN R. ROSZTOCZY

-PRESENTER'S TITLE / BRANCH /DIV: CHIEF REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.: (301) 492-8016 SUBCOMMITTEE: CLASS 9 ACCIDENTS 8 e

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SEVERE ACCIDENT POLICY SJATEMENT .ACTICR ITEf4 P0Ll'CY' STATEMENT NEW APPLICATIONS EXISTING PLANTS '

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• GUIDANCE ON THE ROLE *

' . SYSTEMATIC APPROACH OF PRAs I FOR TiiE EXhMINATION 0F IN0lVIDUAL PLANTS PERFORMANCE CRITERIA *

IMPLEMENT MODIFICATION FOR CONTAINMENT THROUGH BACKFIT POLICY SYSTEMS

• ~CkANGESINRULESAND REGULATORY PRACTICES, AS NEEDED'

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IMPLEMENTATION PROGPAM El.EMENTS

1. EXISTING PLANT EXAMIf{ATION ,

- PEVIEW 0F THE IDCOR INDIVIDUAL PLANT EXAMINATION METHODOLOGY

- DEVELOPMENT OF GUIDELINES AND CRITERIA FOR PLANT EXAMINATIONS 2 DEVELOPMENT OF GUIDANCE ON THE ROLE OF PRAs FOR NEW PLANT APPLICATIONS

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- DETEPMINISTIC REQUIREMENTS

- ACCEPTABLE CONTENT OF PRAs

- CRITERIA FOR THE REGULATORY REVIEW AND INTERPP.ETATION OF THE PRA PESULTS

3. CHANGES IN RULES AND REGULATORY PPACTICE

- SOURCE TERM PELATED CHANGES SEVERE ACCIDENT RELATED CHANGES 2

EXPECTED ACCOMPLISHMENTS PLANT SPECIFIC VULNERABILITIES WILL BE IDENTIFIED AND FIXED (BACKFIT RULE)

IF GENERIC VULNERABILITIES ARE IDENTIFIED, APPROPRIATE DESIGN AND/0R OPERATIONAL CHANGES WILL BE REQUIRED (RULEMAkING)

LESSONS LEARNED WILL HELP DEVELOPMENT OF IMPROVED DESIGNS WITH SAFETY BENEFITS A NEW, MORE REALISTIC REGULATORY APPROACH ON SOURCE TERMS WILL BE PURSUED (SOURCE TERM RELATED CHANGES) 0

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SUMMARY

OF EXPECTED ACCOMPLISHMENTS SEVERE ACCIDENT POLICY IMPLEMENTATION COMPLETE THE NRC ANALYSIS OF SIX 6/86 REFERENCE PLANTS FOR SEVERE ACCIDENTS INCLUDING SOURCE TERM CALCULATIONS RESOLVE IDCOR/NRC TECHNICAL ISSUES 7/86 COMPLETE THE REFERENCE PLANT SENSITIVITY 7/86 STUDIES (EVALUATION OF UNCERTAINTIES)

COMPLETE REVIEW 0F IDCOR METHODOLOGY 10/86 FOR INDIVIDUAL PLANT EXAMINATIONS BRIEF COMMISSION ON THE FINDINGS AND 12/86 RECOMMENDATIONS FOR THE INDIVIDUAL PLANT EXAMINATIONS ISSUES GUIDANCE FOR PUPLIC COMMENT 2/87 ON THE ROLE OF PRAs FOR NEW PLANT APPLICATIONS ISSUE FOR PUBLIC COMMENT RULE CHANGES 4/87 NECESSARY TO RESOLVE GENERIC SEVERE ACCIDENT RELATED VULNERABILITIES e

SUMMARY

OF EXPECTED ACCOMPLISHMENTS SOURCE TERM RELATED CHANGES ISSUE FOR COMPENT REVISED SRP SECTION 6.5.2 9/86 SPECIFYING THE NEED FOR SPPAY ADDITIVES IN PWRs ISSUE FOR COMMENT REGULATORY GUIDE 1.3 AND THE 9/86 APPROPPIATE SECTION OF THE SRP ON FISSION PRODUCT SCRUBBING IN SUPPRESSION POOLS (BWRs)

ISSUE FOR COMMENT PROPOSED CHANGES TO 10 CFR 50.47 2/87 AND 10 CFR 50, APPENDIX E ON EMERGENCY PLANNING REVISE NRR OFFICE LETTER 16 WITH RESPECT TO THE 2/87 USE OF SOURCE TERMS IN SAFETY ISSUE EVALUATION ISSUE FOR COMMENT CHANGES IN CONTAINMENT LEAK 3/87 RATE REQUIREMENTS, INCLUDING POTENTIAL CHANGES IN 10 CFR 50 APPENDIX J REVISE 10 CFR 50.49 AND REGULATORY GUIDE 1.89 6/87 WITH RESPECT TO THE RADIATION ENVIRONMENT FOR EQUIPMENT OVALIFICATION, FOR COMMENT BY ISSUE FOR COMMENT REVISIONS OF SITING CRITERIA 10/87 (10 CFR 100) BASED ON NEW SOURCE TERM INFORMATION ISSUE FOR COMMENT REVISED REGULATORY GUIDE 1.97 12/87 '

ON ACCIDENT MONITORING AND MANAGEMENT 5

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Reference Plant l Anslyses ) Regulatory

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12/86 II Standards for Review of Acceptable IDCOR Mr.thodology Methodology '

LEGEND 3/86 10/86

, Program Activities jk

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Figure 3.1 Program Element 1 - Development of Guidance for Individual Plant Examinations

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PELATIONSHIPS WITH OTHER PROGPAMS RES PROGRAMS

- NUREG-0900 SUPPLEMENT, RESEARCH PLAN FOR

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SEVEPE ACCIDENTS

- NUPEG-0956, REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS

- NUREG-1150,'REFEPENCE PLANT ASSESSMENT

- UPDATE ON SEVERE ACCIPENT RESEARCH NRP PROGRAMS

- SAFETY G0ALS: FINAL VERSION, CONTAINMENT PERFORMANCE DESIGN OBJECTIVE

- UNPESOLVED AND GENERIC SAFETY ISSUES:

STATION BLACK 0UT, SHl'TDOWN DECAY HEAT REMOVAL

- PPA REVIEWS AND INSIGHTS PEPORTS: INDIAN

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POINT, ZION, LIMERICK, AND GESSAP; PPA INSIGHTS REPORTS, PROCEDURES GUIDE AND REVIEW MANUAL INDUSTRY PROGRAMS

- IDCOR: REFERENCE PLANT ANALYSES, TECHNICAL ISSUES, INDIVIDUAL PLANT EXAMINATION METHODOLOGY ,

- AIF: SOURCE TERM ISSUES, PRA ISSUES t

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POTENTIAL ISSUES LARGE DIFFERENCES COULD-EXIST BETWEEN IDCOR AND -

FRC CALCULATIONS QUANTIFICATION OF THE UNCERTAINTIES COULD RUN INTO DIFFICULTIES COMPLETENESS OF PLANT ANALYSES ARE IN OVESTION BECAUSE OF EXTERNAL EVENTS DEVELOPMENT OF PRACTICAL SOURCE TERMS IS A NON-TRIVIAL ISSUE SCHEDULE IS VERY TIGHT, ADDITIONAL SUPPORT MIGHT BE NEEDED

- THER SNO}JMEALOTEfFORPUB)[ICCOMME(S ONsTHE COMMf'SSIONS S,TATEMENT.ON THE OF R AsFOP$hWPLAphPPLIC)ONSANNNTHE EXAMINAil0NOFEXISTINGPLANTS, O

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RECOMMENDATION ON EYTCANAL EVENTS

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ONLY RELATIVELY LIK$LY SEISMIC EVENTS (2 TO 3 TIMES SSE)

SHOULD BE CONSIDERED.

THE CUT 0FF SHOULD BE ESTABLISHED ON THE SAME BASIS AS FOR OTHER SEVERE ACCIDENT ISSUES, FOR EXAMPLE SOURCE TERM.

THE EFFORT SHOULD CONCENTRATE ON SEISMIC VULNERABILITIES.

RESOLUTION OF THE QUESTION WHAT IS THE CONTRIBUTION OF SEISMIC EVENTS TO OVERALL RISK IS NOT A NECESSITY.

THE METHODOLOGY DEVELOPED UNDER THE SEISMIC MARGIN PROGRAM SHOULD BE REVIEWED, MODIFIED (IF NEEDED) AND APPROVED FOR

_ USE FOR THIS PURPOSE.

THE IMPLEMENTATION EFFORT SHOULD BE CLOSELY C0ORDINATED WITH OTHER REGULATORY ACTIONS RELATED TO SEISMIC ISSUES, FOR EXAMPLE THE RESOLUTION OF USI A-46, SEISMIC QUALIFICATION OF EQUIPMENT.

INTERNALLY INITIATED FLOOD AND FIRE WILL BE INCLUDED IN THE -

IDCOR METHODOLOGY FOR INDIVIDUAL PLANT EXAMINATION 13

SELECTION OF SEVERE ACCIDENTS .

F0H FtGULAIURY CONSIDEHAIIONS SEVERE ACCIDENTS ARE DEFINED AS PEACTOR ACCIDENTS MOPE SEVERE THAN DESIGN BASIS ACCIDENTS, THIS DEFINITION REPRESENTS AN INFINITE SPECTRUM, OUR G0AL IS TO CONCENTRATE ON THOSE SEVERE ACCIDENTS WHICH COULD LEAD US TO SIGNIFICANT IMPROVEMENTS WITH RESPECT TO CORE DAMAGE FREQUENCY, CONTAINMENT PERFOR-MANCE OR SOURCE TERMS, PROPOSED APPROACH:

CONCENTRATE ON SELECTION OF SIGNIFICANT SEQUENCES USE A SCREENING PROCESS TO SELECT THE IMPORTANT SEQUENCES, SCREENING WILL BE PERFORMED SEPARATELY FOR ACCIDENT PREVENTION AND ACCIDENT MITIGATION,

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ACCIDENT-PRE'ENTION SCREEN!NG -

INCLUDE SEQUENCES WHICH CONTRIBUTE MORE THAN 5%

TO THE PREDICTED CORE DAMAGE FREQUENCY, INCLUDE SEQUENCES FOR WHICH THE CORE DAMAGE FREQUENCY IS GREATER THAN 2 x 10-6/ REACTOR YEAR, INCLUDE SEQUENCES USEFUL FOR IDENTIFYING SYSTEMS WEAKNESSES OR OPERATOR ACTIONS BASED ON DETERMINISTIC EVALUATIONS, ENGINEERING JUDGEMENT AND CONSIDERATION OF UNCERTAINTIES IN THE PROBABILISTIC ANALYSIS, 4

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ACCIDENT MITIGATION SCRFENING -

INCLUDE SEQUENCES FOR WHICH THE CONDITIONAL PROBABILITY OF A CONTAINMENT FAILURE IS GREATER THAN 10-2.

INCLUDE SEQUENCES WHICH LEAD TO CONTAINMENT BYPASS AND HAVE A PROBABILITY OF OCCURANCE GREATER THAN 10-7 INCLUDE SEQUENCES WHICH EXCEED A SELECTED CONTAINMENT LEAKAGE LIMIT, LIMIT WILL BE SELECTED LATER, INCLUDE SEQUENCES USEFUL FOR IDENTIFYING CONTAINMENT

_. WEAKNESSES OR OPERATOR ACTIONS BASED ON DETERMINISTIC EVALUATIONS, ENGINEERING JUDGEMENT AND CONSIDERATION OF UNCERTAINTIES.

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NRR STAFF PRESENTATION TO THE

-ACRS

SUBJECT:

STATUS OF THE SEVERE ACCIDENT PROGRAM FOR OPERATING REACTORS DATE: FEBRUARY 24, 1986 PRESENTER: RICHARD J. BARRETT PRESENTER'S TITLE / BRANCH /DIV: NUCLEAR ENGINEER REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW AND OVERSIGHT PRESENTER'S NRC TEL. NO.: (301) 492-4563

( SUBCOMMITTEE: CLASS 9 ACCIDENTS

STATUS OF INDIVIDUAL PLANT EXAMINATIONS ,

REVIEW 0F IDCOR INDIVIDUAL PLANT EXAMINATION METHODOLOGY SUBMITTAL OF METHODOLOGY 3/86 SUBMITTAL 0F 2 BWR AND 2 PWR 3/86 TEST CASES TECHNICAL ISSUE RESOLUTION COMPLETION EXPECTED 7/86 T

EVALUATION OF REFERENCE PLANT RISK

' - ACRS BRIEFED ON JANUARY 29, 1986 STRAWMAN GUIDELINES AND CRITERIA GENERAL APPROACH PRELIMINARY MARK I GUIDELINES & CRITERIA e

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GUIDELINES AND CRITERIA CONSIDERATIONS GUIDELINES AND CRITERIA SHOULD BE STRINGENT COMPENSATE FOR INCOMPLETENESS OF RISK INFORMATION INSIGNIFICANT PROBLEMS IDENTIFIED BY THIS PROCESS CAN BE ELIMINATED BY COST-BENEFIT DETERMINISTIC, BUT BASED ON PRA INSIGHTS COMPATIBLE WITH THE IDCOR IPE METHODOLOGY e

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4 DEVELOPMENT OF GUIDELINES AND CRITERIA G0ALS ,, ,

THE THREE BASIC OBJECTIVES OF THE SEVERE ACCIDENT PROGRAM APPLY EQUALLY TO ALL PLANT TYPES DEFINITIONS RELATE THE G0ALS TO THE DESIGN FEATURES AND OPERATING CHARACTERISTICS OF EACH PLANT TYPE INCLUDES BOTH FAVORABLE AND UNFAVORABLE SEVERE ACCIDENT ATTRIBUTES GUIDELINES

- DETERMINISTIC, PLANT-SPECIFIC GUIDANCE ON THE DESIGN FEATURES AND OPERATING CHARACTERISTICS WHICH ARE TO BE EXAMINED BY UTILITIES CRITERIA DETERMINISTIC STANDARDS FOR JUDGING THE ACCEPTABILITY OF PLANT FEATURES O

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G0ALS

1. MITIGATION OF FISSION PRODUCT RELEASES THERE SHALL BE EFFECTIVE MEANS OF MITIGATING THE FISSION PRODUCT RELEASES FOR THE BROAD CLASSES OF ACCIDENT SEQUENCES 4

WHICH LEAD TO CORE DAMAGE

2. PREVENTION OF HIGH CONSEQUENCE SEQUENCES ALL REASONABLE STEPS WILL BE TAKEN TO REDUCE THE FREQUENCY OF SEQUENCES WHICH DO NOT CONFORM TO G0AL 1: THAT IS, SEQUENCES FOR WHICH THE PLANT HAS NO EFFECTIVE MEANS OF MITIGATING FISSION PRODUCT RELEASES

• - 3. REDUCTION OF CORE DAMAGE FREQUENCY A DETERMINISTIC EXAMINATION WILL BE PERFORMED TO IDENTIFY MEANS OF REDUCING CORE DAMAGE FREQUENCY RELATED TO THE LEADING ACCIDENT SEQUENCES i

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t G G0AL 1 MITIGATION OF FISSION PRODUCT RELEASES THERE SHALL BE EFFECTIVE MEANS OF MITIGATING THE FISSION PRODUCT RELEASES FOR THE BROAD CLASSES OF ACCIDENT SEQUENCES WHICH LEAD TO CORE DAMAGE

MARK I BWR DOMINANT CLASSES OF ACCIDENT SEQUENCES CORE DAMAGE FRQUENCY IS BELIEVED TO BE DOMINATED BY STATION BLACK 0UT (TB) AND ATWS OTHER TRANSIENTS AND LOCAs ALSO CONTRIBUTE. IN GENERAL, THE ABILITY TO MITIGATE RELEASES FOR TB AND ATWS ENCOMPASSES THE OTHER TRANSIENTS AND LOCAS THERE ARE SPECIFIC SEQUENCES FOR WHICH MITIGATION BY THE Mt,RK I CONTAINMENT ISE INEFFECTIVE ATWS WITH POWER TRANSIENT BWR INTERFACING SYSTEMS LOCA MITIGATION OF SOURCE TERMS MITIGATION OF SOURCE TERMS FOR MARK I CONTAINMENT REQUIRES SUPPRESSION POOL SCRUBBING WITH NO SIGNIFI-CANT BYPASS MECHANISMS.

DRYWELL SPRAY OPERATION AND REACTOR BUILDING RETENTION PROVIDE AN IMPORTANT DEFENSE-IN-DEPTH CAPABILITY FOR e . FISSION PRODUCT MITIGATION e

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MARK I SOURCE TERMS PLANT VULNERABILITIES

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CORE CONCRETE INTERACTI0NS DRY CONFINED CAVITY LEADS TO AGGRESSIVE CORE CONCRETE INTERACTIONS AND HIGH CORIUM TEMPERATURES

- MAJOR RELEASES OF NONVOLATILE FISSION PRODUCTS, NON-CONDENSIBLE GASES AND HEAT SMALL DRYWELL VOLUME SUSCEPTIBLE TO OVERPRESSURE AND OVERTEMPERATURE FAILURE WITHIN A FEW HOURS OF VESSEL FAILURE REVAPORIZATION OF IN-VESSEL FISSION PRODUCTS HIGH DRYWELL TEMPERATURES PREDICTED BY IDCOR TO REVOLATIZE FISSION PRODUCTS DEPOSITED IN THE VESSEL LINER MELTTHROUGH MOLTEN CORE DEBRIS CAN MIGRATE TO THE DRYWELL LINER AND CAUSE A LOSS OF INTEGRITY LOSS OF CONTAINMENT ISOLATION

- VACUUM BREAKER VALVES CAN STICK OPEN, THEREBY RELEASING FISSION PRODUCTS TO THE WETWELL WITHOUT POOL SCRUBBING

MARK I SOURCE TERM MITIGATING FEATURES DRYWELL INERTING PREVENTS HYDROGEN-BURN INDUCED OVERPRESSURE FAILURE .

DRYWELL SPRAYS PREVENT OVERTEMPERATURE DRYWELL FAILURE

- DELAY, BUT DO NOT PREVENT, OVERPRESSURE FAILURE REDUCE AIRBORNE CONCENTRATIONS OF FISSION PRODUCTS REDUCE CORE DEBRIS MIGRATION AND LINER MELTTHROUGH SUPPRESSION P00L REDUCES PRESSURIZATION DUE TO STEAM SCRUBS FISSION PRODUCTS WETWELL VENTS ALLOW DEPRESSURIZATION OF DRYWELL AND SCRUBBING OF FISSION PRODUCTS REACTOR BUILDING FURTHER REDUCES FISSION PRODUCT RELEASES TO THE ATMOSPHERE

MARK I STRAWMAN GUIDELINES AND CRITERIA

1. RELIABLE OPERATION OF THE DRYWELL SPRAYS SHALL BE DEMON-STRATED FOR THE POST CORE MELT PERIOD OF ALL DOMINANT ACCIDENT SEQUENCES -- -

CRITERIA: (T0 BE DETERMINED)

CONSIDERATIONS:

CAPACITY AC DEPENDENCY TIME REQUIRED FOR ACTIVATION

2. CONTROLLED WETWELL VENTING SHALL BE DEMONSTRATED AS AN EFFECTIVE OPTION FOR ALL RELEVANT SEQUENCES CRITERIA: (CURRENTLY UNDER DEVELOPMENT BY NRC/RES)

CONSIDERATIONS:

- PROCEDURES, TRAINING, HUMAN FACTORS

- VALVE CAPACITY, OPERABILITY AC DEPENDENCY o .3, EMERGENCY OPERATING PROCEDURES SHOULD BE DEFINED S0 AS

. TO MINIMIZE THE LIKELIHOOD OF DEINERTING THE DRYWELL ATMOSPHERE CRITERIA: (T0 BE DETERMINED)

CONSIDERATIONS:

CIRCUMSTANCES UNDER WHICH SPRAY OPERATION CAN OCCUR AFTER WETWELL VENTING

MARK I STRAWMAN GUIDELINES AND CRITERIA (CONTINUED)

4. THE POTENTIAL FOR DRYWEll LEAKAGE OR OTHER OPPORTUNITIES FOR SUPPRESSION 00L BYPAc5 SHALL BE EVALUATED CRITErdA: (T0 BE DETERMINED)

CONSIDERATIONS:

FELIABILITY OF CONTAINMENT ISOLATION /0PERATING EXPERIENCE RELIABILITY OF VACUUM BREAKER VALVES TO RECLOSE DESIGN AND COMPOSITION OF CONTAINMENT PENETRATIONS MEASURED CONTAINMENT LEAKAGE PLANT-SPECIFIC GE0 METRIC VULNERABILITIES

5. THERE SHOULD BE REASONABLE ASSURANCE THAT THE REACTOR BUILDING

WILL RETAIN A SIGNIFICANT FRACTION OF THE FISSION PRODUCT INVENTORY CRITERIA: (T0 BE DETERMINED)

CONSIDERATIONS:

BUILDING INTEGRITY STANDBY GAS TREATMENT SYSTEM FIRE SPRAYS o

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PRELIMINARY PEACH BOTTOM INSIGHTS THESE PRELIMINARY INSIGHTS ARE PRESENTED FOR THE PURPOSE OF ILLUSTRATING THE TYPE OF RESULTS EXPECTED FROM THE REFERENCE PLANT ANALYSES

1. DRYWELL SPRAYS NOT CURRENTLY AVAIL'ABLE FOR TB SEQUENCE

- DIESEL-DRIVEN FIRE PUMPS AVAILABLE, BUT NOT CURRENTLY CONNECTED TO THE SPRAYS

2. CONTROLLED WETWELL VENTING APPEARS FEASIBLE FOR SEQUENCES OTHER THAN TB NRC/RES CURRENTLY EVALUATING ENGINEERING DESIGN AND HUMAN FACTORS

3. DRYWELL DEINERTING NOT YET EVALUATED

4. POOL BYPASS MECHANISMS NOT YET EVALUATED g ,

5. REACTOR BUILDING PERFORMANCE

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BUILDING REMAINS INTACT (EXCEPT FOR BLOWOUT PANELS)

NO FIRE SPRAYS SGTS BYPASSED SIGNIFICANT RETENTION EXPECTED f

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- NEXT STEPS

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. GUIDELINES AND CRITERIA

.  ; REVIEW AND REVISE BWR MARK I GUIDELINES

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.'*,4. CONCENTRATE ON DEFINING BWR MARK I CRITERIA s

IDENTIFY STRAWMAN GUIDELINES AND CRITERIA FOR

- s OTHER PLANT TYPES

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IDCOR IPE METHODOLOGY

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METHODOLOGY AND FOUR PLANT APPLICATIONS DUE FOR SUBMITTAL

-IN MARCH; 1986_

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• REPORT TO IDCOR ON MAJOR SHORTCOMINGS 0F THE METHODOLOGY BY JULY, 1986

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NRR STAFF PRESENTATION TO THE ACRS khhk RE NT US OF E b bhCh khk Nk k lhN i

FEBRUARY 24, 1986 DATE:

PRESENTER: ROBERT Pall.A M CHANICAL ENGINEER

-PRESENTER'S TITLE / BRANCH /DIV: REGULATORY IMPROVEMESTS BRANCH DIVISION OF SAFETY REVIEW & OVERSIGHT PRESENTER'S NRC TEL. NO.: (301) 492 4609 SUBCOMMITTEE: CLASS 9 ACCIDENTS p #

t t IDCOR ISSUE f0N' TORS

!@3 BLS .

1. IN-VESSELFPRELEASE R.PALLA L, GAN

2. RECIRCULATION IN-VESSEL Y. LEUNG J. !iAN

3. PELEASE OF CONT. ED MATLS. R. BAPETT L. CHAN/

T.WAtkER

4. RCS AEROSOL DEPOSIT.10N J, READ L. CHAN

5. IN-VESSEL HgGENERATION 6.PALLA J. HAN

6. CORE SLUMP MDDEL R,PALLA R. WRIGiT

7. STEAM EXPLOSIONS C. ALLEN J.TELF00

, 8. DJPECT HEATING F. ELTAWILA T. LEE j

9. EX-VESSEL HEAT TRANSFER B.HARDIN B,BURSON t0. EX-VESSEL FP RQ EAKE 3. HGDIN B. SURSON

11. REVAPOR12AT10N R. SA M S L,OMN

12. CCMA!! TENT AEOSCF. DEPOSITION J. READ J.TELFORD 13A. SUFPRESSICA' 000L BOSS J. READ J. MITCHELL/

D. PVATT

, 138, FP PENDVAL IN ICE COND. J. READ

14. E ERGENCY RESPCHSE @ DELItG J, READ

15. CONTAINTVT PERFORVANCE F. ELTAWILA J. COSTELLO LB SECONDARY CONTAINENT PERF0MCE R. BARRETT T VALKER

17. 11 IGNITION AND BURNING R,PALLA P. WORTHINGTON 2 ',

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18. E0. D RVIVABILITY S. SANDS W. FAPER f

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a e lh STAFF ASSESSEhT OF IMPORTANCE OF ESOLUTION ON POLICY IELEENTATION STAFF ASSESSE NT .

UNIMPORTANT POTENTIALLY 1SSUE OR RESOLVED IMPORTAN_T ITORTANT

1. IN-VESSEL FP RELEASE X

2. ECIRCULATION IN-VESSEL X

3. RELEASE OF CONT. R0D VATLS. X 14 RCS AEROSOL DEPOSITION X

5. IN-VESSEL HpGENERATION X

6. COE SLUMP NDDEL X

7. STEAM EMPLOSIONS X

8. DIECT HEATING X

9. EX-VESSEL HEAT TRANSFER X LO. EX-VESSEL FP ELEASE X

11. REVAPORIZATION X

12. CONTAINENT AEROSOL DEPOSITION X 13A. SUPPESSION POOL BYPASS X 138. FP REMOVAL IN ICE CON'0. X 114 E E RGENCY E SPONSE NDDELING X

15. CONTIANENT PERF0fFANCE X

16. SECONDARY CONTAINMENT PERFORt'ANCE X

17. H IGNITION AND BURNING X 2

18. E0. SURVIVABILITY X O

NRR STAFF PRESENTATION TO THE ACRS

SUBJECT:

STATUS AND PROGRESS FOR FUTURE PLANTS, ROLE OF PRAS, SEVERE ACCIDENT POLICY IMPLEMENTATION DATE: FEBRUARY 24, 1986 PRESENTER: F.D. C0FFMAN

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PRESENTER'S TITLE / BRANCH /DIV: SECTION LEADER REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW 8 OVERSIGHT PRESENTER'S NRC TEL. NO.: (301) 492-4914 l

SUBCOMMITTEE: CLASS 9 ACCIDENTS -

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TASK 4.2.

GUIDANCE ON THE MINIHJM CONTENT OF PRAs INPlJTS: NUREG/CR-2815, PSA PROCEDURES GUIDE .

NUREG/CR-4550, REFERENCE PLANT lElliOD NUREG/CR-2728, IREP PROCEDURES GUIDE NUREG/CR-2300, PPA PROCEDURES GUIDE N APS STUDY GROUP ON SEVERE ACCIDENT SCREENED BY: IPPORTANCE TO ANTICIPATED USES AND PAST SENSITIVITY STUDIES TO PRODUCE: MINIMJM CONTENT OF PRAs CONCERNING:

SCOPE OF HAZARDS QUALITY & APPLICABILITY OF DATA LEVEL OF DETAIL FOR DEPENDENCIES TREATPENT OF OPERATOR ERROR RATES EFFECTS OF COPMDN-MDDE FAILURES

, SEARCH FOR COPNDN-CAUSE FAILURES

MAGNITUDE OF UNCERTAINTIES i.

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TASK 4.3 GUIDANCE ON CRITERIA FOR REGULATORY USE INPUTS: STAFF TESTIMONY FOR IPPSS. .

NUREG-1068, REVIEW OF LGS PRA AND SARA N'J REG-0979, GESSAR-II SER ACRS LETTER TO CHAIRMAN PALLADINO, ON GESSAR, JANUARY 14,1986 NUREG/CR-3485, PSA REVIEW MANUAL NUREG-1150, INTEGRATED RISK ASSESSW NT REPORT INSIGHTS GAINED FROM PROBABILISTIC RISK ASSESSENTS DATED DECEN ER 3, 1984 NUREG/CR-4405, "PROBABILISTIC RISK ASSESSENT INSIGHTS" SHOREHAM PRA REVIEW OCONEE 3 PRA REVIEW MILLSTONE 3 PRA REVIEW SCREENED BY: E ASURABLE SAFETY BENEFIT TO PRODUCE: GENERIC REQUIREENTS AND ACTION THRESHH0LDS (STAFF ACCEPTANCE CRITERIA ON ACCURACY, CONSISTENCY, AND CLARITY)

ANTICIPATED DIFFICULTIES A. SCHEDULE IS HER0IC:

ONE MONTH FOR C0mlSS10N 8 ACPS REVIEW .

REFERENCE PLANT ANALYSES COULD BE LATER THAN 6/86 PROBING QUESTIONS REPAIN MANY ORGANIZATIONS TO BE COORDINATED B. TECHNOLOGY IS DEVELOPING:

PROCEDURES FOR CONTAINTNT AND CONSEQUENCES ARE ASSOCIATED WITH P0DELING VARIATIONS

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EXPLICIT HANDLING OF UNCERTAINTIES IN DECISIONS WILL REQUIRE SOE RETOOLING PRA KTHODS ARE DIVERSE C. REGULATION IS CONTINUOUS:

POTENTIAL FOR IltEDIATE DISPOSITION OF Iff0RTANT DEVELOPPENTS f

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NRR STAFF PRESENTATION TO THE ACRS

SUBJECT:

SOURCE TERM RELATED REGULATORY CHANGES DATE: FEBRUARY 24, 1986 PRESENTER: LEONARD SOFFER PRESENTER'S TITLE / BRANCH /DIV: SECTION LEADER

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REGULATORY IMPROVEMENTS BRANCH DIVISION OF SAFETY REVIEW & OVERSIGHT PRESENTER'S NRC TEL. NO.: (301) 492-7976 SUBCOMMITTEE: CLASS 9 ACCIDENTS

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POTENTIAL SOURCE TERM CHANGES NEAR-TERM INTERMEDIATE

• LONG-TERM REVISED TREATMENT EMERGENCY PLANNING

• SITING OF ACCIDENTS IN EIS REMOVAL 0F SPRAY CONTAINMENT LEAK RATES

• ACCIDENT ADDITIVES (PWR) AND INTEGRITY MONITORING SUPPRESSION POOL ENVIRONMENTAL QUALIFICATION CREDIT (BWR) 0F EQUIPMENT SAFETY ISSUE EVALUATION CONTROL ROOM HABITABILITY AND AIR FILTRATION SYSTEMS

' STAFF STUDIES PROCEED IN PARALLEL WITH NUREG-1150, BUT NO CHANGES UNTIL AFTER ISSUANCE OF NUREG-1150, 9

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., **O PWR SPRAY ADDITIVES PRESENT PRACTICE MOST PWP'S EMPLOY CONTAINMENT SPRAY SYSTEMS WITH CHEMICAL ADDITIVES (E.G., NA0H).

SYSTEM HAS TWO SAFETY-RELATED FUNCTIONS CONTAINMENT HEAT REMOVAL IN ACCIDENT FISSION PPODUCT (IODINE) CLEANUP IN CONTAINMENT ATMOSPHERE SPRAY SYSTEM CLEANUP PERFORMANCE IS DRIVEN BY REG. GUIDE 1.4 ASSUMPTIONS PLUS REQUIREMFNT (TOGETHEP WITH CONTAINMENT LEAK RATE)

THAT DBA ACCIDENT DOSES MEET GUIDELINE VALVES OF 10 CFR 100.

REG. GUIDE 1.4 ASSUMPTIONS

-100% OF NOBLE GASES 25% OF 10 DINES CONSIDERED INSTANTANEOUSLY AVAILAPLE FOR RELEASE FROM CONTAINMENT ASSUMED CHEMICAL FORM 0F 10 DINE

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91% ELEMENTAL 10 DINE 5% PARTICULATE 10 DINE 4% OPGANIC IODINE ASSUMED APPEARANCE OF 10 DINE (TIMING) AND CHEMISTRY VIPTUALLY REQUIRF IMMEDIATF INITIATION OF SPRAY SYSTEM AUTOMATIC CHEMICAL ADDITIVE e e

OPERATIONAL INSIGHTS USE OF CHEMICAL ADDITIVE ADDS SYSTEM COMPLEXITY IF SPRAY SYSTEM OPERATES PPOMPTLY AS DESIGNFD, INJECTION PHASE MAY BE OVEP BEFORE ANY FISSION PRODUCTS APPEAR IN CONTAINPENT ATMOSPHEPE INADVERTENT SPRAY OPEPATION WITH ADDITIVE EXPOSES EQUIPMENT TO CORROSIVE, POTENTIALLY DAMAGING ENVIRONMENT ese O

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PROPOSFD CHANGES REMOVE REQUIREMENT FOP CHEMICAL-ADDITIVES TO PWR SPPAY SYSTEMS

• RETAIN REQUIREMENT FOR APPROPRIATE DEGR'E OF PH CONTROL IN SUMP, TO PREVENT LONG-TERM IODINE EVOLUTION m

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I t PRESSURE SUPPRESSION P0OLS PRIMARY PURPOSF Is To AeroRB HEAT REG. GUIDE 1.3 FORBIDS ROLE 0F POOLS AS FISSION PRODUCT CLEAN-UP SYSTEMS muser O

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8 S P00Ls' ADVANTAGES IN CLEAN-UP ROLE DISSOLVE SOLUBLE FISSION PRODUCTS SCRUB AEROSOLS Passive m

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FACTORS AFFECTING EFFECTIVENESS CARRIEP GAS COMPOSITION AND TEMPERATUPE AEROSOL DENSITY AND SIZE DISTPIEUTION BY-PASS PATHWAYS m

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PROPOSED REGULATORY CHANGES REVISE REG. GUIDE 1.3 DEVELOP STANDAPD REVIEW PLAN FOR Poots AS FISSION PRODUCT CLEAN-UP SYSTEMS I

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EXPECTED PEGULATORY IMPACT i

REDUCED REQUIREMENTS ON SGTS FILTER EFFICIENCY GREATER IMPORTANCE'0F DRY-WELL LEAKAGE I

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PRESENT REGULATORY APPROACH ON SOURCE TERMS LARGE FISSION PRODUCT RELEASE POSTULATED IN CONTAINMENT DERIVED FROM TI'D-14844, FUPTHER REFINED le PEG GUIDES 1.3/1.4 NOBLE GASES, 10 DINES TIMING 8 IODINE CHEMICAL' FORMS PRESCRIBED

• POSTULATED RELEASE, TOGETHER WITH CONTAINMENT LEAK RATE AND OTHER FISSION PP0 DUCT CLEANUP SYSTEMS, USED TO CALCULATE DOSES TO HYPO-THETICAL INDIVIDUALS AT EAB AND LPZ DOSES CALCULATED CONSERVATIVELY DOSES COMPARED WITH VALVES IN 10 CFR 100 CONSEQUENCES ACCEPTABLE lF DOSES BELow 10 CFR 100 VALVES POSTULATED RELEASE IS A MAJOR FACTOR IN DETERMINING WHETHER CONSE-OUENCES OF DBA'S ARE RAD 10 LOGICALLY ACCEPTABLE m

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REGULATORY APPLICATIONS OF TID-14844 RELEASE USED TO SET CONTAINMENT LEAK RATE FOR LIMITING DBA USEDTOASSEssPERFORMANkEFORFissIONPRODUCTCLEANUPSYSTEMS (SPRAYS, FILTERS, ICE CONDENSERS)

USED TO DETEPMINE SUITABILITY OF PLAPT-SITE COMBINATION ALSO SETS CONTROL ROOM HABITAPILITY UNDER ACCIDENT CONDITIONS USED TO DEFINE RADIATION ENVIRONMENT FOR SAFETY-RELATED EQUIPMENT USED TO SET REQUIREMENTS FOR MSIV - LEAKAGE COLLECTION SYSTEMS 4

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e e r_ - _ _ , - _ . _ _ , . ._ _ _ , . . . , , , _ . ____.__-.._,,m.,_ - _ . . . , , _f.. 7r-,,. .,m,. _ - . _ _ . .,e .. . ~ .-. ,,_ - - . , . . , .

ADVANTAGES OF PRESENT APPROACH SIMPLE SOURCE TERM FAMILIAR '

LARGE FISSlos PRODUCT RELEASE PLUs EMPHASIS ON Low LEAKAGE CONTAINMENT AND ADDITIONAL CLEANUP SYSTEMS ENSUPES HIGH LEVEL OF SAFETY AND THAT CONSE0VENCES OF LESSEP ACCIDENTS VERY Lost i

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DISADVANTAGES OF PRESENT APPPOACH FISSION PRODUCT RELEASE ASSUMPTIONS NOT IN ACCORD WITH RESEARCH RESULTS

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NOT TIED To ANY PAPTICULAR ACCIDENT OP GPOUP 0F ACCIDENTS

• RELEASE ASSUMPTIONS EMPHASIZE 10D;NE MITIGATION SYSTEMS -

NOT CLEAR HOW.0THER NUCLIDES AFFECTED CONFUSION AND CONTROVERSY AS TO WHETHEP RELEASE REPPESENTS A DESIGN BASIS OP A SEVERE ACCIDFNT h

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n POSSIBLE REVISED APPP0ACH -

WISH TO RE-EXAMINE THE USE OF THE TID-14844 RELEASE IN LICENSING POSSIBLE APPROACH l CONSIDER A NUMBER OF ACCIDENT SEQUENCES LEADING TO CORE DEGRADATION, MELT AND RELEASE INTO CONTAINMENT l

• l EVALUATE ACTIONS OF NATURAL REMOVAL PROCESSES AS l WELL AS FISSION PRODllCT CLEANUP SYSTEMS TO EVALUATE i

PROGRESSION OF ACCIDENT AND CALCULATE RELEASE OF FISSION PRODUCTS TO CONTAINMENT FROM AB0VE, DETERMINE FISSION PRODUCT TYPES, AMOUNTS AND TIME-DEPENDENT CONCENTRATIONS IN CONTAINMENT THAT GENERALLY ENVELOPE SEQUENCES ESTIMATE TIME-DEPENDENT CONTAINMENT LEAK RATE AND CALCULATE FISSION PRODUCT LEAKAGE FROM CONTAINMENT CALCULATE DOSES TO HYPOTHETICAL INDIVIDUALS AT EAB AND LPZ AND COMPARE TO PART 100 (MAY HAVE TO ADD OTHER ORGAN DOSE CRITERIA)

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