ML20245B802
| ML20245B802 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/27/1986 |
| From: | Leech P Office of Nuclear Reactor Regulation |
| To: | Lainas G Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20244D847 | List:
|
| References | |
| FOIA-88-198 NUDOCS 8904260285 | |
| Download: ML20245B802 (79) | |
Text
iziKZa%TXL: M lirum m:.2..M Lc:3:cu::V& u..6%.:3.2 m r.wvuw m :m .a.tw L August 27, 1986 NOTE TO: Gus Lainas, Assistant Director,'RWR Division THRU: John Zwolinski, Director, BWD#1 FROM: Paul Leech, Project Manaoer, BWD#1-
SUBJECT:
RESPONSE TO SHOW CAUSE PETITION, RE: PILGRIM STATION As you are aware, Jerry Klingler of IAE has requested our input for the response to the Show Cause Petition from William B. Golden and others with respect to Pilgrim Station by September 29, 1986. At Jerry Hulman's suggestion, I h6ve developed the following schedule to meet that request:
September 10 - Draft by Hulman and Kudrick provided to Lainas.
September 15 - Comments by Lainas to Hulman/Kudrick.
September 19 - Proposed final version provided by Lainas to Bernero.
, September 23 - Changes by Bernero provided to Lainas, Hulman/Kudrick and Leech.
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September 26 - Final version forwarded by Rernero memorandum to I&E. '
I propose that preparation of the NRR input to I&E proceed in accordance with the above schedule.
ORIGIRAL SIGNED BY Paul Leech, Pro,iect Manaaer BWR Pro.fect Directorate #1 cc: R. Bernero L. Hulman J. Kudrick J. Zwolinski G. Klingler BWD#1 R/F l
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i 1 DISCUSSION OF A GENERIC LETTER ON BWR CONTAINMENT PERFORMANCE - SEPTEMBER 11, 1986 i l' ROBERT M. BERNER0, USNRC D 4
p=.awnm.= e=u:a ..=n:.x.ax: .;;.. . . .u.. .. = :. , =.1a . a:. .. ..:;~ . a. - 1 . 1 GENERIC LETTER ON .. BWR CONTAINMENT PERFORMANCE THE SETTING: PLANT EVALUATIONS UNDER THE SEVERE ACCIDENT POLICY STATEMENT THE KEY REGULATIONS: GDC 16 AND GDC 50 THE SUBJECTS: 37 BWRS WITH PRESSURE SUPPRESSION CONTAINMENTS THE METHOD: A GENERIC LETTER OF REQUIREMENTS TO IMPLEMENT CHANGES BASED ON GENERIC EVALUATION i . l I L= ._ = __ _ . ~~
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, 2 NRC SEVERE ACCIDENT l: POLICY STATEMENT l
l i l i e THE MOST COST-EFFECTIVE OPTIONS FOR REDUCING THIS I VULNERABILITY SHALL BE IDENTIFIED AND A DECISION SHALL BE REACHED CONSISTENT WITH THE COST-EFFECTIVENESS CRITERIA 0F THE COMMISSION'S BACKFIT POLICY AS TO WHICH OPTION OR SET OF OPTIONS (IF ANY) ARE JUSTIFIABLE AND REQUIRED TO BE IMPLEMENTED. e IN THOSE INSTANCES WHERE THE TECHNICAL ISSUE G0ES BEYOND CURRENT REGULATORY REQUIREMENTS, GENERIC RULEMAKING WILL BE THE PREFERRED SOLUTION. IN OTHER CASES, THE ISSUE SHOULD BE DISPOSED OF THROUGH THE CONVENTIONAL PRACTICE OF ISSUING
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BULLETINS AND ORDERS OR GENERIC LETTERS WHERE MODIFICATIONS ARE JUSTIFIED THROUGH BACKFIT POLICY, OR THROUGH PLANT-SPECIFIC DECISION MAKING ALONG THE LINES OF THE INTEGRATED SAFETY ASSESSMENT PROGRAM (ISAP) CONCEPTION. i e
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CRITERION 16 - CONTAINMENT DESIGN. --AN ESSENTIALLY LEAK-TIGHT BARRIER AGAINST THE UNCONTROLLED RELEASE OF RADI0 ACTIVITY TO THE ENVIRONMENT AND TO ASSURE THAT THE CONTAINMENT DESIGN CONDITIONS IMPORTANT TO SAFETY ARE NOT EXCEEDED FOR AS LONG AS POSTULATED ACCIDENT CONDITIONS REQUIRE " a 1 L_- _ _ _ _ . '_~~~~~~~ ' ~ ;' "*-
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1 GDC 50: CRITER10N 50 - CONTAINMENT DESIGN BASIS. --ASREQUYREDBY j SECTION 50.44, ENERGY'FROM METAL-WATER AND OTHER CHEMICAL REACTIONS THAT MAY RESULT FROM DEGRADATION BUT NOT TOTAL FAILURE OF EMERGENCY CORE COOLING FUNCTIONING, (2) THE LIMITED EXPERIENCE AND EXPERIMENTAL DATA AVAILABLE FOR DEFINING ACCIDENT PHENOMENA AND CONTAINMENT RESPONSES, AND (3) THE CONSERVATISM 0F THE CALCULATIONAL MODEL AND INPUT PARAMETERS." 9 - - -_ ___ ______.u __._mm_ _-.. _ .m a . .A
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6 INSTITUTIONAL PROCESS e CLOSURE OF' SEVERE ACCIDENT ANALYSIS FOR CONTAINMENT NO FURTHER ANALYSIS UNLESS EXCEPTION IS TAKEN e SPECTRUM 0F OPTIONS RULEMAKING 50.5l4F LETTER FOLLOWED BY ORDER GENERIC LETTER FROM DIRECTOR NRR OR DIRECTOR DBWRL e PROCESS CHARACTERISTICS
- BASED ON TECHNICAL WORK AVAILABLE, IDCOR, SOURCE TERM, PLANT SPECIFIC WORK, NUREG-1050, NUREG-1150 OPEN TO PUBLIC FOR COMMENT AND PARTICIPATION h
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7 I l A BWR - MARK I J FOR REFERENCE BEFORE
- o. CORE MELT FREQUENCY: 1x10-4/YR l A' FULL SPECTRUM OF SEQUENCES INCLUDING BLACK 0UTS e CONTAINMENT CAPABILITY: UNCERTAIN AND VARIABLE BUT ASSUME 1 l OUT OF 2 CORE MELTS GIVES FAIRLY LARGE RELEASE AFTER e CORE MELT FREQUENCY: 1x10-4/HR IPE FOR FRONT END MAY REDUCE BUT NO CREDIT IS TAKEN HERE e CONTAINMENT CAPABILITY: SUBSTANTIAL ASSURANCE THAT CONTAINMENT WILL MITIGATE CONSEQUENCES, DEGREE VARIABLE FROM
[ PLANT TO PLANT BUT 1 OUT OF 50 CORE MELTS GIVING A FAIRLY LARGE RELEASE SHOULD BE REPRESENTATIVE L l g .
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8 HYDR 0 GEN CONTROL PROPOSED REQUIREMENTS e CHANGE TECH. SPEC. AT END OF OPERATION FROM 24-H.00R ' ALLOWANCE TO 12-HOUR ALLOWANCE OF NON-INERTED OPERATION AI REDUCED POWER e PERMIT 12-HOUR PERIOD AT REDUCED POWER WITHIN THE OPERATING CYCLE TO SEARCH FOR UNIDENTIFIED LEAKAGE RATIONALE e DEINERTING TYPICALLY TAKES 4-8 HOURS e LEAKAGE INSPECTION AND MINOR REPAIR CAN BE REASONABLY ACHIEVED IN 4-8 HOURS e REDUCED POWER ( 6 33%) SUBSTANTIALLY REDUCES SHORT-LIVED FISSION PRODUCT INVENTORY AND DYNAMICS OF POSSIBLE ACCIDENTS -____._______-___m______.____m__ .m. . _ _ _ _____ _ . . __. .
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. 9 DRYWELL SPRAY PROPOSED REQUIREMENTS e REDUCE DESIGN SPRAY RATE (CHANGE N0ZZLES) TO AB0UT 10% OF ,
PRESENT VALUE e PROVIDE AC-POWERED BACKUP WATER SUPPLY FOR SPRAY AND AC-INDEPENDENT WATER SUPPLY, AVAILABILITY BY REMOTE MANUAL OPERATION OR BY SIMPLE RELIABLE PROCEDURE DESIRABLE , e MAK ALTERNATE WATER SOURCES AVAILABLE TO COOL CORE DIRECTLY e 90/10 MODE OF RHR OPERATION RATIONALE h a WATER SUPPLIES AND EQUIPMENT ARE ALREADY AVAILABLE FOR LOWER FLOWS 4 e LOWER. FLOWS PROVIDE ALL BENEFITS EXCEPT LOW 4 T DECAY HEAT REMOVAL AND DO NOT RAPIDLY FLOOD CONTAINMENT e ASSURED DRYWELL SPRAY SUBSTANTIALLY REDUCES PROBABILITY AND SIGNIFICANCE OF DRYWELL FAILURE OR SUPPRESSION POOL BYPASS
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WmT:mErzn : .:n amm:.s:sa::.. T. o ....c-:.w. .:'.: s . > . 10 PRESSURE CONTROL PROPOSED REQUIREMENTS e RELIABLE CAPABILITY TO VENT WETWELL AT EPG PRESSURE LEVEL WITH OR WITHOUT AC POWER. FOR VENTING WITHOUT AC POWER MANUAL PROCEDURE IN ADVANCE MAY BE USED IF NITROGEN PURGE IS AVAILABLE e VENT OF 18-INCH DIAMETER OR GREATER DESIRABLE j e ABILITY TO VENT SLOWER SEQUENCES THROUGH STANDBY GAS TREATMENT SYSTEM e BURST RESISTANCE DUCTING IN REACTOR BUILDING TO MINIMIZE COMPLICATIONS RATIONALE e RELIABLE VENTING PREVENTS UNCONTROLLED OVERPRESSURE FAILURE ! WHICH CAN CAUSE CORE MELT e VENTING WITH DRYWELL SPRAY GIVES GREAT ASSURANCE OF RELEASE MITIGATION _ _ _ _ _: " - - ' L : ~
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'.. 11 CORE DEBRIS PROPOSED REQUIREMENTS ,
e ASSURE RETENTION OF WATER AT LEAST'3 FEET DEEP IN TORUS ROOM IF TORUS LEAKS ENTIRE CONTENTS DESIRABLE e CONCRETE CURBS OR OTHER BARRIERS WHICH WOULD RETARD DEBRIS ATTACK OF DRYWELL SHELL e AVOID LOSS OF RECOVERY SYSTEMS FROM WETTING BY TORUS ROOM WATER RATIONALE ( e DRYWELL FAILURE BY DEBRIS ATTACK IS MADE LESS LIKELY AND LESS SIGNIFICANT BY DRYWELL SPRAY AND VENTING e RETENTION OF TORUS WATER ENSURES DEBRIS QUENCHING AND SHOULD FACILITATE ACCIDENT RECOVERY i i I l i
M i l w . % D L 3 1 5 b:L W , J o M . x .. s . u . u : . .w . ..: . ,: - CHRONOLOGY e JUNE 16, 1986: MEETING WITH BWROG/IDCOR PROPOSED A GENERIC LETTER, PRESCRIPTIVE SOLUTION, BY BACKFIT e JUNE 30, 1986: VERMONT YANKEE COMMITS TO GOV. KUNIN TO DO A SPECIAL 60-DAY CONTAINMENT STUDY e JULY 25, 1986: BOSTON EDISON COMPANY BOARD DECIDES TO FIX PILGRIM CONTAINMENT e AUGUST 19, 1986: BWROG EXECUTIVES VOTE TO FUND AND CONTINUE DIALOGUE ON THIS WITH NRC, CONTACT NUMARC ABOUT BWR VS, PWR e SEPTEMBER 11, 1986: MEETING WITH BWR0G TO COMPARE BACKFIT NOTES AND STRAWMAN GENERIC REQUIREMENTS e SEPTEMBER 11, 1986:. MEETING WITH VERMONT YANKEE TO REVIEW CONTAINMENT STUDY e SEPTEMBER 23, 1986: NRC/IDCOR MEETING ON BWR/ MARK I ANALYSES e SEPTEMBER 23, 1986: ACRS SUBCOMMITTEE ON CONTAINMENT PERFORMANCE TO DISCUSS HARPERS FERRY WORKSHOP RESULTS AND BWR CONTAINMENT SENERIC APPROACH e SEPTEMBER 24, 1986: ACRS SUBCOMMITTEE ON CLASS 9 ACCIDENTS TO DISCUSS BWR/ MARK I ANALYSES AND SEVERE ACCIDENT PROGRAM e NOVEMBER 19, 1986: CRGR REVIEW 0F DRAFT GENERIC LETTER ON BWR CONTAINMENT REQUIREMENTS (T0 BE PUBLISHED FOR COMMENT) e DECEMBER 17, 1986: ISSUE DRAFT GENERIC LETTER ON BWR CONTAlt.'NENT REQUIREMENTS FOR PUBL'C COMMENT e APRIL 1987: ISSUE FINAL GENERIC LETTER ON BWR. CONTAINMENT REQUIREMENTS .___L_-__- _ _ _ _: ' _:_ _ ~_"'~ ~~~~ ~' ' A
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- Established ECOG Precracs to al.ircss FjArrgen Fule requira ents f:r
" Degraded Ccre Accide.ts" .
- Quader S: ale Testing Frcgra .
- Analytical effort
- Station Blackout as a h i .r gen genera *.icn eve.t (:-vZ) withir. the context of recoverable d-gr-ded ccres is an issue b2ing addressed.
- Current HCOG evaluaticn ir. dica *e +'-=' <30 is not a credible H E
- HCDG respcnding to h7C cuestions !
- The need for an inde;r:nder.t pc.er su,q-ly for ig-dters in the event cf an 5B0 ide .tified by the NPC in the cct c.xt of Severe Accide.ts, l
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ELEMENT 2 -
- 3E EE !5 Ji?..T.lyEl
^
REDUCE LIKE LIHCCD 7 AI LURE EY DIREC7 A TTACK ocepi:rvc"
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_. JSE 3 ACTICA L DEEEIS .EIAPDING EARRIE3E arc rawasAL
- 2. .. CfBM suFF-E5SI rJ; POOL #TER AS A
',UENC h I NG P 03 L moG to supply the design criteria for a backup power supply to the hydrogen igniters.
* @fCM%/foieKs%giltUMit$rgbe safety related
- Identify impact of addressing Severe Accidents on the design of a backyp p@ferdnMply.
- Ntrber of igniters required in the event of an SBo o CONCEPTS CONSIDERED
- Make use of existing data base and criteria
- Additional tastiAq)j@n}f; 1F_M@M4ry]PEM N7o11ow capletion of current Test Program - end of this year INCREASE SUMP SIZE INSIDE PEDESTA L Responsibility of individual Mark III owners with support of HCOG as required.
CURB CUTSIDE PEDESTA L OPENING
- Define backup power supply source
- Define associateglpaJtAT DRYWELL LINER /FLOCR JUNCTION
- Meet and discuss with the NBC the details of the design, costs, and benefits.of AI)ppp;g[ippteegepgl% to:'$htfugf(tef8) PROMOTE EVEN
- Decision and timppg-lfpr3(rDpe*Hng
- NEGATIVE IMPACTS LOCA C CNSIDEFATIONS SEISMIC INTEPACTICNS A LAPA C Ct4C ERNS .
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. ,, SPRAYS , .
- - - ..iJ ._ . ~ . . . . .--0 ..2: - - -
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GJECTiVE : 'yp' SPRAY {i/CDATER TO}:-
- 1. yort ncNT r
- i v7tr a i QUEBCR DEBdrIS 2"1PRfAN7),v0DE LS OF DEERIS e.- ..:a :, :. ;: -y n:
r: ;7<EROSOLS (SEdOSDARY) a7;p;
- 2. 'SCROB A
- 3. L04ER PRESSURE (SEC WDARY)
- ^'i WE 1 4. COOL VULNERABLE EQUIPMENT (SECONDARY)
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REQUIR5 MENT 5f,"s,-* ~' '~" ~ ~~ ~ ' ' " ' '" ' ' ' " p *, , ; ' " 3
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WATER IN TC:?.5 C'AMEEP 0F. 3ROVIDE PROTEC TI G TC L cpi iCASPMi l l# D3YWEqLLcc3,gg Roms
- 2. BACKUP WATER SOURCES AND PUMPS
- QUA LITATIVE C H EE-MEC_t(ONS
- USE OF FIREMAINS
- - NO NEGAi!VE "fFACTS IDCOR/BWROG EVALUATIONS:
0 ED 0 ThllAFfMFMANN 10,000 GPM/ HEADER o CONCEPTS CONSIDERED o CONNECTIm TO HOSE STATIM IN REACTG BUILDING
- APPRmIMATE FLOV PROVIDED 200 GPM
- DES NOT PROVIDE SPPAY L____-______--____
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- v- eb.APRROXIf%TF, i sum:u.-et in FLCW.PROSIDED ot.tn c .u 1 1/3i g. FIRE PUMP RATING 3 EQUI.8EvENIX:
- CLOSE OFF APPROXIl%TELY 70% T N0ZZLES
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ue : ,/3. #u.L,-. .e. .M-o POTENTIAL BENEFIT / RISK WARRANTS FURTHER STUDY o :EV. 4 IMPLEY,ENTATION SY A LL UTI LITIES C BSISTENT WITH PREVI OJS P CST-TMI CCMMITHENT C PROPCSE REVIEW OF REV. 4 WITH .. 31GHTS FRG SEVERE ACCIDENT STUDIES
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OBJECTIVES: 1. AVERT UNCONTROLLED OVERPRESSURE FAILURE l o AUGUST 19 MEETING - 20 0F 23 BWROG UTILITIES REPRESENTED
- 2. CONTROL RELEASE PATH (SCRUBBING) i o C0i,T N.E CONSTRUCTIVE DIALOGUE WITH NRC l REQUIREMENTS:
L n F,0 CEMENTS S Ik"I b ? M k kA 10f 0f kEVih b TO ED'iS , 2. REMOTE / RELIABLE CONTROL OF VENT VALVE NUMARC CONTACTED TO CONSIDER SEVERE ACCIDENT
^bbdTbhkAffGENERICINDUSTRYISSUE t
CONTINUE WORKING WITH NRC TO BETTER DEFINE !$50F5 IDCOR/BWROG EVALUATIONS: PUK HtSOLUTION o CONCEPTS CONSIDERED PROPOSE TO BWROG SEVERE ACCIDENT INSIGHT o UPGRADERBUCTINOFA89GSTMI)BX GAS TREATMENT SYSTEM (SBGTS) TO CONTAINMENT DESIGN PRESSURE CAPABILITY
- NOT FEASIBLE TO UPGPADE SBGTS i
o HARDPIPED BYPASS AROUND SBGTS o HARDPIPED DEDICATED VENT o COSTS ARE C CiPARABLE FOR HARDPIPED CPTIWS
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l ELEMENT 3 - PRESSURE (Continued) ! l L i o VENT SIZING UNDER REVIEW .- :/ C r {', e d p f x o ATWS 6 cfg o DHR o . NEGATIVE IMPACTS NEEDING FURTHER REVIEW o SECONDARY CONTAINMENT CONTAMINATION o DELIBERATE RELEASE l L e- - - .- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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l ELEMENT 4 - C ME DEBRIS ( OBJECTIVE: REDUCE LIKELIHOOD E FAIWRE BY DIRECT ATTACK REQUIREMENTS:
- 1. USE PRACTICAL DEBRIS RETARDING BARRIERS
- 2. CWSERVE SUPPRESSIW P0OL WATER AS A QUENCHING POOL LDCCR/BWRG EVALUATIONS:
o DRYWELL o CONCE' PTS CONSIDERED PLUG IN PEDESTAL OPENING INCREASE SUMP SIZE INSIDE PEDESTA L 1 l - CURB OJTSIDE PEDESTA L OPENING CURB AT DRYWELL LINER / FLOOR JUNCTION ADDITIGAL PEDESTA L OPENINGS TO PROMOTE EVEN DISTRIBUTION o NEGATIVE IMPACTS LOCA CWSIDERATIONS SEISMIC INTERACTIWS 1
- ALARA CmCERNS
- s@MMWrun ~.3,1:. . ?.h2E%mLJW N::ab.::.: a:,a n :::.r ' 1.C ::7:si c.: '
l ELEMENT f4 - CORE DEBRIS (Continued) 4 e QUALITATIVE BEllEFIT LOW DEPENDENT ON ANALYTICAL MODELS OF DEBRIS MOBILITY WHICH ARE VERY UIEERTAIN o WETWELL 0 MCST PLANTS CURRENTLY HAVE CAPABILITIES TO HOLD WATER IN TORUS CHAMBER OR PROVIDE PROTECTIOi.TO CRITICAL EQUIPMENT Ill C WNER R00iS o QUALITATIVE BENEFITS LOW o NO NEGATIVE IMPACTS o NO FURTHER STUDY WARRANTED l l 1
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@JECTIVE: ENSURE OPEPATCRS ARE READY TO USE PLANT FEATURES TO BEST ADVANTAGE IN SEVERE ACCIDENTS REQUIREMENTS:
- 1. C LEAR SYMPTm BASED STRATEGIES (INTEGRATED)
- 2. REMOVAL OF UNNECESSARY INHIBITIWS
- 3. TRAINING /PR[EDURES IDCOR/BWROG EVALUATIONS:
o REV. 4 IMPLEMENTATION BY ALL UTILITIES C WSISTENT WITH PREVI OJS PGT-TMI COMMITMENT o PROPGE REVIEW 0F REV. 4 WITH INSIGHTS FR01 SEVERE ACCIDENT STUDIES e . _ _ . . . _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _
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, CONCLUSIONS - BWR EXECUTIVE MEETING o AUGUST 19 MEETING - 20 0F 23 BWROG UTILITIES REPRESENTED o CONTINUE CONSTRUCTIVE DIALOGUE WITil NRC o AGREEMENTS:
COMMIT T0 IMPLEMENTATION OF REVISION 4 TO EPGS NUMARC CONTACTED TO CONSIDER SEVERE ACCIDENT CONTAINMENT ISSUE AS A GENERIC INDUSTRY ISSUE CONTINUE WORKING WITH NRC TO BETTER DEFINE ISSUES FOR RESOLUTION PROPOSE TO BWROG SEVERE ACCIDENT INSIGHT REVIEW 0F EPG REV 4 1 l' l l
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OFFICE MEMORANDUM [6102584
, Boston Edison Company RMG Control Number
%vu To:(C. S. BrennfoiP) From: R. N. Swanson
- Record Type A4.08 Date: August 27, 1986 Dept. Doc. NED86- 649 Non-Safety Related
Subject:
Lead Engineer Assignment for Primarv Containment _ Issues _ Distribution: JE Howard PT Antonopoulos TJ Tracy RV Fairbank J Pawlak RE Grazio JL Rogers S Dasgupta JM Fulton JD Keyes DA Bryant WS Clancy PT Kahler R Velez GV Mileris CH Minott VJ Zukauskas T. Hauske You are hereby assigned as the Nuclear M'weiing Department's lead engineer for all primary containment related Lsues, *n this capacity you are to:
- 1) Monitor relevant industry (NRC, AIF, BWROG etc.) efforts and maintain a complete file of relevant infonnation.
- 2) Evaluate and apprise me of the impact / potential impact on PNPS of any primary containment related initiatives or concerns.
- 3) Perform, or have performed, all plant specific technical evaluations or analyses related to the issue.
- 4) Assume Lead Engineer responsibilities as described in NED Procedure 1.01 for implementation of any Mark I Containment related modifications.
By copy of this memo those on distribution are requested to support Steve in this assignment, include him on distribution, as applicable, and 7.dvise him of relevant industry meetings or events. DOCUWLNT ELEASED A(.$ 2 9 G86 i l FOR USE S t&TI ON
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QUESTION 9. NRC's chief safety officer has reportedly e told the industry that he is concerned about
-k5ctw (, rag sunus Mark I containment buildings because given a
-)3} q{ . ,
core melt accident they have a 90 percent chance of failure. Please answer th'e : following: - (A) Is a 90 percent chance of failure in the event of a core meltdown an acceptable failure rate? e ANSWER. - The NRC holds the position that the likelihood of core melt ac:idents in any plant should be very low and, in addition, that there should be substantial assurance that the containment will , mitigate the consequences of a core melt should one occur in order to ensure low risk to the public. It is not merely a questionofhavinglowriskbutofhavingasweY1the - defense-in-depth assurance of combined protection by prevention and mitigation. This concept was identified in the Commission's Policy Statement as noted in response to Question 8. However, a number of regulatory actions have been taken, as outlined below, which will reduce the probability of containment failure for Mark I containment buildings.
~ _ _ . _ - . .-
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QUESTION 9. (Continued The recent expression of concern cited in Question 9 was expressed with the intention of encouraging utility owners of BWR MK I plants to give priority to the continuing need for ensuring MK I containment integrity. The 90 percent chance estimate was a rough approximation of the results of the WASH-1400 assessment published in 1975. That assessment was based on the Peach Bottom plant as it stood at that time, and on the operating and emergency procedures of that time. The results, which indicated a virtual certainty of a fairly large release with every core melt, also showed many circumstances where lesser accident conditions could lead to overpressure failure of containment --
- which failure in turn actually caused the core melt. The risk -
dominant accident sequences were transients such as anticipated transients without scram (ATWS) in which excessive energy is released to containment or other transients with an associated scram where containment heat removal capability was lost. It is important to note that the WASH-1400 analysis still found that the overall risk of the BWR was equivalent to the risk of the PWR even though the PWR containment was estimated to proyide greater ., mitigation of core melt consequences. This was because WASH-1400 estimated a lower likelihood of core melt in the BWR due to its dtverse and flexible water supply systems. l l L - _ ---. .
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QUESTION 9. (Continued Since the TMI accident a great deal has been done and is still in progress to lower the risk of the BWR MK I plants, both by . suppressing the likelihood of core melt accidents and by ensuring consequence mitigation by the containment. Improved training and l symptom-based procedures for plant operators have been a'dopted. l These are especially effective in BWRs bec'ause of the inherent flexibility of the plant systems. One of the major risk contributors ATWS, has been substantially reduced by NRC rule mandated improvements. Since further studies showed that the BWR suppression pool water had a much greater capability to scrub releases than was recognized by WASH-1400, a filtered vent containment strategy was adopted. While within the design basir envelope, the BWR containment remains sealed. If, through some beyond design basis failure, the pressure threatens to cause uncontrolled failure of containment, the strategy calls for venting the containment from the wetwell, above the pool, so that releases pass through the water where essentially all the major constituents.except the noble gases are removed. Containment venting procedures are being developed and imple'mented at many - BWRs. Some small design changes may be necessary to implement or improve the ability to avert containment failure from overpressure or other failure modes.
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i l _ QUESTION _9. (Continued - 4-Both the NRC and the indu;try are active in this process through our independent work and interaction with the Industry Degraded Core (IDCOR) group and the owners directly. Because.of the continuing work and the incomplete implementation, it is not possible to give a representative MK I containment i performance figure at this time but, consiste.t with the NRC Severe Accident Policy Statement, the current process for evaluating and improving existing plants is expected to achieve I conditions where a HK I c.ontainment provides substantial i l assurance of mitigating the consequences of a core melt should one occur. Further, with adoption also of the many ways to - reduce the likelihood of core melt, the risk of BWR MK I plants is low. i
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i.wna nummsda:;on W:::%D;e5 .:v '. :J.icca:h;.w wL:0;Y ;cbVm. m -- QUESTION 9 (B) Does the NRC believe that Pilgrim's containment building is more vulnerable to failure given a core melt accident than other s types of containment buildings in use at
, other plan.ts? -
ANSWER Studies of severe accidents for many years have indicated that BWR MK I containments such as the Pilgrim plant has are more vulnerable to failure given a core melt than some other types of containment buildings due to smaller relative size, compact - geometry, etc. Nevertheless, the work referred to in response to 9A above indicates that substantial assurance of core melt consequence mitigation can be achieved with the BWR MK I containment. As to Pilgrim specifically they, like other plants of this type.- are in the process of developing and implementing the necessary improvements. For example, emergency operating procedures up through Revision 2 of the BWR Emergency Procedure Guidelines have been implemented; this includes containment venting. Two of the three elements of the ATWS improvements have been made. But the Individual Plant Evaluation for severe accidents has not yet been conducted so no specific risk estimate can be made at this time for the Pilgrim plant. - m-- -_-:---__= - _,
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a:,r. xp.v;;mr.i.:.znwu;m.u.mun:c:adeu.. ; '.Fr.: ':;n 2 u. ivs ic m; e ,i :.,:-- * .;w . l l QUESTION 9: (C) Rank the 17 plants with management problems in descending order of the estimated containment
. failure rate."
ANSWER. It is not feasible to do such a ranking on a plant by plant basis men-4ngfully due to the uncertainties inherent in current predictions of containment behavior during severe accidents, and in
,. -plant specific design differences.
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x w : m ..,.: w r a u.e x .w .: scu = ,::.n a .>. a.-u, x:.w. .o m . u.aa,:w.m era: . . .. i . ,a q l RUESTION 9. (D) Is the NRC considering any new requirements or backfits relating to containment issues?
?
. 1
. i ANSWER. .
As indicated in the answers to parts a through b of Question 9 the NRC expects to consider new requirements and backfits in the implementation of its severe accident policy. 7 - i
- e 6 + e
% E T70I,2%.IIld Min:_'d d KiWF53id58d?).222M'B;.52221snFEA F.?! M 4.t $ 4 .~r.dU h d 6 ? MEETING
SUMMARY
DISTRIBUTION Licensee: Vermont Yankee Nuclear Power Company
-* Copies also sent to those people on service-(cc) list for subject plant (s).
Docket File NRC POR L POR BWD-2 Rdg DMuller VRooney OGC-MNBB 9604 EJordan BGrimes ACRS-10 HAbelson. RAuluck DShum TElsasser PLeech MThadani JKopeck RHouston MHodges Glainas RBernero i I
D M ii 2 LH 2 2kih6 D g.D M %: M ?d % M U W .:; X u.x M Z % kW S .1 7-~ 4 5 G r 1..... & . Septembsr 9, 1986-Docket No. 50-271-i LICENSEE: Vermont Yankee Nuclear Power Corporation FACILITY: Vermont Yankee Nuclear Power Station
SUBJECT:
AUGUST 6, 1986 MEETING WITH THE VERMONT YANKEE NUCLEAR POWER CORPORATION (VYNPC) Re: Containment safety study status On August 6, 1986, a meeting was held at the NRC headquarters in Bethesda, Maryland to discuss the status of the Vermont Yankee Mark I containment safety study. Enclosure 1 is a list of individuals that attended the meeting. Enclosure 2 is a handout of the slides presented by VPNPC at the meeting. VYNPC reported that the Vermont Yankee plant had been compared to Mark I containment design assumed for generic studies with respect to: hydrogen control, drywell sprays, containment pressure control, core debris control, and training and procedures. These comparisons were almost complete, and at the time of the meeting, inputs by consultants were being condensed and integrated into a single report. The report should be complete at the beginning of September, as scheduled. C.T. m
. ,, .c.'.v.v!
. . . '. y Vernon L. Rooney, Project Manager BWR Project Directorate #2 Division of BWR Licensing
Enclosures:
As stated cc w/ enclosures: See next page BWR#2 VRooney j/f/86
?dTMDuM.%2MMMn:n u.%:OWinWlt GW2 2:i ' MhOa:.,rk : n':L. &FW:6 ~ .,..w 1 i Mr. R. W. Capstick Vemont Yankee Nuclear Power Corporation Vermont Yankee Nuclear Power Station - cc: Mr. J. G. Weigand W. P. Murphy, Vice President & President & Chief Executive Officer Manager of Operations Vermont Yankee Nticlear. Power Corp. Vermont Yankee Nuclear Power Corp. R. D. 5, Pox 369 R. D. 5, Box 169 Ferry Road Ferry Road Brattleboro Vement 05301 Brattleboro, Yemont 05301 Mr. Donald Hunter Vice President Mr. Gerald Tarrant Commissioner Vermont Yankee Nuclear Power Corp. Vermont Department of Public Service 1671 Worcester Road 120 State Street Framingham, Massachusetts 01701 Montpelier, Vermont 05602 New England Coalition en Nuclear Pollution Hill and Dale Fam Public Service Board R. D. 2. Box 223 State of Vermont Putney, Vemont 05346 120 State Street Mr. Walter Zaluzny Chairman, Board of Selectman Vermont Yankee Decomissionirg Post Office Box 116 Alliance Vernon, Vermont 05345 Box 53 Montpelier, Yement 05602-0053 J. P. Pelletier, Plant Manager , Vemont Yankee Nuclear Power Corp. Post Office Box 157 Resident Inspector Vernon, Yement 05354 U. 5. Nuclear Regulatory Corrission Post Office Box 176 Raymond N. McCandless Vernon, Yement 05354 Yemont Division of Occupational
& Radiological Health Yemont Public Interest Administration Building Research Group, Inc.
10 Baldwin Street 43 State Street Montpelier, Vemont 05602 Montpelier, Vermont 05602 Honorable ,1ohn J. Easton Thomas A. Murley Attorney General Regional Administrator State of Vermont Region I Office 109 State Street U. 5. t!uclear Regulatory Corrission Montpelier, Vement 05602 631 Park Avenue King of Prussia, Pennsylvania Ic@6 John A. P.itscher, Esquire Ropes & Gray 225 Franklin Street . Boston, Massachusetts 02110
m m m wc;;,.wz:Pw Ln.:.xo . 50w:w :.ta.w.. ::s x : :::;." r ;a wm c. . m . u :zu . e. Enclosure 1 LIST OF MEETING ATTENDEES August 6 -1986 l Name Organization H. Abelson NRC/ DBL /PD#? D. Huller NRC/ DBL /PD#2 J. Thayer Yankee Atomic Electric Co. R. Lodwick VYNPC W. Murphy VYNPC S. Shultz Yankee Atomic Electric Co. R. Auluck NRC/ DBL /PD#1 P. Paull State of Vermont G. Tarrant State of Vemont S. Murphy NIRS D. Shum NRC/ DBL /F0B T. Elasser NRC/RI P. Leech NRC/ DBL /PD#1 M. Thadani NRC/ DBL /PD#2 R. Bachmann NRC/0GC J. Gray, Jr. New York Power Authority J. Kopeck NRC/PA J. Coates Vermont Public Interest Research Group D. Hew Harmon & Weiss R. Houston NRC/ DBL M. Hodges NRC/ DBL /RSB G. Lainas NRC/ DBL /BWAD R. Bernero NRC/ DBL O A _i__ _ . - . - . ,_ * -
NXEEKi?@KWCURCEMU iO2HN;iFRWZMW M"En212WSC':Mnr@ Er l- '- 0 - Enc 1osure.2 VERMONT YANKEE CONTAINMENT SAFETY STUDY- j
- 1. PURPOSE AND OBJECTIVE OF EACH TASK INCLUDING PRELIMINARY RESULTS MARK 1 DESIGN REVIEW o WASH 1400 DESIGN COMPARIS0N o KEY DESIGN DIFFERENCES o VERMONT YANKEE DESIGN
SUMMARY
VERMONT YANKEE CONTAINMENT CAPABILITY o VERMONT YANKEE APPROACH ; o STUDIES AND RESOURCES USED o ACCIDENTS AND TRANSIENTS STUDIED o RESULTS UTILIZING KEY DIFFERENCES AT VERMONT YANKEE o DEFINITION OF "90%" o VY CONTAINMENT CONDITIONAL FAILURE PROBABILITY CURRENT MARK I ISSUES o DEFINITION AND TECHNICAL STATUS OF 5 ISSUES o APPLICABILITY AND NEED FOR FURTHER STUDY o ACTIVE INDUSTRY AND NRC EFFOTS I. STATUS
- RESULTS TO DATE
- SCHEWLE FOR COMPLETION )
)
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CONTAINMENT SAFETY STUDY ) DESIGN AND OPERATIONAL FEATURES COMPARISON ppsir.n/ Operational Data Vemont Yankee (WASH-1400) Q neral Plant Data Plant Type CE BWR GE BWR Containment Type Mark I Pressure Suppression Mark I Pressure Suppression , R2ted Thermal Power, MWt 1,593 3,293 1 Rnted Core Flow, lb/hr 48.0 x 10E6 102.5 x 10E6 R;ted Steam Flow, Ib/hr 6.43 x 10E6 13.381 x 10E6 Reactor Data Incide Height, ft-in 63 - 1.5 72 - 11 Inside Diameter, in. 205 251 Containment Data Internal Design Pressure, psis 56 56 Drywell Data cylinder Diameter, ft. 33 38.5 ft. 62 67 Spherical Diameter,3 Free Air volume, ft 134,000 159.000 Torus Data Major Diameter, ft. 98 111.5 Minor Diameter, ft. 27.66 31 W:ter Volume, min / max., ft3 68,000/70,000 123,000 Free Air Volume, ft3 114,200/112,200 119,000 Vent Pipes Data Number 8 8 Internal Diameter, ft. 6.75 6.75 Downcomer Pipe Data j Number 96 96 Internal Diameter, ft. 2 2 Submergence, ft. (nominal) 4 4 Number of WW-DW Vacuum 10 12 Breakers l l _ __ _ _ _ - _ _ l
[ESUSf53$iO.. 'TlO2f2OIdidOb6!_TSE22Eit iMd5Ef 7 % Y fi716%ODY, eM5, 4-Design / Operational Data Verwont Yankee (WASH-1400)
- Secondary Containment Data Fr:e Air Volume 2,120,000 2,400,000-Mitimation Systems Deslan Data HPCI System Number of Trains or Subsystems 1 1 Number of Pumps / Train 1 1 Design Flow / Train 4,250 gpm at 1,120 to 150 psid 5,000 spm at 1,100 to 150 psig Electrical Power DC only (turbine-driven) DC only (turbine-driven)
RCIC System Number of Trains or Subsystems 1 1 Number of Pumps / Train 1 1 Design Flow / Train 400 600 spm Electrical Power DC only (turbine-driven) DC only (turbine-driven) RHR System Number of Trains or Subsystems 2 2 Number of Pumps / Train 2 2
- Number of KKs/ Train 1 2 Design Flow / Train 14,400 gPm 20,000 gym Electrical Power Emergency ac and de Emergency ac and de Source of Water Torus Torus Service water (river) Service water Ultimate backup (Diesel fire pump !
cross-tie capability) Emergency Diesel Generator Systems Number of Emergency Buses 2 4 per unit l Number of Emergency Diesels 2* 4* - Shared between two units
- Single-Unit Site - Requires *Two-Unit Site - Requires two one out of two emergency out of four emergency diesels diesels for safe shutdown. for safe shutdown.
__..____1___ __ _ _
.. ,w a.; w w m a m . w w c u m r z a :1 we. w . w m ::.u x ie :.it w ;g r w +. w a m w a m x+ x ---- e l VERMONT YANKEE - WASH 1400 KEY DESIGN DIFFERENCES
. RAfl0
- DESIGN PARAMETER VERMONT YANKEE / WASH 1400 RATED THERMAL POWER .48 DRYWELL VOLUME / POWER 1.75 TORUS WATER VOLUME / POWER 1.19 TORUS AIR VOLUME / POWER 1.96 HPCI PUMPING CAPACITY / POWER 1.77 RCIC PUMPING CAPACITY / POWER 1.40 j
l; _.'_i___.s._.___,u_ ._..2_ s .
w 4 du:GB%Mu %.u?C22nudWcwee;;1MW1;.GMVn%' ic9mcsus"'w J-' VERMONT YANKEE DESIGN
SUMMARY
o SMALL REACTOR PLANT (NSSS) IN LARGE MARK 1 CONTAINMENT , o ENGINEERED SAFETY FEATURES CAPACITY o RESIDUAL HEAT REMOVAL CAPABILITY o ELECTRIC DRIVEN MAIN FEED PUMPS o DC SYSTEM CAPACITY & DIVERSITY
- 8 HOUR BATTERY RATING
- APPENDIX R BATTERIES
- SPECIAL PURPOSE BATTERIES
- UNINTERRUPTIBLE POWER SUPPLIES o AC SYSTEM RELIABILITY & DIVERSITY
- VERNON HYDRO TIE LINE
- NORTHEAST GRID RELIABILITY
- SEPARATE HIGH LINE RIGHT OF WAYS o DIESEL FIRE PUMP CROSSTIE TO RHR SYSTEM
pwamw;wurmutw.i.sw=+.mmaea:n. :.cw m.wss,. - ~m L.: n w m r x :x-l , 9 1 STUDY APPROACH-o IDENTIFY THE DOMINANT ACCIDENT SEQUENCES WHICH CAN LEAD TO SEVERE ( ACCIDENTS AT VERMONT YANKEE. ! 1 o QUANTIFY THE DOMINANT SEQUENCES USING A REFERENCE MARK 1 MODIFIED FOR THE UNIQUE VERMONT YANKEE FEATURES. o DEVELOP THE CONTAINMENT EVENT TREE TO DISPLAY THE PATHWAYS TO SAFE MITIGATION AND POTENTIAL RADIONUCLIDES RELEASE TO THE ENVIR?NMENT. o QUANTIFY THE CONTAINMENT EVENT TREE USING AVAILABLE ESTIMATES OF MITIGATION RELIABILITY AND STANDARD MODFLS. o CALCULATE THE CONDITIONAL FAILURE PR08 ABILITY OF CONTAINMENT. l i 1 ( i
KERwx514:k:22 % B x c a m n,s s a : m n. c : m a w a.. m w s :v.=.-.- m ; ;: a .: a ruu w w . M IIElES& RESOURCES STUDIES o WASH 1400 o IDCOR
- IPE METHODOLOGY
- TECHNICAL
SUMMARY
REPORT
- TASK REPORT DOCUMENTS o DRAFT BWR MARK 1 PSA's
- o SEVERE ACCIDENT SEQUENCE ANALYSIS PROGRAM
_ RESOURCES o YANKEE ATOMIC ELECTP,1C COMPANY o DELIANCORPORATION L 1 o FAUSKE AND ASSOCIATES o RISK MANAGEMENT ASSOCIATES o GENERAL ELECTRIC COMPANY o PICKARD LOWE & GARRICK, INC.
...c w g.> m = an . m.z.:.a...w a.n..vm = n.a a m m .?eu w rr.w .= w :1. x.y,w ~ w ,a.-n:=7: n m;1 a m
- ] !
. , d 1 P f
BWR DOMINANT ACCIDENT SEQUENCES CLASS I- LOSS OF COOLANT MAKEUP IA HIGH PRESSURE IB STATION BLACK 0UT-IC ATWS
-ID LOW PRESSURE IE LOSS OF DC CLASS'll LOSSOFCONTAINMENTHEATREM0hAL
- - CLASS III LOCA CLASS IV ATWS i
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iD Rimux-itMM&:12 mac:2?. c13sr=usmamism;w: . mixm. uahwm u mm: n;us; :e. , VERMONT YANKEE RESULTS
' DOMINANT SE00ENCES STATION BLACK 0UT KEY DIFFERENCES:
o DIESEL FIRE PUMP CROSSTIE o VERNON HYDR 0 TIE LINE o DC SYSTEM CAPABILITY o DIESEL GENERATOR RELIABILITY o NEW ENGLAND GRID $TABILITY ATEi KEY DIFFERENCES: o CONTAINMENT SIZE VS. POWER o ET! STING ATWS MODIFICATIONS AND COMMITMENTS o LARGE RHR SYSTEM HEAT REMOVAL CAPACITY o MSIV RE0 PEN CAPABILITY o EMERGENCY OPERATING PROCEDURES L
p%m3Rstaa pi:.GTwa,!Diw G:bmwnumauwi:?.:rmu.r.b d=.;:W!im:'n, M:: :W l' L. DEFINITION OF "90%" WASH-1400 ASSUMPTIONS o ALL SEQUENCES (MELT /NON-MELT) ARE ASSOCIATED WITH CONTAIN".ENT FAILURE o CORE MELT RELEASE CATEGORIES
- 1. STEAM EXPLOSION
- 2. OVERPRESSURE FAILURE DIRECT TO ATMOSPHERE ,
- 3. OVERPRESSURE FAILURE THROUGH REACTOR BUILDING I4. ISOLATION FAILURE o RELEASE CATEGORIES 2 & 3 DOMINATE RISK o SEQUENCES WHICH CONTRIBUTE COMPRISE 90%
ATWS LOSS OF RESIDUAL CORE HEAT REMOVAL CURRENT STUDY o DIFFERENT DOMINANT SEQUENCES IDENTIFIED o KNOWLEDGE OF CONTAINMENT FAILURE PHENOMENA AND CRITERIA HAS EVOLVED 1 3 i 1 1 1
- _ = _ _ - - _ _ . _ _ _ _ _ _ _ _ _ _ _ _
l
I!Z5m..esmMEMFf5?s:Ef%~2mKWSea.;a.cdmetsrssd2iese "aem4Fe my,;2 ' i i l CONTAINMENT CONDITIONAL FAILURE PROBABILITY I o CALCULATE THE CONTRIBUTION OF EACH ACCIDENI CLASS 10 10TAL LIKELlH00D OF CORE MELT l l o DElERMINE THE CONDITIONAL FAILURE PROBABILITY OF CONTAINMENT FOR EACH ACCIDENT CLASS o CALCULATE THE WEIGHTED AVERAGE OF PRIMARY CONTAINMENT FAILURE PROBABILITY
%EfMA:2?,l,lGON di 7> bBJi3 :r134BC2L"D2Sn-v;;U:6N:6l*TL:Tb 7%Li di%R2'W"lN5%% & 5 '
4 CONDITIONAL CONTAINMENT FAILURE PROBABILITY CUANTIFICA110N DOMINANT FRACTIONAL CONTAINMENT CONTRIBUTION AcclDENT CONTRIBUTION FAILURE TO WEIGHTED SEQUENCE' (CORE MELT) PROBABILITY AVERAGE LOSS OF C00LAN1 MAKEUP A A .A x A LOSS OF CONTAINMENT HEAT REMOVAL s B sxB LOCA c C cxC ATWS o D oxD n 1.0 E( ) = CFFP i O WHERE CCFP = CONDITIONAL CONTAINMENT FAILURE PROBABILITY i l
p a m:a m m a x as a w m . m a w. a m wa:ee m u aa.x ra m e n = m:. O ?.
- l CURRENT TECHNICAL ISSUES
- 1. HYDR 0 GEN CONTROL
- 2. DRYWELL SPRAYS.
- 3. CONTAINMENT PRESSURE CONTROL
- 4. CORE DEBRIS CONTROL
. 5. TRAINING AND PROCEDURES i
1 1 l 1
-l i
i
wn::.:sxxznw;xnxarwawwwunemanau.~ue.a:w::mm.=ww. mw~ .c ?.v. :'. RYDR0GENCONTROL
- OBJECTIVE: PREVENT HYDR 0 GEN COMBUSTION CAUSED FAILURE e
SUGGESTIONS: 0XYGEN CONTROL INERT TO START CONTROL INGRESS OF OXYGEN ISSUES IDENTIFIED BY NRC: WHEN AND HOW LONG NOT INERTED PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS
- 1. -CONTAINMENT.INERTED 2.- ELECTRIC POWER NOT REQUIRED TO MAINTAIN INERT
- 3. TECH. SPECS. CONTROL DEINERT TIME ,
11 . PLANT SHUTDOWN IF TECH. SPEC. CANNOT BE MET e vn i. F =I % a __ _.-_______2.________i
j.a : max =wsananm;awx:anxun:xx::mna=w.cm.; .z:w.ox.u m.uwiw=w.=a=r a i i DRYWELLSPRAlS. OBJECTIVE: SPRAY WATER T0: i
.1. LOWER PRESSURE j
- 2. COOL VULNERABLE EQUIPMENT
- 3. QUENCH DEBRIS
- 4. SCRUB AEROSOLS SUGGESTIONS:
- 1. SPRAY IN DRYWELL
- 2. INDEPENDENT BACKUP WATER SOURCES AND PUMPS H0SE CONNECTIONS USE OF FIREMAINS ISSUES IDENTIFIED BY NRC:
- 1. RISK 0F IMPLOSION
- 2. RISK 0F HYDR 0 GEN COMBUSTION AFTER STEAM CONDENSATION
- 3. MANUAL ACTIONS AND TIMING !
i PRESENT VY CAPABILITY RELATIVE TO PROPOSED RE0VIREMENTS: I I
- 1. EXISTING FLOW PATH FROM DIESEL FIRE PUMPS TO VESSEL OR DRYWELL SPRAYS (2500 GPM)
- 2. AVAILABLE FOLLOWING STATION BLACK 0UT
- 3. IF EMERG DIESEL AVAILABLE, SYSTEM OPERABLE FROM OUTSIDE REACTOR BUILDING
- 4. EXISTING OPERATING PROCEDURE DESCRIBES USE OF SYSTEM i
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Tar::auwan:-a~ n s.:w::mst:w mamnsawmasmw.w:c u.12:u,x nsiam CONTAINMENT - PRESSURE CONTROL OBJECTIVES: 1. AVERT UNCONTROLLED OVERPRESSURE FAILURE
- 2. CONTROL RELEASE PATH (SCRUBBING) I SUGGESTIONS: ,
- 1. SUBSTANTIAL CAPABILITY TO VENT WETWELL
- 2. REMOTE / RELIABLE CONTROL OF VENT VALVE
- 3. ABILITY TO RECLOSE VENT ISSUES IDENTIFIED BY NRC:
- 1. DELIBERATE RELEASE OF RADI0 ACTIVITY
- 2. WHAT CONSTITUTES REMOTE / RELIABLE CONTROL?
- 3. IS VENTING TO SECONDARY CONTAINMENT ACCEPTABLE?
- 4. WHAT IS APPROPRIATE ACTION PRESSURE?
PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS:
- 1. EIGHTEEN INCH ATMOSPHERIC CONTROL SYSTEM. VENTS
- 2. THREE INCH ATMOSPHERIC CONTROL SYSTEM VENT
- 3. TWENTY AND EIGHTEEN INCH NITROGEN PURGE LINES
- 4. SIX INCH NITROGEN PURGE LINES l S. ONE INCH NITROGEN CAD LINE - EXISTING PROCEDURE w- -- __ . .
p.m.. mana4cc.wa. ' wm2m22:eerwar.xm =arr.3,mre. .e...cuv.22anc.;;wmr 4 9 CORE DEBRIS CONTROL OBJECTIVE: ' REDUCE LIKEllH00D OF FAILURE BY DIRECT CONTACT OF CORE DEBRIS WITH DRYWELL WALL. l l SUGGESTIONS:
- 1. USE PRACTICAL DEBRIS RETARDING BARRIERS
- 2. CONSERVE SUPPRESSION P0OL WATER AS A QUENCHING P0OL ISSUE IDENTIFIED BY NRC: WHAT IS PRACTICAL?
PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS:
- 1. SMALL CORE DEBRIS VOLUME COMPARED TO PREVIOUS STUDIES
- 2. ,
DRYWELL SUMPS COMBINED VOLUME APPR0XIMATELY 200 FT3
- 3. >1200 FT DRYWELL FLOOR SURFACE AREA
- 4. DOWNCOMERS APPR0XIMATELY ONE FOOT AB0VE DRYWELL FLOOR
a m ia.Z E ais e .w s . m i m s n/c.h w m u s. s e v a: e m s w c a m . w . ,. m m u .s. w.c.-:e m m. - o 1 4 TRAINING AND PROCEDURES j OBJECTIVE: ENSURE OPERATORS ARE READY T0'USE PLANT FEATURES TO BEST ADVANTAGE IN SEVERE ACCIDENTS SUGGESTIONS:
- 1. CLEAR SYMPT 0M BASED STRATEGIES (INTEGPATED)
- 2. REMOVAL 0F UNNECESSARY INHIBITIONS
- 3. TRAINING ISSUES IDENTIFIED BY NRC:
- 1. COMPETING SAFETY REQUIREMENTS
- 2. DEGREE OF TRAINING PRESENT VY CAPABILITY RELATIVE TO PROPOSED REQUIREMENTS:
- 1. REV. 3 0F EMERGENCY PROCEDURE GUIDELINES (EPG's) IMPLEMENTED
- 2. ACTIVE PARTICIPATION IN REV. 4 DEVELOPMENT
- 3. VY PLANT SPECIFIC SIMULATOR COMPLETED IN 1985
- 4. EXTENSIVE OPERATOR TRAINING PROGRAM INCLUDES SEVERE ACCIDENT PREVENTION AND MITIGATION
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