IR 05000293/1986030

Mgt Meeting Insp Rept 50-293/86-30 on 860730.Major Areas Inspected:Plant Status,Salp Improvement Status,Overtime Control,Startup Checklist,Mark I Containment Enhancement & Other Issues.Related Info Encl

ML20211C040
Person / Time
Site:	Pilgrim
Issue date:	08/19/1986
From:	Mcbride M, Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20210F497	List: IR 05000293/1986030 ML20210F489 ML20211C003 ML20211C010 ML20211C025 ML20211C047 ML20211K596
References
FOIA-87-10 50-293-86-30-MM, NUDOCS 8702190675
Download: ML20211C040 (42)

Text

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. U. S. f,'UCLEAR REGULATORY COMMISSION

REGION I

Yieting N /86-30

::cket N License N OPR-35 Licensee: Boston Edison Company M/C Nuclear 800 Boylston Street Boston, Massachusetts Cacility Name: Pilgrim Nuclear Power Station Meeting At: USNRC, Region I, King of Prussia, PA Feeting Conducted: July 30 1986 Frepared by:,, _ d  -b2' d__   _

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 - w . McBride, Senior Resident inspector   Date i

h proved by: - ML C SfdkI-

 -. Strosnider, Chief, Projects Section  Date
 , ORP Peeting Summary: A management meeting was held at NRC Region I on July 30,
'986 to discuss plant status, SALP improvement status, overtime control, startup checklist, Mark I Containment enhancement and other issues. The reeting lasted about two hours. Members of the news tredia attended the eetin B70;!19067L G70213 PDR FDIA 1YEB7-010 PDR     i

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!            DETAIL _5 1.0 Purpose

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The management conference was held at the request of NRC Region The j licensee was asked to discuss recent program i provements and other topics of

'    interest at the plant. The licensee announced prior to the meeting, that
'    the current outage would be extended until early 198 I   2.0 Meeting Summary

Dr. T. Murley, NRC Region I Administrator, opened the meeting with a brief ' discussion of recent regulatory actions, including the Augmented Inspec-tion Team, Confirmatory Action Letter (CAL) 86-10, and the agreement to seek Region I approval prior to starting the plant. Mr. A. L. Oxsen, Vice

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President of Nuclear Operations, opened Boston Edison's presentation by

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" I discussing recent licensee program changes, including a management realign-ment under an Executive Vice President and the decision to enter into re-fueling outage number seven. Mr. Oxsen indicated that Boston Edison will i complete fire protection modifications require:: by 10 CFR 50, Appendix F

during this outage. Boston Edison also plans to work on further improv - t

ments in management control processes and implement certain Mark I contain- ! ment enhancements during the outage.

! Plant Status j I Current outage work was discussed, including inspections of the resi-I dual heat removal (RHR) pump impeller wear rings and ongoing contain- ! ment leakage testing. No visual indications have been detected in the wear rings of the two RHR pumps that had been disassembled. The l remaining two RHR pumps will be inspected in the near future. The , t licensee discussed bearing wear that had been detected in one of the pump motors and indicated that the pump impeller wear rings will be tested for hardness and dye penetrant checked.

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The status of local leak rate testing (LLRT) was reviewed. Problems with the leakage test procedure for the feedwater check valves that l - may have contributed to a falsely high as-found leakage rate were discussed. The licensee stated that the LLRT procedure would be l modified prior to as-lef t testing. The Itcensee's plans for inte- j ! grated leakage testing were also discussed.

l SALP l_mprovement Status Improvements in management programs that relate to SALP-fdentified

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l . weaknesses were discussed, including the work control process at the  ;' ! z station, the development of performance indicators, communication of problems within the licensee organization, and staff accountabilit ! l $ The licensee stated that full-time planners would be assigned to  ! ! . I

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each managerert section head, to help track and pric-iti:e section wor Periodic morning meetings between the Plant Manager, QA, and the section heads were also describe The licensee's preparations for this year's emergency planning exercise were discussed, The licensee indicated that several practice drills were planne c. Overtime Control During Startup The projected number of licensed operators available for the plant start up in 1987 were discussed. The licensee indicated that five shif ts will be manned and that an additioral senior licensed operator will be added to each watch. Boston Edisor agreed to discuss planned overtime controls during the next plant startup in more detail during the next NRC management meetin d. Startup Check List The preparation of a detailed check 1 ,t of items trat should be com-pleted prior to a plant startup was discussed. The licensee indicated th3t the startup will last several weeks and will contain hold points at particular power levels. Regional man..gement indicated that the checklist and the startup itself will be reviewed ir considerable detail and that Regional Administritive approval will be required prior to proceding past certain critical hold points to be identified later. The Regional Administrator indicated that be will not grant approval for a restart until af ter the NRC has reviewed the licensee's startup pla e. Mark I containment Enhancements The licensee discussed possible improvements for the Mark I contain-ment at Pilgrim. The changes will probably be made to enhance such features as containment oxygen control, containment venting, and con-tainment spray. The need to consider these changes carefully and avoid inadvertently degrading other aspects of containment was emphasized by regional management. The licensee agreed and indicated that the enhancements would receive a thorough review. The licensee also indicated that they would be working closely with Vermont Yankee, another plant that is reviewing the Mark I containment desig :

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f. Other Topics Other discussion topics included maintenance planning, the maintenance backlog, once per cycle surveillance test scheduling, outage a: tivi-ties, and security program problem The licensee indicated that efforts were underway to reduce the main-tenance backlog and implement a maintenance planning grou The li-censee does not expect the planning group to be fully developed until the end of the current outage (although planning procedures should be in place by this fall). The licensee stated that the number of out-standing maintenance requests has been decreasing, although the number increased during the recent Boston Edison strike. The licensee indicated that the number of open maintenance requests by itself was not a good indicator of the plant physical condition. On another main-tenance item, the licensee indicated that the deficiency stickers for the control rod position indication system in the control room could not be cleared without extensive under vessel work. The status of these repairs was questioned during a recent visit to the site by an NRC commissic a The licensee briefly discussed some of this recent surveillance test schedulic,9 problens, and indicated that the testing f requency for once per-cycie tists would be clarified with NRR. The NRC regional staf f discussed 'ne importance of carefully reviewing the surveil-lance frequency for each type of once per-cycle test and noted that certain safety-related electrical equipment, required to be cali-brated once per-cycle, had recently been found out of calibratio Security staf fing problems and the amount of inoperable security equipment (requiring compensatory measures) was discusse Boston Edison indicated that staffing levels for the contractor security force would be increased and overtime closely monitored by Boston Edison management in the future. Boston Edison agreed to furnish the Pilgrim Senior Resident Inspector with the expected completion date for engineering evaluations of the scope of security equipment improvements. Subsequently, the licensee indicated that the evalua-tions would be completed by August 18, 198 Boston Edison stated that a major decontamination effort would be made in the plant during the current refueling outage. The licensee expects to decontaminate areas with safety-related equipment in the reactor building quadrant rooms so that these areas can be toured without protective clothin . . -w .

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3.0 Concluding Statements At the end of the meeting, Dr. Murley requested that Boston Edison be prepared to discuss the following issues during the next management meeting:

-- Overtime control during the plant startup
-- The expected scope of startup testing
-- Contain ent leakage test results and fixes
--

Mark I containment enhancements

-- Updated outage schedule Dr. Murley also indicated that the NRC will need to review Boston Edison's start up plars and hold point ,. _ _ _ _ _ __- ..  - . _ - - _ _ _ _ _ _  - _ __ ._ _ _ _ - . -- _ . _ _ - - - _ _ _ _ _ -

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ATTACHMENT TO MEETING REPORT 50-293/86-30 MEETING ATTENDEES Boston Edison Company (BEco) J. Lydon Chief Operating Officer E. Howard Vice President - Engineering & QA A. Oxsen Vice President - Nuclear Operations A. Pedersen Plant Manager K. Roberts Director of Outage Management R. Swanson Nuclear Engineering Manager S. Hudson Operations Section Head C. Gustin Manager, Public Information U. S. Nuclear Regulatory Commission T. Murle, Regional Administrator R. Starostecki Director, Division of Reactor Projects S. Ecneter Director, Divisicn of Reactor Safety H. Kiste* Chief, Reactor Projects Branch #1 L. Bettenhau en Chief, Operations Branch J. Dure Chief, Engineering Branch J. Strosnider Chief, Reactor Projects Section IB M. Shanbaky Chief, Facilities Radiation Protection Section M. McBride Senior Resident Inspector, Pilgrim G. Nejfelt Resident Inspector, Hatch G. Smith Safeguards Specialist P. Leech Project Manager, NRR Press J. Gehorgam Associated Press J. Impemba Boston Herald

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VERM3f;T YAf;KEE

*-  CONTAINMENT SAFETY STUDY I INTRODUCTION
 . PERSPECTIVE SCOPE RESULTS SUMMARY 11 MARK 1 C0llTAINMENT DESIGN REVIEW 111 SEVERE ACCIDENT ANALYSIS AND QUANTIFICATION BWR DOMINANT SE0VENCES CONTAltST CONDITIO!;A'_ FAILURE PROBABI'_lTY o METHODOLOGY o ACCIDENT SE0VEriCE QUA!1TIFICAT10N o INITIATING EVENT FRE0VEliCIES o CORE MELT FRE0VENCIES o CONTAINMENT EVENT TREES o CONTAINMENT PERFORMANCE RESULTS IV MARK 1 CONTAINMENT POLICY ISSUES o HYDR 0 GEN CONTROL o DRYWELL SPRAY CAPABILITY o CONTAINMENT VENTING o CORE DEBRIS CONTROL o SEVERE ACCIDENT TRAINING PROCEDURE V DECISION ANALYSIS FOR RECOMMENDATIONS
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o RECOMMENDATIONS AND INSIGHTS

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a PERSPECTIVE e INITIATED - TO RESPOND T0 QUOTED STATEMENT THAT MARK 1 CONTAINMENTS HAVE A 90% CONTAINMENT CONDITIONAL FAILURE PROBABI kr< u OcAukc~f e AGREED UPON BY VERMONT YANKEE, THE NRC AND THE GOVERNOR OF THE STATE OF VEPMONT ON JUNE 30, 1986 e PR03RESS

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INITIATED JULY 1, 1986

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INTERIM STATUS MEETifd IN EETHESDA ON AUGUST 6, 1986

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FINAL REPORT C0rPLETE AND TRANSMITIED TO NRC AND STATE OF VERMONT ON SEPTEMBER 2, 1986

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a STUDY SCOPE e MARK I CONTAINMENT DESIGN REVIEW

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e QUANTIFY A VERMONT YANKEE SPECIFIC CONTAINMENT CONDITIONAL FAILURE PROBABILITY GIVEN A SCENARIO LEADING TO CORE MELT e ADDRESS MARK I CONTAINMENT POLICY ISSUES

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s STUDY RESULTS e VERMONT YANKEE CONDITIONAL CONTAINMENT FAILURE PROBABILITY OF 7% 4 e CONTAINMENT CAPABILITY IMPROVEMENTS IDENTIFIED

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INSIGHTS BASED ON SEVERE ACCIDENT SEQUENCES

i i a v,- , - - - -, ,,--e- . - - - , . a .v-, ------r-----r- -,,-----,-r-+v-t +,w r~ ,-,.m , w rw r-w --

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a MARK I CONTAINMENT DESIGN REVIEW e SCOPE: COMPARE VERMONT YANKEE WITH THE MARK I REFERENCE PLANT TO IDENTIFY DESIGN AND OPERATIONAL FEATURE DIFFERENCES SIGNIFICANT TO CONTAINMENT PERFORMANCE DURING SEVERE ACCIDENT SEQUENCES e RESULTS: la SIGNIFICANT DESIGN FEATURES WERE IDENTIFIED

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- RATIO 0F CONTAlfiMEf;T SIZE TO REACTOR POWER 28. #-
- ELECTRICAL PLANT DIVERSITY DIESEL GENERATORS TIE Life 10 ADJACEtil HYDROELECTRIC STAT 10fiS
- CON!ECTION FROM DIESEL FIRE PUMP TO LPCI/

CONTAINMENT SPRAY

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, CONTAINMENT FAILURE PROBABILITY EVALUATION

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e BACKGROUND

- WASH - 11400 (90 0  r
- DEFINIT 10fl 0F CCFP Q[,h A dC e VERMONT YANKEE ANALYSES
- APPROACH
- ACCIDENT SEQUENCE IDENTIFICAT10f1
- ACCIDENT SEQUEf1CE QUANTIFICATI0f1
- CONTAINMENT EVENT TREE DEVELOPf1ENT
- CET QUANTIFICAT10f1 e SUMMARY OF RESULTS

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Craphic Display of Containment Performance 10- 0 . /s i I l l l l l 1E-1 L l NORMAL OPERATIONS l i l

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      (SEVERE ACCIDENTS)

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DEFINITION OF "90%" WASH-1400 ASSUMPTIONS e ALL SEQUENCES (MELI /NON-MELT) ARE ASSOCIAlED WlIH CONTAINMENT FAILURE e CORE MELT RELEASE CATEGORIES STEAM EXPLOSION OVERPRESSURE FAILURE DIRECT TO ATMOSPHERE OVERPRESSURE FAILURE THROUGH REACTOR BUILDING ISOLAll0N FAILURE e RELEASE CATEGORIES 2 & 3 DOMINATE RISK e SEQUENCES WHICH CONTRIBUTE COMPRISE 93% ATWS LOSS OF RESIDUAL CORE HEAT REMOVAL CURRENT STUDY e DIFFERENT DOMINANT SEQUENCES IDENTIFlED e KNOWLEDGE OF CONTAINMENI FAILURE PHENOMENA AND CRITERIA HAS EVOLVED l

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, CONTAINMENT CONDITIONAL FAILURE PROBABILITY e CALCULATE THE CONTRIBUTION OF EACH ACCIDENT CLASS TO TOTAL LIKEllH00D OF CORE MELT e DETERMINE THE CONDITIONAL FAILURE PROBABILITY OF CONTAINMENT FOR EACH ACCIDENT CLASS e CALCULATE THE WEIGHTED AVERAGE OF PRIMARY CONTAINMENT FAILURE PROBABILITY

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, CONDill0NAL CONI AlNPEN1 FAllllRE Prk? Bill 1Y 00AN11 FICA 110N DOMINANT FRACTIONAL CONTAINME ACCIDENT CONTRIBUTION CONTRIBUTION FAILURE SEQUENCE IO WEIGHTED (CORE MELT) PRosasitiir AVERAGE LOSS OF COOLANT MAKEUP A A AxA LOSS OF CONTAINPENT HEAT REIOVAL s B ExB LOCA c C cxC ATWS o D DxD I( ) = CFF? WHERE CCFP = CONDil10NAL CONTAINPENT FAILURE PROEABILITY

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CALCULATION PROCESS FLOW CHART (PAatfd $ I DENEFY IDENTIFY 1)tE DEVELOP THE DOWlNANT ACCDENT SE0VENCE CONTAINWENT SEQUENCE TWES FREQUENCES EVENT TREE

  #  # FOR THE LEADING TO SEWRC ACCIDENT   CONTAINWENT OR CONDITIONS   RESPONSE W UN1QUE FEARlRES)

It CALCULATE THE COND:TIONAL QUANTlfY FAILURE THE I PROBAB:UTY CCT OF CONTA:NWENT

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SMM PHENOWENOLOGICAL O b"O WITICADON CALCULADONS WLNERABluTY EVALUADON EVALUA W

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BWR DOMINANT ACCIDENT SEQUENCES CLASS 1 LOSS OF COOLANT MAKEUP IA HIGH PRESSURE IB STATION BLACK 0UT IC ATWS ID LOW PRESSURE lE LOSS OF DC CLASS 11 LOSS OF CONTAINMENT HEAT REMOVAL CLASS 111 LOCA CLASS IV ATWS

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DOMINANT ACCIDENT SEQUENCE QUANTIFICATION ALTERNATIVES e PLANT - SPECIFIC PRA e VY INDIVIDUAL PLANT EVALUATION e MODIFICATION OF REFERENCE BWR MARK I PRA TO ACCOUNT FOR KEY DIFFERENCES u PEACHBOTTOM BWR IPE ANALYSIS UTILIZED AS THE REFERENCE P0lNT l

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. INITIATING EVENT FREQUENCIES Number of Vermont Yankee Best Estimate Category Types of Events Events in Frequency Data Base (yr-1)

Transients Resulting in 20 Reactor Scram with Bypass 1.62 Potentially Available 2 Events Bounded by Loss of 2 Feedwater 0.16 3 MSIV Closure 4 0.32

Events with Bypass Net 8 Available 0.65 5 Inadvertent Opening of One 0 or More SRVs 0.062

Loss of Off-site Power 1 0.071

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DERIVATION OF VERNONT YANKEE SEVERE ACCIDF.NT SEQUENCE FREQUENCIES VERMONT YANKEE SURROCATE PLANT VERMONT YANKEE DOMINANT DOMINANT SEQUENCES UNIQUE FEATURES ACCIDENT SEQUENCES AND FREQUENCIES DESIGNATOR FREQUENCY FEATURE DELTA DESICNATOR FREQUENCY (per Rx yr) (per RK yr) Class IB Class IB

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TEQUV PHASE I 1E-6 -Initiator (x.044/.05) K.35 TEQUV PHASE I 3 .12- 7 ** ' o - 2)u, >

   -2 D/C (x2)

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   -No. of Switchyards-Underground transmission

TEQUV PIIASE II 2E-6 2 -V -No Event Specific K.52 TgQUV PHASE II 1.OE-6 Station Blackout Proceduren (Approx. snme Phase II) Phase III, IV (X3)

TEQUV PilASE III 4E-6 9 "# -Dattery Life K TEQUV PHASE III 4.4E-6 (Approx. same) h Phase II, III. IV TEQUV PilASE IV 1.6E-6 7g[w HPCI Potential Larger K.26 TEQUV PHASE IV 4.2E-7 DC Power Reqm't (K2)

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s SUNKARY OF CORE MELT FREQUENCY TYPES AT VERMONT YANKEE Class Description Frequency Fraction (yr" ) IA Loss of Makeup; RPV 1.33E-5 43 at high pressure IB Loss of Makeup: 6.2E-6 .20 Station Blackout IC Loss of Makeup: 2.6E-6 .08 ATWS ID Loss of Makeup: 3.9E-6 .12 RPV at low pressure IE Loss of Makeup: 2.4E-7 .01 Loss of DC buses II Loss of Containment 2.1E-6 .07 Heat Removal III LOCAs 7.3E-7 .02 ( IV ATWS 2.2E-6 .07 ! '

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I V Interfacing LOCA Netlitible Negligible i i TOTAL 3.1E-5 100% l i

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CONTAINMENT EVENT TREE DEVELOPMENT e CORE MELT ACCIDENT CLASS (ENTRY STATE) e CONTAINMENT MITIGATING SYSTEM RESPONSE e CONTAINMENT RESPONSE-FAILURE PATH E SIZE e REACTOR / AUXILIARY BUILDING RESPONSE

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CONTAINMENT EVENT TREE END STATES TIMING T (BEYOND CORE MELT) EARLY (E) T< 10 HOURS INTERMEDIATE (I) 10 < T< 24 HOURS LATE (L) T> 24 HOURS NO CONTAINMENT FAILURE (NCF) INTACT

! RELEASE QUANTITIES I (E0VIVALEllT 10 DINE) PLUS NOBLE GASEL

' HIGH (H) I> 10% MEDIUM (M) 0.1 < I < 10% ! LOW (L) I< 0.1% l

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RNCTIONAL EVENTS: GMAlHNT EVENT TEE e CONTAltNENT ISOLATION AND INITIAL INTEGRITY eACTIVEMITIGATION: DRYWEu. T/P CONTROL e PASSIVE MITIGATION: CONTAltNEh7 0F DEBRIS e COraAlfEENT HEAT REtOVAL-RHR

 - VENTING e COfRAltNENT BREACH SIZE
- Ji.AKAGE
- WERPRESSURE FAILURES e C0f3AlfEENT BREACH LOCATION YWELL-flETWELLAIRSPACE
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CONTAINMENT EVENT TREES TOP EVENTS ACT14 Wif1 CATION CONTA!NW(NT NTECRITY ACT)W AccDENT CON TAjWWEN T WITICATION PASSW CCNTANWENTCONTAINW[N1 NO NO ORYMll M TwCLL PEACIM CLASS ISOLAT)DN TEWPDI ATUp[ WIT lC A T10N HEAT PRESSURE LIAXACE CER- FNLURE AIR $PACC BJONC gy TP y & Nif)AL CON TR OL PEWOVAL CONTROL PRESSURE STATE NTICRITY (INJLCTION) (VENTINC) FAFLURE cu c, T 9 w v 0 c N 3 p 1 CLAS$i + B acccCNT . CONhN'JED DR1MLL ACh4 TAINWEN' WITICAft0N CLAS$ CONTANVENT NO NO CWR- DRTWELL WETwCLL COOLANT PASSW PEACTD' I OLAT 1IN pg gg ENTRY IEE33 # LIAXACE PPE$5URE FAILURE AtRSPACE WAKEUP CONDOL g p y AL CONTROL FAFLURE TO ftPV WII M IT STATE (ENTINC) ONXCEM CW y (D) (C) (H) (J) C a 1 P R I lY

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CONTAINMENT EVENT TREE QUANTIFICATION BASES

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e PAST PRA EXPERIENCE e DETERMINISTIC EVALUATIONS (VY)

 - MARCH /RMA
 - MAAP  f&AlwOEJ4 (4p[rps
 - HAND CALCULATIONS e IDCOR TECHfilCAL REPORT INFORMATION e ENGINEERlliG JUDGMENT
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SUMMARY OF REI. EASE FREQUENCY OF EACil RELEASE TYPE FOR EACil ACCIDF.NT Cl. ASS: DASE CASE

   "BEST ESTIMATE ** CASE Accident Class   Release Frequency (yr" ) of Release Type Designator Core Melt,g)

Frequency (yr Elf EM E LM LL WCF At Jyj h .} /w & IA 1.33E-05 7.38E-09 (l,l' 7h ./

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2.58E-08 1.2SE-07 6.89E-08 QQ 1.81E-07 1.29E-05 IB 6.20E-06 7.07E-09 1.11E-07 5. 0 7 E-0 7 3.56E-07 9.03E-07 4.31E-06 IC 2.60E-06 5.27E-07 2.83E-07 1.61E-08 9.31E-09 2.44E-08 1.74E-06 ID 3.90E-06 3.32E-09 4.60E-08 2.86E-08 5.44E-07 1.54E-06 1.74E-06 II 2.10E-06 4.73E-10 5.39E-07 1.56E-06 - - - III 7.30E-07 5.53E-10 9.06E-10 6.56E-09 4.55E-08 1.28E-07 5.49E-07 IV 2.20E-06 5.44E-09 6. 55 E-0 7 1.54E-06 - - - Sum 3.10E-05 5.51E-07 1.66E-06 3.78E-06 1.02E-06 2.78E-06 2.12E-05

% of Total Core Melt Frequency 100%  7.1% 12% 3.3% 9.0% 68%

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_ CONTAINMENT CONDITIONAL FAILURE PROBABILITY (D BASE AND SENSITIVITY CASE RESULTS CATAGORIZATION

EH/M EL__ LM LL

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_NCf_ BASE 7 12 4 9 68 S1 10 12 3 9 66 S2 3 9 4 10 74 S3 1 3 6 16 74 WHERE: S1 -

  .EARLY CONTAINMENT FAILURE INCREASE FOR BOTH HIGH AND LOW PRESSURE MELT DOWNS S2 -

INCREASED OPERATOR ACTION CREDIT FCR ATWS EVENTS S3 - S2 PLUS CLASSIFICATION OF TW CLASS AS LATE VERSUS EARLY RELEASE

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'I MARK I CONTAINMENT POLICY ISSUES e HYDR 0 GEN CONTROL . e DRYWELL SPRAY CAPABILITY l e CONTAINMENT VENTING e CORE DEBRIS CONTROL

e SEVERE ACCIDENT TRAINING AND PROCEDURES

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_ O HYDR 0 GEN CONTROL e OBJECTIVE: PREVENT HYDR 0 GEN COMBUSTION FOLLOWING SEVERE ACCIDENTS e VY STATUS

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CONTAINMENT INERTED

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DE-INERT TIME CONTROLLED BY TECHNICAL SPECIFICATIONS

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OPERATING HISTORY INDICATES DE-INERT TIME IS 1.1% OF TIME PLANT IS IN RUN MODE

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VY RECENTLY CONVERTED TO N2 CAD e EVALUATION: CURRENT VY PERFORMANCE AND DESIGN MEETS OBJECTIVE e RECOMMENDAT10llS: NONE

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DRYWELL SPRAY CAPABill'TY e OBJECTIVE: PROVIDE AC INDEPENDENT CAPABILITY 10'SPR k TO LOWER PRESSURE AND TEMPERATURE e VY STATUS:

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NORMAL CAPABILITY FROM RHR, BACKED UP BY SERVICE WATER (DEPENDENT ON AC POWER)

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DIESEL DRIVEN FIRE PUMP CAPABLE OF BEING ALIGNED TO LPC1/ CONTAINMENT SPRAY (AC INDEPENDENT)

- WATER SUPPLY FROM COOLING TOWER DEEP BASIN (AC DEPENDENT)

e EVALVATION: CURRENT CAPABILITY ACCEPTABLE, ENHANCEMENTS POSSIBLE 10 INCREASE RELIABILITY OF SUPPLY:FROM DIESEL FIRE PUMP e RECOMMENDATIONS:

- ENHANCE THE OPERATION OF THE DIESEL FIRE PUMP TO DRYWELL SPRAY FLOW PATH
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IMPROVE ABILITY TO RESTORE AC POWER .

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' CONTAINMENT VENTING

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e OBJECTIVE: PREVENT UNCONTROLLED LOSS OF CONTAINMENT INTE e VY STATUS:

- ONLY VERY SMALL VENTING PATHWAY PROCEDURALIZED
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STUDY IDENTIFIES 6 VENTING PATHWAYS

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RISKS OF VENTING DISCUSSED e EVALUATION: CAPABillTY FOR WETWELL VENTING, TO ACHIEVE FISSION PRODUCT SCRUBBING, SHOULD BE ENHANCED e RECOMMENDATIONS:

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ENGINEERING EVALUATION OF UPGRADING WETWELL VENTING TO INITIATED

- VENTING GUIDELINE DOCUMENT PREPARED OUTLINING 32 POTEN VENT PATHS, REQUIREMENTS FOR USE, ACCESS, ET '
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OBJECTIVE: PRECLUDE CONTAINMENT FAILURE DUE TO CORIUM/DNYWE SHELL CONTACT o e' VY STATUS:

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   - SMALL CORE VOLUME IN LARGE DRYWELL PROVIDES ASSURANCE CORIUM WILL NOT MIGRATE 10 DRYWELL SHELL
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   - DRYWELL SPRAYS SWILL " REFREEZE" CORIUM
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e EVALUATION: CbRRENTVYCONFIGURATIONMEETSOBJECTIVE e RECOMMENDATIONS: NONE s F l (

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,. SEVEERE ACCIDENT TRAINING AND PROCEDURES e

OBJECTIVE: ENSURE OPERATORS READY 10 USE PLANT FEATURES TO THE BEST ADVANTAGE IN SEVERE ACCIDENTS e VY STATUS:

- VY HAS IMPLEENTED E0P's BASED ON REV. 3 0F THE EPG's e EVALUATION: VY E ETS CURRENT GUIDANCE e RECOMENDAll0N:
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IMPLEENT REV. la 0F EPG's WHEN AVAILABLE

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CONSIDER REVISION 10 EXISTING E0P's FOR: o STATION BLACK 0UT o CONTAINMENT VENTING o DRYWELL SPRAY USING DIESEL FIRE PUMP o GUIDANCE TO SUPPLY CST VIA FIRE SYSTEM o SAMPLE SUPPRESSION P00L BEFORE TRANSFER TO CST o PROCEDURALIZE H22 /0 CONTROL t o COMPLETE DEVELOPENT OF VY SPECIFIC LEVEL / POWER CONTROL

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. DECISION ANALYSIS FOR RECOMENDA110NS e TWO PART TEST

- WHICH SEVERE ACCIDENTS WARRANT CONSIDERATION?
- WHAT IS ACCEPTABLE CONTAINMENT PERFORMANCE?

e SEVERE ACCIDENT SEQUENCES

- MOST CRITICAL TO CONTAINMENT PERFORMANCE ARE:

STATION BLACK 0UT ATWS WITH MSIV CLOSURE e ACCEPTABLE CONTAINMENT PERFORMANCE

- CONTAINMENT SHOULD REMAIN INTACT, WITH00i EXCESSIVE LEAKAGE FOR 211 HOURS FOLLOWING EVENT INITIATION e RECOMMENDATIONS REPRESENT THE CONCENSUS OF THE STUDY TEAM REGARDING PROCEDURAL AND DESIGN ENHANCEMENTS TO IMPROVE CONTAINMENT PERFORMANC MAXIM 1ZE USE OF EXISTING PLANT FEATURES
- AUGMENT EXISTING DESIGN WHERE CLEAR BENEFIT IS APPARENT

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RECOMMENDATIONS

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e -PROCEDURE ADDITIONS / ENHANCEMENTS e ISSUES FOR FURTHER EVALVATION e INSIGHTS BASED ON SEVERE ACCIDENT SEQUENCES l ! ! l l

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PROCEDURE ADDITIONS / ENHANCEMENTS - e PROVIDE PROCEDURAL GUIDANCE FOR RESTORATION OF AC POWER FOLLOWIN A STATION BLACK 0UT e PROVIDE A REPAIR PROCEDURE OUTLINING THE RESTORATION OF THE VERN TIE LINE IF DAMAGE SHOULD OCCUR e PROVIDE PROCEDURAL GUIDANCE FOR OPTIMIZATION OF DC POWER SOURCES FOLLOWING A STATION BLACK 0UT e PROVIDE PROCEDURAL GUIDANCE FOR USE OF THE SRV's FOR MANUAL-DEPRESSURIZATION FOLLOWING A STATION BLACK 0UT e PROVIDE PROCEDURAL GUIDANCE TO ALIGN THE DIESEL FIRE PUMP FOLLOWING STATION BLACK 0UT e PROVIDE LEVEL / POWER GUIDANCE FOR ATWS

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ISSUES FOR FURTHER EVALUATION e . UPGRADE E0P's TO REV. 4 0F EPG's e UPGRADE SELECTED VALVE OPERATORS IN FLOW PATH FROM. DIESEL FIRE PUMP TO LPCI/ CONTAINMENT SPRAY e ENHANCE RELIABILITY AND CAPABILITY OF CONTAINMENT WETWELL VENT , l

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*' INSIGHTS BASED ON SEVERE ACCIDENT SEQUENCES e PROVIDE ADDITIONAL EVALUATION, GUIDANCE AND OPERATOR TRAINING ON RESPONSE OF THE RHR PUMPS TO HIGH SUPPRESSION P0OL TEMPERATURES e IMPROVE RELIABILITY OF THE NITROGEN SUPPLY 10 THE ADS VALVES e EVALUATE DESIGN AND POSI-ACCIDENT RELIABILITY OF THE SERVICE WATER SYSTEM e EVALUATE STANDBY LIQUID CONTROL SYSTEM DESIGN AND RELIABILITY

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