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| | issue date = 01/06/1997 | | | issue date = 01/06/1997 |
| | title = LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr | | | title = LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr |
| | author name = PFITZER W A, WEBRING R L | | | author name = Pfitzer W, Webring R |
| | author affiliation = WASHINGTON PUBLIC POWER SUPPLY SYSTEM | | | author affiliation = WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| | addressee name = | | | addressee name = |
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| =Text= | | =Text= |
| {{#Wiki_filter:CATEGORY1REGULRZfINFORMATION DZSTRIBUTZOlYSTEM (RIDE)~~ACCESSION NBR:9701140016 DOC.DATE: | | {{#Wiki_filter:CATEGORY 1 REGULRZf INFORMATION DZSTRIBUTZOlYSTEM (RIDE) |
| 97/01/06NOTARIZED: | | ~ |
| NOFACIL:50-397 WPPSSNuclearProject,Unit2,Washington PublicPoweAUTH.NAMEAUTHORAFFILIATION PFITZER,W.A.
| | ACCESSION NBR:9701140016 DOC.DATE: 97/01/06 NOTARIZED: NO DOCKET FACIL:50-397 |
| Washington PublicPowerSupplySystemWEBRING,R.L. | | ~ WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION PFITZER,W.A. Washington Public Power Supply System WEBRING,R.L. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION |
| Washington PublicPowerSupplySystemRECIP.NAME RECIPIENT AFFILIATION DOCKET05000397 | |
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| ==SUBJECT:== | | ==SUBJECT:== |
| LER96-008-00:on 961205,failure tocomplywithTSactionrequirement foremergency corecoolingsysactuation instrumentation occurredduetounidentified inoperability condition.PMR willbeconducted.W/970106 ltr.DISTRIBUTION CODE:ZE22TCOPIESRECEIVED:LTR iENCL4SIZE:TITLE:50.73/50.9 LicenseeEventReport(LER),IncidentRpt,etc.NOTES:ERECIPIENT IDCODE/NAME PD4-2PDINTERNAL:
| | LER 96-008-00:on 961205,failure to comply with TS action requirement for emergency core cooling sys actuation instrumentation occurred due to unidentified inoperability condition.PMR will be conducted.W/970106 ltr. |
| ACRSAEOD/SPD/RRAB NRR/DE/ECGB NRR/DE/EMEB NRR/DRCH/HICB NRR/DRCH/HQMB NRR/DSSA/SPLB RES/DET/EIB EXTERNAL:
| | DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR i ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc. |
| LSTLOBBYWARDNOACMURPHYFG.A NRCPDRCOPIESLTTRENCL111122111111111111111111RECIPIENT IDCODE/NAME COLBURNFT PDRABFILECNTEELBNRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRPM/PECB NRR/DSSA/SRXB RGN4FILE01LITCOBRYCEFJHNOACPOOREFWNUDOCSFULLTXTCOPIESLTTRENCL11221111111111111'111111D0NTNOTETOALL"RIDS"RECIPIENTS:
| | 4 SIZE: |
| PLEASEHELPUSTOREDUCEWASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMOWFN5D-5(EXT.
| | E NOTES: |
| 415-2083) | | RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD4-2 PD 1 1 COLBURNFT 1 1 INTERNAL: ACRS 1 1 PD RAB 2 2 AEOD/SPD/RRAB 2 2 FILE C NTE 1 1 NRR/DE/ECGB 1 1 ELB 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DET/EIB 1 1 RGN4 FILE 01 1 D EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCEFJ H 1 1 0 NOAC MURPHYFG.A 1 1 NOAC POOREFW 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N |
| TOELIMINATE YOURNAMEFROMDISTRIBUTION LISTSFORDOCUMENTS YOUDON'TNEED!FULLTEXT.ONVERSION REQUIREDTOTALNUMBEF.OFCOPIESREQUIRED:
| | T NOTE TO ALL "RIDS" RECIPIENTS: |
| LTTR26ENCL26 WASHINGTON PUBLICPOWERSUPPLYSYSTEMP.O.Box968~Richlaufi, washington 99352-0968 January6,1997G02-97-003 DocketNo.50-397U.S.NuclearRegulatory Commission Attn:DocumentControlDeskWashington, D.C.20555Gentlemen:
| | PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED! |
| | FULL TEXT .ONVERSION REQUIRED TOTAL NUMBEF. OF COPIES REQUIRED: LTTR 26 ENCL 26 |
| | |
| | WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ Richlaufi, washington 99352-0968 January 6, 1997 G02-97-003 Docket No. 50-397 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen: |
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| ==Subject:== | | ==Subject:== |
| NUCLEARPLANTWNP-2,OPERATING LICENSENPF-21LICENSEEEV1PITREPORTNO.96-008-00 Transmitted herewithisLicenseeEventReportNo.96-008-00 forWNP-2.Thisreportissubmitted pursuantto10CFR50.73(a)(2)(i)(B) anddiscusses theitemsofreportability, corrective actiontaken,andactiontoprecluderecurrence.
| | NUCLEAR PLANT WNP-2, OPERATING LICENSE NPF-21 LICENSEE EV1PIT REPORT NO. 96-008-00 Transmitted herewith is Licensee Event Report No. 96-008-00 for WNP-2. This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) and discusses the items of reportability, corrective action taken, and action to preclude recurrence. |
| Shouldyouhaveanyquestions ordesireadditional information regarding thismatter,pleasecallmeorMs.L.C.Fernandez at(509)377-4147.
| | Should you have any questions or desire additional information regarding this matter, please call me or Ms. L. C. Fernandez at (509) 377-4147. |
| Respectfully, | | Respectfully, |
| .L.Webring~VicePresident, Operations Support/PIO MailDropPE08RLW/CDMEnclosure | | . L. Webring ~ |
| (!iCC:LJCallan-NRCRIVJEDyer-NRCRIVKEPerkins,Jr.-NRCRIV,WalnutCreekFieldOfficeNSReynolds-Winston&StrawnTGColburn-NRRDLWilliams-BPA/399NRCSr.ResidentInspector | | Vice President, Operations Support/PIO Mail Drop PE08 RLW/CDM Enclosure |
| -927N/~ililAR~9'70ii400i6 | | (!i CC: LJ Callan - NRC RIV JE Dyer - NRC RIV KE Perkins, Jr. - NRC RIV, Walnut Creek Field Office NS Reynolds - Winston & Strawn TG Colburn - NRR DL Williams - BPA/399 NRC Sr. Resident Inspector - 927N |
| '770i06PDRADOCK05000397SPDR FACILITYNAHE(1)LICENSEEEVENEPORT(LER)OCk.c<NUHBER(2)AGE(3)Nhshinton/@clearPlant-Lhit2050003971FITLE(4)FAILURETOCOMPLYWITHATECHNICAL SPECIFICATION ACTIONREQUIREMENT FORTHEEMERGENCY CORECOOLINGSYSTEMACTUATION INSTRUMENTATION DUETOUNIDENTIFIED INOPERABILITY CONDITION EVEHTDATE(5)TEARHOHTHDAYTEARLERNUHBER(6)EQUENTIALUHBEREVISIOHUHBERREPORTDATE(7)YEARDAYOTHERFACILITIES INVOLVED(8)OCKETNNlBERS(S)
| | ~ |
| FACILITYNNlES50120596960080001069750ERATIHGE(9)HISREPORTISSUBMITTED PURSUAHTTOTHEREQUIREHEHTS OF10CFR5:(CheckoneormoreofthefoioNng)(11)RLEVEL(10)0.402(b)0.405(a)(1)(i) 0.405(a)(1)(ii) 0.405(a)(1)(iii) 0.405(a)(1)(iv) 0.405(a)(1)(v) 0.405(C)0.36(c)(1) 0.36(c)(2) 0.73(a)(2)(i) 0.73(a)(2)(it) 0.73(a)(2)(iii) 0.73(a)(2)(iv) 0.73(a)(2)(v) 0.73(a)(2)(vii) 0.73(a)(2)(viii)(A) 0.73(a)(2)(viii)(B) 0.73(a)(2)(x)
| | il/ il A R ~ |
| .71(b).71(c)THER(SpecifyinAbstractIoHandinText,NRCForm366A)AHELICEHSEECONTACTFORTHISLER(12)W.A.Pfitzer,Technical Specialist REAC(X)ETELEPHONE NNIBER0977.2419COHPLETEOHELINEFOREACHCOHPOHEHT FAILUREDESCRIBED IHTHISREPORT(13)CAUSESYSTEllCOHPOHENT PSHANUFACTURER EPORTABLE hery''PCAUSESTSTEH0NPRDS38CONPONEHT HANUFACTURER EPORTABLE 0HPRDS~BR!$;4?;f.SUPPLENEHTAL REPORTEXPECTED(14)YES(lfyes,completeEXPECTEDSUBHISSIOH DATE)NOTRACfUOIXPECTEDSUSHISSIONHONTHDAYYEARATE(15)OnDecember5,1996withtheplantinMode1at100%reactorpower,itwasdetermined thatWNP-2mayhavefailedtocomplywithaTechnical Specification actionrequirement fortheEmergency CoreCoolingSystem(ECCS)Actuation Instrumentation. | | 9'70ii400i6 '770i06 PDR ADOCK 05000397 S PDR |
| Basedonsubsequent
| | |
| : analysis, itwasdetermined thatthepressureswitchesdesignedtoinitiatetheHighPressureCoreSpray(HPCS)systemonhighcontainment drywellpressurehadexceededtheirTechnical Specification allowable valuesonseveraloccasions duringtheperiodfromJune10,1996throughNovember24,1996.ContrarytoECCSActuation Instrumentation Technical Specification 3.3.3.b,actionwasnottakenwithin24hourstoperformAction30ofTable3.3.3-1becausetheinoperability condition hadnotbeenidentified.
| | LICENSEE EVEN EPORT (LER) |
| Action30requiresthattheinoperable instrumentation channel(s) beplacedinthetrippedcondition within1hourortheHPCSsystembedeclaredinoperable.
| | FACILITY NAHE (1) OCk.c< NUHBER (2) AGE (3) |
| Inaccordance withTechnical Specification 3.5.1,theHPCSsystemcanbeinoperable forupto14daysbeforeadditional actionisrequired.
| | Nhshin ton /@clear Plant - Lhit 2 0 5 0 0 0 3 9 7 1 F ITLE (4) |
| Therootcauseofthiseventwasaprogramdeficiency inthatnowarehouse controlswereplacedontheissuanceofthepressureswitches.
| | FAILURE TO COMPLY WITH A TECHNICAL SPECIFICATION ACTION REQUIREMENT FOR THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO UNIDENTIFIED INOPERABILITYCONDITION EVEHT DATE (5) LER NUHBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8) |
| APlantModification Request(PMR)wasinitiated whichwouldhavepromptedtheappropriate engineering analysispriortoinstallation oftheswitches, butthePMRwaslatercanceledin1989andnoprocesstieexistedbetweenthePMRandtheswitches.
| | HOHTH DAY TEAR TEAR EQUENT IAL EV IS IOH DAY YEAR FACILITY NNlES OCKET NNlBERS(S) |
| Thisresultedinthereplacement pressureswitchesnotbeinginstalled intheventedconfiguration asrequired.
| | UHBER UHBER 50 1 2 0 5 9 6 9 6 0 0 8 0 0 0 1 0 6 9 7 50 ERAT IHG HIS REPORT IS SUBMITTED PURSUAHT TO THE REQUIREHEHTS OF 10 CFR 5: (Check one or more of the foi oN ng) (11) |
| Toensureoperability oftheHPCSsystemhighdrywellpressuretripfunction, immediate corrective actionwastakentoventtheassociated drywellpressureswitchestothereactorbuildingatmosphere andverifythesetpoints inaccordance withtheChannelFunctional Test(CFT)surveillance procedures, Furthercorrective actionshavebeencompleted toestablish alimitation onusefortheaffectedpressureswitchesandrequiring anengineering evaluation beperformed priortouseinotherapplications toensurethereplacement pressureswitchesarecorrectfortheapplication.
| | E (9) |
| Thiseventposednothreattothehealthandsafetyofeitherthepublicorplantpersonnel, I+I LICENSEEEVENTREPORT)TEXTCONTINUATION FACILITYNAKE(1)Washington NuclearPlant-Unit2OCKETNUKSER(2)05000397LERKUKSER(8)earwherev.Now6080AGE(3)2F5ITLE(4)PAILCM!TQCOMPLYWITHATECHNICAL APECIPICATIQN ACTIONIIHQCECIIQMT PQETHEEMERGENCY CORECOOLINGSYSTEMACTUATION INSTRUMENTATION DUETOANUNIDERI'IFIED INOPERABILITY CONDITION vnDOnDecember5,1996withtheplantinMode1at100%reactorpower,itwasdetermined thatWNP-2mayhavefailedtocomplywithaTechnical Specification actionrequirement fortheEmergency CoreCoolingSystem(ECCS)Actuation Instrumentation.
| | R LEVEL 0.402(b) 0.405(C) 0.73(a)(2)(iv) .71(b) |
| Duringaninvestigation ofasetpointdriftproblemrelatedtopressureswitchesMS-PS-47B and47CtPS),itwasdiscovered thatthepressureswitchcaseswerenotventedtothereactorbuildingatmosphere asassumedintheirsetpointcalculation.
| | (10) 0.405(a)(1)(i) 0.36(c)(1) 0.73(a)(2)(v) .71(c) 0.405(a)(1)(ii) 0.36(c)(2) 0.73(a)(2)(vii) THER (Specify in Abstract 0.405(a)(1)(iii) 0.73(a)(2)(i) 0.73(a)(2)(viii)(A) IoH and in Text, NRC 0.405(a)(1)(iv) 0.73(a)(2)(it) 0.73(a)(2)(viii)(B) Form 366A) 0.405(a)(1)(v) 0.73(a)(2)(iii) 0.73(a)(2)(x) |
| ThesepressureswitchesaretheChannelBandCsensors,respectively, forHighPressureCoreSpray(HPCS)system[BG]initiation onhighcontainment drywellpressure.
| | LICEHSEE CONTACT FOR THIS LER (12) |
| Basedonfurtherinvestigation, itwasdetermined thatthepressureswitchcasesforMS-PS<7Aand47DP'S],whicharetheChannelAandDsensors,respectively, werealsonotventedtothereactorbuildingatmosphere asassumedinthesetpointcalculation.
| | AHE TELEPHONE NNIBER W.A. Pfitzer, Technical Specialist REA C(X)E 0 9 7 7 . 2 4 1 9 COHPLETE OHE LINE FOR EACH COHPOHEHT FAILURE DESCRIBED IH THIS REPORT (13) |
| TheTechnical Specification tripsetpointforthesepressureswitchesis~1.65psigandtheallowable valueis61A85psig.Anunventedandsealedpressureswitchcaseissubjecttopressurechangeswithinthecaseduetoambienttemperature variations and,becausethetripsetpointandallowable valueareclosetoatmospheric
| | CAUSE SYSTEll COHPOHENT HANUFACTURER EPORTABLE hery''P CAUSE STSTEH CONPONEHT HANUFACTURER EPORTABLE 0 NPRDS 0 HPRDS ~BR! |
| : pressure, thesetemperature variations cancreateinternalpressurechangeswhichcouldaffecttheswitchsetpoint.
| | $ ;4?;f. |
| Atlowsetpointpressures, anincreaseininternalcasetemperature willcauseanincreaseininternalcasepressure.
| | P S 3 8 SUPPLENEHTAL REPORT EXPECTED (14) XPECTED SUSHI SS ION HONTH DAY YEAR ATE (15) |
| Thisincreased casepressureactsagainstthesenseddrywellpressuretoshiftthesetpointinthenonconservative direction suchthatahigherdrywellpressurewouldberequiredforHPCSsysteminitiation.
| | YES (lf yes, complete EXPECTED SUBHISSIOH DATE) NO TRACf UOI On December 5, 1996 with the plant in Mode 1 at 100% reactor power, it was determined that WNP-2 may have failed to comply with a Technical Specification action requirement for the Emergency Core Cooling System (ECCS) Actuation Instrumentation. Based on subsequent analysis, it was determined that the pressure switches designed to initiate the High Pressure Core Spray (HPCS) system on high containment drywell pressure had exceeded their Technical Specification allowable values on several occasions during the period from June 10, 1996 through November 24, 1996. Contrary to ECCS Actuation Instrumentation Technical Specification 3.3.3.b, action was not taken within 24 hours to perform Action 30 of Table 3.3.3-1 because the inoperability condition had not been identified. Action 30 requires that the inoperable instrumentation channel(s) be placed in the tripped condition within 1 hour or the HPCS system be declared inoperable. In accordance with Technical Specification 3.5.1, the HPCS system can be inoperable for up to 14 days before additional action is required. |
| : Moreover, theswitchsetpointcanalsobeaffectedbychangesinatmospheric pressure.
| | The root cause of this event was a program deficiency in that no warehouse controls were placed on the issuance of the pressure switches. A Plant Modification Request (PMR) was initiated which would have prompted the appropriate engineering analysis prior to installation of the switches, but the PMR was later canceled in 1989 and no process tie existed between the PMR and the switches. This resulted in the replacement pressure switches not being installed in the vented configuration as required. |
| Following eachpressureswitchcalibration, anunventedcaseiseffectively sealedbyinstallation ofthecoverplate.Withthepressureswitchcaseunventedandsealed,achangeinatmospheric pressurebetweencalibrations willbeevidentbyasetpointdriftobservedatthenextcalibration. | | To ensure operability of the HPCS system high drywell pressure trip function, immediate corrective action was taken to vent the associated drywell pressure switches to the reactor building atmosphere and verify the setpoints in accordance with the Channel Functional Test (CFT) surveillance procedures, Further corrective actions have been completed to establish a limitation on use for the affected pressure switches and requiring an engineering evaluation be performed prior to use in other applications to ensure the replacement pressure switches are correct for the application. |
| Ahighatmospheric pressureatthetimethepressureswitchcaseissealedfollowing calibration willresultinashiftinthesetpointinthenonconservative direction (higherdrywellpressurerequiredtoinitiateHPCS)astheatmospheric pressurewithinthedrywellandreactorbuildingdecreases betweencalibrations.
| | This event posed no threat to the health and safety of either the public or plant personnel, |
| OnDecember15,1996,ananalysisofthetemperature andatmospheric pressureeffectsdescribed abovedetermined thatthepressureswitchesfortheHPCSsystemhighdrywellpressuretripfunctionhadexceededtheirTechnical Specification allowable valuesonseveraloccasions duringtheperiodfromJune10,1996throughNovember24,1996.ContrarytoECCSActuation Instrumentation Technical Specification 3.3.3.b,actionwasnottakenwithin24hourstoperformAction30ofTable3.3.3-1becausetheinoperable condition hadnotbeenidentified.
| | |
| Action30requiresthattheinoperable instrumentation channel(s) beplacedinthe,trippedcondition within1hourortheHPCSsystembedeclaredinoperable.
| | I+I LICENSEE EVENT REPORT ) |
| Inaccordance withTechnical Specification 3.5.1,theHPCSsystemcanbeinoperable forupto14daysbeforeadditional actionisrequired.
| | TEXT CONTINUATION FACILITY NAKE (1) OCKET NUKSER (2) LER KUKSER (8) AGE (3) ear wher ev. Now Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 6 08 0 2 F 5 ITLE (4) |
| ImmirrivActinToensureoperability oftheHPCSsystemhighdrywellpressuretripfunction, immediate actionwastakenonDecember5,1996toventpressureswitchesMS-PS-47A, 47B,47C,and47Dtothereactorbuildingatmosphere andverifythesetpoints inaccordance withtheChannelFunctional Test(CFT)surveillance procedures 7.4.3.3.1.53 and7.4.3.3.1.54.
| | PAILCM!TQ COMPLY WITH A TECHNICAL APECIPICATIQN ACTION IIHQCECIIQMT PQE THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO AN UNIDERI'IFIED INOPERABILITYCONDITION vn D On December 5, 1996 with the plant in Mode 1 at 100% reactor power, it was determined that WNP-2 may have failed to comply with a Technical Specification action requirement for the Emergency Core Cooling System (ECCS) Actuation Instrumentation. |
| Thepressureswitcheswereventedbyremovingthevendorinstalled caseventcaps.I LICENSEEEVENTREPORT)TEXTCONTINUATION FACILITYRAIIE(I)Washington NuclearPlant-Unit2OCKETRLNBER(2)05000397LERNUIIBER(8)eerurberev.Ho.6080AGE(3)3F5ITLE(4)~AILCREI'CCOMPLYWITEATBCIINICAL PPECIPICATICN ACIICNIE!QIBI!MENTPCRTHEEMERGENCY CORECOOLINGSYSTEMACTUATION INSTRUMENTATION DUETOANUNIDENTIFIED INOPERABILITY CONDITION FrE'nrrivAinFrEvn2.3.Pursuantto10CFR50.73(a)(2)(i)(B),
| | During an investigation of a setpoint drift problem related to pressure switches MS-PS-47B and 47C tPS), it was discovered that the pressure switch cases were not vented to the reactor building atmosphere as assumed in their setpoint calculation. These pressure switches are the Channel B and C sensors, respectively, for High Pressure Core Spray (HPCS) system [BG] initiation on high containment drywell pressure. Based on further investigation, it was determined that the pressure switch cases for MS-PS<7A and 47D P'S], which are the Channel A and D sensors, respectively, were also not vented to the reactor building atmosphere as assumed in the setpoint calculation. The Technical Specification trip setpoint for these pressure switches is ~ 1.65 psig and the allowable value is 6 1A85 psig. |
| thiseventisbeingreportedasacondition prohibited bytheWNP-2Technical Specifications.
| | An unvented and sealed pressure switch case is subject to pressure changes within the case due to ambient temperature variations and, because the trip setpoint and allowable value are close to atmospheric pressure, these temperature variations can create internal pressure changes which could affect the switch setpoint. At low setpoint pressures, an increase in internal case temperature will cause an increase in internal case pressure. |
| PressureswitchesMS-PS-47A, 47B,47C,and47DwerereplacedduringtheSpring1996(R-11)maintenance andrefueling outageastheywereapproaching theirqualified life.TheoriginalStatic-0-Ring (vendor)suppliedpressureswitchmodel(12NAASX10TI',
| | This increased case pressure acts against the sensed drywell pressure to shift the setpoint in the nonconservative direction such that a higher drywell pressure would be required for HPCS system initiation. Moreover, the switch setpoint can also be affected by changes in atmospheric pressure. Following each pressure switch calibration, an unvented case is effectively sealed by installation of the cover plate. With the pressure switch case unvented and sealed, a change in atmospheric pressure between calibrations will be evident by a setpoint drift observed at the next calibration. A high atmospheric pressure at the time the pressure switch case is sealed following calibration willresult in a shift in the setpoint in the nonconservative direction (higher drywell pressure required to initiate HPCS) as the atmospheric pressure within the drywell and reactor building decreases between calibrations. |
| orTT)wasreplacedwithanewmodel(12N6BB4NXC1AJJTIX12, orX12)becausetheoriginalmodelcouldnolongerbeprocuredQualityClass1.Thenewmodelpressureswitchdifferedfromtheoriginalmodelinthatthediaphragm materialwaschangedfromkaptontostainless steelandthenewmodelincludedavendorsuppliedintegralair-tight conduitsealandcaseventport.Thecaseventportwascappedbut,atthecustomer's option,thecapcouldberemovedtoventthepressureswitchcase.Thevendorprovidedtheoptiontocap(unvent)thepressureswitchcaseforthoseapplications wherethepressureswitchisexpectedtoremainfunctional inextremeenvironments.
| | On December 15, 1996, an analysis of the temperature and atmospheric pressure effects described above determined that the pressure switches for the HPCS system high drywell pressure trip function had exceeded their Technical Specification allowable values on several occasions during the period from June 10, 1996 through November 24, 1996. Contrary to ECCS Actuation Instrumentation Technical Specification 3.3.3.b, action was not taken within 24 hours to perform Action 30 of Table 3.3.3-1 because the inoperable condition had not been identified. Action 30 requires that the inoperable instrumentation channel(s) be placed in the, tripped condition within 1 hour or the HPCS system be declared inoperable. In accordance with Technical Specification 3.5.1, the HPCS system can be inoperable for up to 14 days before additional action is required. |
| Theoriginalpressureswitchdidnotincludeanintegralair-tight conduitseal.Revision0ofthesetpointcalculation forMS-PSQ7A, 47B,47C,and47Destablished thesetpointlimitsfortheoriginalpressureswitchmodelTTbasedonthedevicebeingventedtothereactorbuildingatmosphere.
| | Imm i rr iv Acti n To ensure operability of the HPCS system high drywell pressure trip function, immediate action was taken on December 5, 1996 to vent pressure switches MS-PS-47A, 47B, 47C, and 47D to the reactor building atmosphere and verify the setpoints in accordance with the Channel Functional Test (CFT) surveillance procedures 7.4.3.3.1.53 and 7.4.3.3.1.54. The pressure switches were vented by removing the vendor installed case vent caps. |
| Revision1ofthecalculation addressed thechangesresulting fromreplacement oftheoriginalpressureswitchmodelwiththenewmodelX12.Therevisedcalculation assumedthatthenewpressureswitchmodelwouldbevented.However,thenewpressureswitchmodelwasnotinstalled inaventedconfiguration becauseofthepresenceoftheintegralair-tight conduitsealandthefailuretouncaptheventport,Hence,thetemperature andatmospheric pressureeffectsonthenewpressureswitcheswerelargerthanassumedinthesetpointcalculation.
| | I |
| Basedonananalysisoftheseeffects,itwasdetermined thatpressureswitchesMS-PS-47A, 47B,47C,and47DhadexceededtheirTechnical Specification allowable valuesonseveraloccasions duringtheperiodfromJune10,1996throughNovember24,1996.Asdiscussed above,theeffectsfromchangesinatmospheric pressurerelativetotheatmospheric pressurepresentatthetimeofcalibration wereintroduced bythefailuretouncaptheventport.Thesetpointdriftproblemobservedfollowing installation ofthenewpressureswitchmodelhasbeenattributed tothefailuretouncaptheventportandtheeffectsfromchangesinatmospheric pressure.
| | |
| Thus,itisbelievedthatproperventingofthepressureswitcheswillresolvethesetpointdriftphenomenon andrestorethepressureswitchestoreliableoperation. | | LICENSEE EVENT REPORT ) |
| Tovalidatethisconclusion andensurecontinued operability, thesetpoints forpressureswitchesMS-PS-47A, 47B,47C,and47Dwillbeverifiedweeklyuntilthepressureswitchesexhibitapatternofacceptable setpointdriftinaccordance withtheadministrative limitsoftheCFTsurveillance procedure.
| | TEXT CONTINUATION FACILITY RAIIE (I) OCKET RLNBER (2) LER NUIIBER (8) AGE (3) eer urber ev. Ho. |
| ThenewpressureswitchmodelX12wasusedtoreplacepressureswitchesMS-PS-47A, 47B,47C,and47DduringtheR-11outage.Thenewpressureswitcheswereinstalled inthefieldinanunventedconfiguration becausetheworkorderforinstallation andcalibration didnotincludeinstructions toremovetheventcap.Furthermore, therewasnoexplicitdesigndocumentrequirement toremovetheventcapbecausetherewasnoengineering evaluation (substitution LICENSEEEVENTREPORT(L)TEXTCONTINUATION FAC!LITTHARE(1)Washington NuclearPlant-Unit2OCKETHUHBER(2)05000397LERHUHBER(8)mherev.Ho.6080AGE(3)4F5iTLE(4)FAILURETOCOMPLYWITHATECHNICAL SPECIFICATION ACTIONREQUIREMENT FORTHEEMERGENCY CORECOOLINGSYSTEMACTUATION INSTRUMENTATION DUETOANUNIDENTIFIED INOPERABILITY CONDITION evaluation, designchangeevaluation, orequivalent changeevaluation) performed whichassessedreplacement oftheoriginalpressureswitchmodelTTwiththenewmodel.Thisevaluation wasnotperformed becausenowarehouse controlswereplacedontheissuanceofthepressureswitches.
| | Washington Nuclear Plant - Unit 2 0 0 3 9 7 0 5 0 6 08 0 3 F 5 ITLE (4) |
| In1984,PlantModification Request(PMR)84-1125-0 wasissuedtoreplaceStatic-0-Ring pressureswitchmodelTt'ithmodel12N6BB4NXC1AJJTIX6 (X6)astheoriginalmodelcouldnolongerbeprocuredQualityClass1.In1988,beforeanyofthenewmodelpressureswitcheswereinstalled, Static-0-Ring issueda10CFRPart21notification againsttheX6modelpressureswitchbecauseofsetpointdriftduetoprocesspermeation throughthediaphragm.
| | ~ AILCRE I'C COMPLY WITE A TBCIINICALPPECIPICATICN ACIICN IE!Q IBI!MENT PCR THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO AN UNIDENTIFIED INOPERABILITYCONDITION F rE 'n rr iv A in F rEv n Pursuant to 10 CFR 50.73(a)(2)(i)(B), this event is being reported as a condition prohibited by the WNP-2 Technical Specifications. |
| Static-0-Ring subsequently replacedmodelX6withmodelX12toresolvethe10CFRPart21concern.Theonlychangewasthatthediaphragm materialwaschangedfromkaptontostainless steel.OnMarch9,1989,SupplySystemSubstitution Evaluation 567,Revision0,waspreparedtoauthorize modelX12asareplacement formodelX6.Thesubstitution'evaluation identified thataPMRwasrequiredforinstallation ofthenewmodelpressureswitch.However,thisrequirement wasnotenteredintotheMaterialManagement System(MMS)asa"Limitation onUse"becauseatthetimetherewasnoprocedural requirement todoso.Asanunrelated action,thesubstitution evaluation procedure (SPES-1,Section7.47)wasrevisedapproximately twoyearslater,onJune15,1991,requiring a"Limitation onUse"(includes entrymtheMMS)foritemswhereaPMRisrequiredforinstallation. | | : 2. Pressure switches MS-PS-47A, 47B, 47C, and 47D were replaced during the Spring 1996 (R-11) maintenance and refueling outage as they were approaching their qualified life. The original Static-0-Ring (vendor) supplied pressure switch model (12NAASX10TI', or TT) was replaced with a new model (12N6BB4NXC1AJJTIX12, or X12) because the original model could no longer be procured Quality Class 1. The new model pressure switch differed from the original model in that the diaphragm material was changed from kapton to stainless steel and the new model included a vendor supplied integral air-tight conduit seal and case vent port. The case vent port was capped but, at the customer's option, the cap could be removed to vent the pressure switch case. The vendor provided the option to cap (unvent) the pressure switch case for those applications where the pressure switch is expected to remain functional in extreme environments. The original pressure switch did not include an integral air-tight conduit seal. |
| OnAugust3,1989,PMR84-1125-0 wascanceledforunknownreasons.Thiseffectively eliminated therequirement foranengineering evaluation ofthedifferences betweentheoriginalpressureswitchmodelandmodelX6.Substitution Evaluation 567,Revision0,onlyaddressed thedifferences betweenpressureswitchmodelX6andmodelX12(i.e.,thechangefromakaptontoastainless steeldiaphragm)
| | Revision 0 of the setpoint calculation for MS-PSQ7A, 47B, 47C, and 47D established the setpoint limits for the original pressure switch model TT based on the device being vented to the reactor building atmosphere. Revision 1 of the calculation addressed the changes resulting from replacement of the original pressure switch model with the new model X12. The revised calculation assumed that the new pressure switch model would be vented. However, the new pressure switch model was not installed in a vented configuration because of the presence of the integral air-tight conduit seal and the failure to uncap the vent port, Hence, the temperature and atmospheric pressure effects on the new pressure switches were larger than assumed in the setpoint calculation. Based on an analysis of these effects, it was determined that pressure switches MS-PS-47A, 47B, 47C, and 47D had exceeded their Technical Specification allowable values on several occasions during the period from June 10, 1996 through November 24, 1996. |
| ~BecausethePMRwascanceledandnowarehouse controlswereplacedonissuanceofthepressureswitches, noengineering evaluation wasperformed whichauthorized theuseofeithermodelsX6orX12asareplacement forpressureswitchesMS-PS-47A, 47B,47C,and47D.IftheMMShadcontained auseagelimitation againstthepressureswitchesrequiring aPMRforinstallation, thislimitation wouldhaveensuredthattheappropriate engineering evaluation (i.e.,substitution evaluation, designchangeevaluation, orequivalent changeevaluation) wasperformed andadequateinstructions wereprovidedforinstallation. | | As discussed above, the effects from changes in atmospheric pressure relative to the atmospheric pressure present at the time of calibration were introduced by the failure to uncap the vent port. |
| A"Limitation onUse"hasbeenenteredintotheMMStoensurethatStatic-0-Ring pressureswitchmodelX12isinstalled intheplantonlyafteraproperengineering evaluation hasbeenperformed. | | The setpoint drift problem observed following installation of the new pressure switch model has been attributed to the failure to uncap the vent port and the effects from changes in atmospheric pressure. Thus, it is believed that proper venting of the pressure switches will resolve the setpoint drift phenomenon and restore the pressure switches to reliable operation. To validate this conclusion and ensure continued operability, the setpoints for pressure switches MS-PS-47A, 47B, 47C, and 47D will be verified weekly until the pressure switches exhibit a pattern of acceptable setpoint drift in accordance with the administrative limits of the CFT surveillance procedure. |
| Additionally, anengineering evaluation wascompleted onNovember27,1996toverifythatStatic-0-Ring pressureswitchmodelX12isthecorrectmodelfortheMS-PS-47A, 47B,47C,and47Dapplication. | | : 3. The new pressure switch model X12 was used to replace pressure switches MS-PS-47A, 47B, 47C, and 47D during the R-11 outage. The new pressure switches were installed in the field in an unvented configuration because the work order for installation and calibration did not include instructions to remove the vent cap. Furthermore, there was no explicit design document requirement to remove the vent cap because there was no engineering evaluation (substitution |
| Therevisiontothesubstitution evaluation procedure providesassurance thatsinceJune15,1991theMMShasbeenupdatedwitha"Limitation onUse"wheneverasubstitution requiresaPMRforinstallation.
| | |
| However,therecouldbeothercaseswherematerialwasprocuredforaPMRpriortotheprocedure revisionsuchthataPMRuseagelimitation wasnotenteredintotheMMS,thematerialwasstoredinthewarehouse (notinstalled intheplant),andthenthePMRwascanceled. | | LICENSEE EVENT REPORT (L ) |
| Toaddressthispossibility, asearchofthePlantTrackingLog(PTL)wasconducted forsimilarcases.Nosimilarcaseswerefound.AreviewofopenandcanceledPMRswillalsobeperformed toensurethereisnomaterialorderedforaPMRwhichdoesnothavealimitation onuse.
| | TEXT CONTINUATION FAC!LITT HARE (1) OCKET HUHBER (2) LER HUHBER (8) AGE (3) mher ev. Ho. |
| LICENSEEEVENTREPORT(R)TEXTCONTINUATION lFACILITYRNIE(1)Washington NuclearPlant-Unit2OCKETRNIBER(2)05000397LERRNIBER(8)earsaberev.Ho.6080AGE(3)5F5ITLE(4)PAILQIIEPQCQANLPIHIHATECHNICAL EPEEICAIIQNACQQNQWTHEEMERGENCY CORECOOLINGSYSTEMACTUATION INSTRUMENTATION DUETOANUNIDENTIFIED INOPERABILITY CONDITION EQQkXQ11M Therootcauseofthiseventwasaprogramdeficiency inthatnowarehouse controlswereplacedontheissuanceofthepressureswitches.
| | Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 6 08 0 4 F 5 iTLE (4) |
| Thisresultedinnoengineering analysisbeingdonetoaddresschangesbetweenpressureswitchmodelTI'ndX12designs.ThisresultedinpressureswitchesMS-PS-47A,47B,47C,and47Dnotbeinginstalled intheventedconfiguration asrequired.
| | FAILURE TO COMPLY WITH A TECHNICAL SPECIFICATION ACTION REQUIREMENT FOR THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO AN UNIDENTIFIEDINOPERABILITYCONDITION evaluation, design change evaluation, or equivalent change evaluation) performed which assessed replacement of the original pressure switch model TT with the new model. This evaluation was not performed because no warehouse controls were placed on the issuance of the pressure switches. |
| hrrr'inAreviewoftheprocurement andPMRprocesses willbeconducted andprocessimprovements willbemadeasnecessary toassurethatdisposition ofmaterials procuredforPMRswhicharelatercanceledisaddressed.
| | In 1984, Plant Modification Request (PMR) 84-1125-0 was issued to replace Static-0-Ring pressure switch model Tt'ith model 12N6BB4NXC1AJJTIX6 (X6) as the original model could no longer be procured Quality Class 1. In 1988, before any of the new model pressure switches were installed, Static-0-Ring issued a 10 CFR Part 21 notification against the X6 model pressure switch because of setpoint drift due to process permeation through the diaphragm. |
| finifIThiseventhadminimalsafetysignificance andposednothreattothehealthandsafetyofeitherthepublicorplantpersonnel.
| | Static-0-Ring subsequently replaced model X6 with model X12 to resolve the 10 CFR Part 21 concern. The only change was that the diaphragm material was changed from kapton to stainless steel. On March 9, 1989, Supply System Substitution Evaluation 567, Revision 0, was prepared to authorize model X12 as a replacement for model X6. The substitution'evaluation identified that a PMR was required for installation of the new model pressure switch. However, this requirement was not entered into the Material Management System (MMS) as a "Limitation on Use" because at the time there was no procedural requirement to do so. As an unrelated action, the substitution evaluation procedure (SPES-1, Section 7.47) was revised approximately two years later, on June 15, 1991, requiring a "Limitation on Use" (includes entry m the MMS) for items where a PMR is required for installation. |
| PressureswitchesMS-PS-47A, 47B,47C,and47DarethesensorsforHPCSsysteminitiation onhighdrywell~~ressure.TheprimarypurposeoftheHPCSsystemistomaintainreactorvesselinventory following smallreaklossofcoolantaccidents (LOCAs)thatdonotdepressurize thereactorvessel.TheHPCSsystemalsoprovidesspraycoolingheattransferduringLOCAswherethecorebecomesuncovered.
| | On August 3, 1989, PMR 84-1125-0 was canceled for unknown reasons. This effectively eliminated the requirement for an engineering evaluation of the differences between the original pressure switch model and model X6. Substitution Evaluation 567, Revision 0, only addressed the differences between pressure switch model X6 and model X12 (i.e., the change from a kapton to a stainless steel diaphragm) ~ Because the PMR was canceled and no warehouse controls were placed on issuance of the pressure switches, no engineering evaluation was performed which authorized the use of either models X6 or X12 as a replacement for pressure switches MS-PS-47A, 47B, 47C, and 47D. If the MMS had contained a useage limitation against the pressure switches requiring a PMR for installation, this limitation would have ensured that the appropriate engineering evaluation (i.e., substitution evaluation, design change evaluation, or equivalent change evaluation) was performed and adequate instructions were provided for installation. A "Limitation on Use" has been entered into the MMS to ensure that Static-0-Ring pressure switch model X12 is installed in the plant only after a proper engineering evaluation has been performed. Additionally, an engineering evaluation was completed on November 27, 1996 to verify that Static-0-Ring pressure switch model X12 is the correct model for the MS-PS-47A, 47B, 47C, and 47D application. |
| NocreditistakenfortheHPCSsystemhighdrywellpressureinitiation functioninthedesignbasisaccident(DBA)ortransient analyses.
| | The revision to the substitution evaluation procedure provides assurance that since June 15, 1991 the MMS has been updated with a "Limitation on Use" whenever a substitution requires a PMR for installation. However, there could be other cases where material was procured for a PMR prior to the procedure revision such that a PMR useage limitation was not entered into the MMS, the material was stored in the warehouse (not installed in the plant), and then the PMR was canceled. To address this possibility, a search of the Plant Tracking Log (PTL) was conducted for similar cases. No similar cases were found. A review of open and canceled PMRs will also be performed to ensure there is no material ordered for a PMR which does not have a limitation on use. |
| Thehighdrywellpressureinitiation functionisretainedforoverallredundancy anddiversity oftheHPCSfunction.
| | |
| TheHPCSsystemisassumedtobeinitiated onlowreactorvesselwaterlevelintheDBAandtransient analyses.
| | LICENSEE EVENT REPORT ( R) |
| Furthermore, basedonanalysis, duringthetimethepressureswitchventswerecapped,theHPCSsystemwouldhaveinitiated onhighdrywellpressureatapressure~2.50psig.TheTechnical Specification allowable valueforHPCSsysteminitiation onhighdrywellpressureis~1.85psigandthedesignbasisanalytical valueis~2.00psig.Boththeallowable valueandtheanalytical limitprovidesignificant margintotheprimarycontainment designpressure. | | TEXT CONTINUATION l FACILITY RNIE (1) OCKET RNIBER (2) LER RNIBER (8) AGE (3) ear saber ev. Ho. |
| imilrEvnTherehavenotbeenanyprevioussimilarreportable eventsatWNP-2involving improperuseofmaterials procuredforaPMRwhichwassubsequently cancelled.}}
| | Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 6 08 0 5 F 5 ITLE (4) |
| | PAILQIIE PQ CQANLP IHIH A TECHNICAL EPEE ICAIIQNACQQN Q W THE EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION DUE TO AN UNIDENTIFIED INOPERABILITYCONDITION EQQkXQ11M The root cause of this event was a program deficiency in that no warehouse controls were placed on the issuance of the pressure switches. This resulted in no engineering analysis being done to address changes between pressure switch model TI'nd X12 designs. This resulted in pressure switches MS-PS-47A, 47B, 47C, and 47D not being installed in the vented configuration as required. |
| | hr rr in A review of the procurement and PMR processes will be conducted and process improvements will be made as necessary to assure that disposition of materials procured for PMRs which are later canceled is addressed. |
| | f i nifI This event had minimal safety significance and posed no threat to the health and safety of either the public or plant personnel. |
| | Pressure switches MS-PS-47A, 47B, 47C, and 47D are the sensors for HPCS system initiation on high drywell ressure. The primary purpose of the HPCS system is to maintain reactor vessel inventory following small |
| | ~ |
| | reak loss of coolant accidents (LOCAs) that do not depressurize the reactor vessel. The HPCS system also provides spray cooling heat transfer during LOCAs where the core becomes uncovered. No credit is taken for |
| | ~ |
| | the HPCS system high drywell pressure initiation function in the design basis accident (DBA) or transient analyses. The high drywell pressure initiation function is retained for overall redundancy and diversity of the HPCS function. The HPCS system is assumed to be initiated on low reactor vessel water level in the DBA and transient analyses. Furthermore, based on analysis, during the time the pressure switch vents were capped, the HPCS system would have initiated on high drywell pressure at a pressure ~2.50 psig. The Technical Specification allowable value for HPCS system initiation on high drywell pressure is ~ 1.85 psig and the design basis analytical value is ~2.00 psig. Both the allowable value and the analytical limit provide significant margin to the primary containment design pressure. |
| | imil r Ev n There have not been any previous similar reportable events at WNP-2 involving improper use of materials procured for a PMR which was subsequently cancelled.}} |
|
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Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B1111997-11-10010 November 1997 LER 97-011-00:on 971010,HPCS Battery Charger Failed.Caused by Failure of a Phase Firing Control Circuit Board Due to Aging During 7 Yrs of Use.Hpcs Sys Was Immediately Declared inoperable.W/971110 Ltr ML17292B1151997-11-0707 November 1997 LER 97-010-00:on 970906,discovered That TS SR 3.4.5.1 for Identified Portion of RCS Total Leakage Would Not Be Able to Perform within Time Limits of SR 3.0.2.Caused by Inadequate Methods.Method of Meeting SR 3.4.5.1 Established ML17292B0641997-09-24024 September 1997 LER 97-004-01:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Indication of Entry Into Region a of power-to-flow Map.Caused by Inadequate Attention to Detail.Established Event Evaluation teams.W/970924 Ltr ML17292B0241997-08-18018 August 1997 LER 97-009-00:on 970717,discovered Error in Recently Performed Inservice Testing procedure,OSP-TIP/IST-R701. Caused by Procedure Inadequacy.Plant Procedure OSP/TIP/IST-R701 Will Be changed.W/970818 Ltr ML17292B0291997-08-15015 August 1997 LER 97-008-00:on 970716,wire Seal Used to Lock Containment Instrument Air Pressure Control valve,CIA-PCV-2B,found Not Intact.Cause of Misadjustment of CIA-PCV-2B Unknown.Event Will Be Communicated to Plant employees.W/970815 Ltr ML17292B0201997-08-15015 August 1997 LER 97-S01-00:on 970718,failure to Take Compensatory Measure for Inoperative Microwave Security Zone Occurred. Caused by Personnel Error.Training Will Be Conducted W/ Appropriate Members of Security force.W/970815 Ltr ML17292A9481997-07-23023 July 1997 LER 97-007-00:on 970611,voluntary Rept of Automatic Start of DG-1 & DG-2 Was Experienced.Caused by Undervoltage Condition on Electrical Busses SM-7 & SM-8.Circulating Water Pump CW-P-1C Control Switch Placed in pull-to-lock.W/970723 Ltr ML17292A9201997-06-26026 June 1997 LER 97-006-00:on 970527,non-performance of Surveillance Requirement 3.6.1.3.2 for Blind Fanges,Was Noted.Caused Because Misunderstanding of Intent of Specs.Added Five Structural Assemblies for SP.W/970626 Ltr ML17292A8331997-04-28028 April 1997 LER 97-004-00:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Entry Into Region a of power-to-flow Map Following Planned Trip of Single Mfp. Event Evaluation teams,established.W/970428 Ltr ML17292A8311997-04-28028 April 1997 LER 97-005-00:on 970327,valid Reactor Scram Signal Received Due to Low Water Level Condition During Preparations for SRV Testing.Caused by Risks & Consequences of Decisions Not Completely Identified.Restored Water level.W/970428 Ltr ML17292A8251997-04-21021 April 1997 LER 97-003-00:on 970320,notification of Noncompliance W/Ts as TS SRs for Response Time Testing Were Not Being Met for Specified Instrumentation in Rps,Pcis & Eccs.Requested Enforcement Discretion for One Time exemption.W/970421 Ltr ML17292A7431997-03-20020 March 1997 LER 97-002-00:on 970218,determined That Rod Block Monitor (RBM) Calibr Values Were Not Set IAW Tech Specs.Caused by Calibr Procedures Inadequacies.Revised & re-performed RBM Channel Calibr procedures.W/970330 Ltr ML17292A7401997-03-13013 March 1997 LER 97-001-00:on 970211,reactor Water Cleanup Sys Blowdown Flow Isolation Setpoint Was Slightly Above TS Allowable Valve Occurred Due to Calculation Error.Lds Fss LD-FS-15 LD-FS-16 Were Declared inoperable.W/970313 Ltr ML17292A6641997-01-22022 January 1997 LER 96-009-00:on 961220,miscalculation of Instantaneous Overcurrent Relay Settings Resulted in Inoperability of safety-related Equipment.Caused by Utilization of Inappropriate Design.Testing Was completed.W/970122 Ltr ML17292A6461997-01-0606 January 1997 LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr ML17292A6371996-12-19019 December 1996 LER 96-007-00:on 961122,electrical Breakers Were Not Seismically Qualified in Test/Disconnect Position.Circuit Breaker Mfg Did Not Consider Raced Out Breaker Position During Testing.Relocated Circuit breakers.W/961217 Ltr ML17292A4121996-08-0808 August 1996 LER 96-006-00:on 960709,average Power Range Monitor Rod Block Downscale Surveillance Not Performed Prior to Entry Into Mode 1.Caused by long-standing Misinterpretation of Requirements of Tss.Procedures revised.W/960808 Ltr ML17292A3801996-07-24024 July 1996 LER 96-004-00:on 960624,plant Was Manually Scrammed by Control Room Personnel Due to Reactor Water Level Transient Experienced During Testing of Digital Feedwater Sys.Caused by Programming Error.Sys Was corrected.W/960724 Ltr ML17292A3771996-07-24024 July 1996 LER 96-005-00:on 960624,determined Missed Surveillance Test Re Channel Check of Average Power Range Monitor.Caused by Inadequate Procedures.Revised Surveillance Procedure Re When APRM Checks Must Be performed.W/960724 Ltr ML17292A3641996-07-12012 July 1996 LER 96-003-00:on 960615,required Surveillance Test Not Performed When Required by TS 3.4.1.3.Caused by Inadequate Procedures.Implementing Surveillance Procedure & Reactor Plant Startup Procedures revised.W/960712 Ltr ML17292A3361996-06-20020 June 1996 LER 96-002-00:on 960504,critical Bus SM-8 Lost Power When Supply Breaker 3-8 Tripped.Caused by Personnel Error. Operators Counselled & Procedures revised.W/960620 Ltr ML17292A2861996-05-24024 May 1996 LER 96-001-00:on 960425,inadvertent ESF Actuations Occurred Due to Tripping of Temporary Power Supply to IN-3.Caused by Personnel Error.Operations Restored to IN-3 Loads & Reset ESF actuations.W/960524 Ltr ML17291B0891995-10-19019 October 1995 LER 95-011-00:on 950919,failed to Comply W/Ts SR for RCIC Sys Due to Analysis Deficiency That Resulted in Inadequate Surveillance Test Procedure.Surveillance Procedure Revised to Correct deficiency.W/951019 Ltr ML17291A9021995-07-0707 July 1995 LER 95-010-00:on 950609,HPCS DG Declared Inoperable Due to Discovery That TS Test Method Incomplete.Caused by Inadequate Testing Procedure.Test Procedure for HPCS DG Reviewed & Special Test Procedures written.W/950707 Ltr ML17291A9031995-07-0707 July 1995 LER 95-009-00:on 950607,inadvertent MSIV Closure Occurred During Surveillance Test Due to Poor Communication Between Test Team.Determined That MSIV Closure Not Valid Because Closure Not Triggered by Plant conditions.W/950707 Ltr ML17291A8501995-06-0808 June 1995 LER 95-006-01:on 950405,reactor Scram Occurred During Surveillance Testing Due to Protective Sys Relay Failure. Replaced Failed Relay Before Plant Startup ML17291A8101995-05-12012 May 1995 LER 95-008-00:on 940125,TS Wording Lead to Potential TS Violation.Caused by Lack of Clarity in Ts.Submitted Improved TS for Plant to Provide Addl clarity.W/950512 Ltr ML17291A7841995-05-0505 May 1995 LER 95-007-00:on 950222,emergency Diesel Start Occurred Due to Voltage Transient on BPA Grid.Confirmation Was Received at 17:51 H That Disturbance Had Originated in BPA Grid ML17291A7801995-05-0404 May 1995 LER 95-006-00:on 950405,main Turbine Trip Occurred During Performance of Surveillance Test Due to Protective Sys Relay Failed.Replaced Failed Relay Before Plant startup.W/950504 Ltr ML17291A7851995-05-0303 May 1995 LER 95-005-00:on 950222,inoperable IRM Had Been Relied Upon to Meet TS Requirements During Reactor Startup.Caused by Lack of Neutron Source to Test Instrumentation. Sys Knowledge Gained Will Be incorporated.W/950503 Ltr ML17291A7071995-03-25025 March 1995 LER 95-004-00:on 950226,malfunction in Main Turbine DEH Control Sys Caused All Four High Pressure Turbine Governor Valves to Rapidly Close.Caused by Blown Fuse.Suspected Faulty Circuit Card replaced.W/950325 Ltr ML17291A7011995-03-20020 March 1995 LER 95-002-00:on 950218,automatic Reactor Scram Occurred. Caused by Erroneous Positioning of Control During Performance of Scheduled Periodic Functional Test.Control repositioned.W/950320 Ltr 1999-07-20
[Table view] Category:RO)
MONTHYEARML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B1111997-11-10010 November 1997 LER 97-011-00:on 971010,HPCS Battery Charger Failed.Caused by Failure of a Phase Firing Control Circuit Board Due to Aging During 7 Yrs of Use.Hpcs Sys Was Immediately Declared inoperable.W/971110 Ltr ML17292B1151997-11-0707 November 1997 LER 97-010-00:on 970906,discovered That TS SR 3.4.5.1 for Identified Portion of RCS Total Leakage Would Not Be Able to Perform within Time Limits of SR 3.0.2.Caused by Inadequate Methods.Method of Meeting SR 3.4.5.1 Established ML17292B0641997-09-24024 September 1997 LER 97-004-01:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Indication of Entry Into Region a of power-to-flow Map.Caused by Inadequate Attention to Detail.Established Event Evaluation teams.W/970924 Ltr ML17292B0241997-08-18018 August 1997 LER 97-009-00:on 970717,discovered Error in Recently Performed Inservice Testing procedure,OSP-TIP/IST-R701. Caused by Procedure Inadequacy.Plant Procedure OSP/TIP/IST-R701 Will Be changed.W/970818 Ltr ML17292B0291997-08-15015 August 1997 LER 97-008-00:on 970716,wire Seal Used to Lock Containment Instrument Air Pressure Control valve,CIA-PCV-2B,found Not Intact.Cause of Misadjustment of CIA-PCV-2B Unknown.Event Will Be Communicated to Plant employees.W/970815 Ltr ML17292B0201997-08-15015 August 1997 LER 97-S01-00:on 970718,failure to Take Compensatory Measure for Inoperative Microwave Security Zone Occurred. Caused by Personnel Error.Training Will Be Conducted W/ Appropriate Members of Security force.W/970815 Ltr ML17292A9481997-07-23023 July 1997 LER 97-007-00:on 970611,voluntary Rept of Automatic Start of DG-1 & DG-2 Was Experienced.Caused by Undervoltage Condition on Electrical Busses SM-7 & SM-8.Circulating Water Pump CW-P-1C Control Switch Placed in pull-to-lock.W/970723 Ltr ML17292A9201997-06-26026 June 1997 LER 97-006-00:on 970527,non-performance of Surveillance Requirement 3.6.1.3.2 for Blind Fanges,Was Noted.Caused Because Misunderstanding of Intent of Specs.Added Five Structural Assemblies for SP.W/970626 Ltr ML17292A8331997-04-28028 April 1997 LER 97-004-00:on 970327,plant Operators Manually Scrammed Reactor as Required by TS Due to Entry Into Region a of power-to-flow Map Following Planned Trip of Single Mfp. Event Evaluation teams,established.W/970428 Ltr ML17292A8311997-04-28028 April 1997 LER 97-005-00:on 970327,valid Reactor Scram Signal Received Due to Low Water Level Condition During Preparations for SRV Testing.Caused by Risks & Consequences of Decisions Not Completely Identified.Restored Water level.W/970428 Ltr ML17292A8251997-04-21021 April 1997 LER 97-003-00:on 970320,notification of Noncompliance W/Ts as TS SRs for Response Time Testing Were Not Being Met for Specified Instrumentation in Rps,Pcis & Eccs.Requested Enforcement Discretion for One Time exemption.W/970421 Ltr ML17292A7431997-03-20020 March 1997 LER 97-002-00:on 970218,determined That Rod Block Monitor (RBM) Calibr Values Were Not Set IAW Tech Specs.Caused by Calibr Procedures Inadequacies.Revised & re-performed RBM Channel Calibr procedures.W/970330 Ltr ML17292A7401997-03-13013 March 1997 LER 97-001-00:on 970211,reactor Water Cleanup Sys Blowdown Flow Isolation Setpoint Was Slightly Above TS Allowable Valve Occurred Due to Calculation Error.Lds Fss LD-FS-15 LD-FS-16 Were Declared inoperable.W/970313 Ltr ML17292A6641997-01-22022 January 1997 LER 96-009-00:on 961220,miscalculation of Instantaneous Overcurrent Relay Settings Resulted in Inoperability of safety-related Equipment.Caused by Utilization of Inappropriate Design.Testing Was completed.W/970122 Ltr ML17292A6461997-01-0606 January 1997 LER 96-008-00:on 961205,failure to Comply with TS Action Requirement for Emergency Core Cooling Sys Actuation Instrumentation Occurred Due to Unidentified Inoperability Condition.Pmr initiated.W/970106 Ltr ML17292A6371996-12-19019 December 1996 LER 96-007-00:on 961122,electrical Breakers Were Not Seismically Qualified in Test/Disconnect Position.Circuit Breaker Mfg Did Not Consider Raced Out Breaker Position During Testing.Relocated Circuit breakers.W/961217 Ltr ML17292A4121996-08-0808 August 1996 LER 96-006-00:on 960709,average Power Range Monitor Rod Block Downscale Surveillance Not Performed Prior to Entry Into Mode 1.Caused by long-standing Misinterpretation of Requirements of Tss.Procedures revised.W/960808 Ltr ML17292A3801996-07-24024 July 1996 LER 96-004-00:on 960624,plant Was Manually Scrammed by Control Room Personnel Due to Reactor Water Level Transient Experienced During Testing of Digital Feedwater Sys.Caused by Programming Error.Sys Was corrected.W/960724 Ltr ML17292A3771996-07-24024 July 1996 LER 96-005-00:on 960624,determined Missed Surveillance Test Re Channel Check of Average Power Range Monitor.Caused by Inadequate Procedures.Revised Surveillance Procedure Re When APRM Checks Must Be performed.W/960724 Ltr ML17292A3641996-07-12012 July 1996 LER 96-003-00:on 960615,required Surveillance Test Not Performed When Required by TS 3.4.1.3.Caused by Inadequate Procedures.Implementing Surveillance Procedure & Reactor Plant Startup Procedures revised.W/960712 Ltr ML17292A3361996-06-20020 June 1996 LER 96-002-00:on 960504,critical Bus SM-8 Lost Power When Supply Breaker 3-8 Tripped.Caused by Personnel Error. Operators Counselled & Procedures revised.W/960620 Ltr ML17292A2861996-05-24024 May 1996 LER 96-001-00:on 960425,inadvertent ESF Actuations Occurred Due to Tripping of Temporary Power Supply to IN-3.Caused by Personnel Error.Operations Restored to IN-3 Loads & Reset ESF actuations.W/960524 Ltr ML17291B0891995-10-19019 October 1995 LER 95-011-00:on 950919,failed to Comply W/Ts SR for RCIC Sys Due to Analysis Deficiency That Resulted in Inadequate Surveillance Test Procedure.Surveillance Procedure Revised to Correct deficiency.W/951019 Ltr ML17291A9021995-07-0707 July 1995 LER 95-010-00:on 950609,HPCS DG Declared Inoperable Due to Discovery That TS Test Method Incomplete.Caused by Inadequate Testing Procedure.Test Procedure for HPCS DG Reviewed & Special Test Procedures written.W/950707 Ltr ML17291A9031995-07-0707 July 1995 LER 95-009-00:on 950607,inadvertent MSIV Closure Occurred During Surveillance Test Due to Poor Communication Between Test Team.Determined That MSIV Closure Not Valid Because Closure Not Triggered by Plant conditions.W/950707 Ltr ML17291A8501995-06-0808 June 1995 LER 95-006-01:on 950405,reactor Scram Occurred During Surveillance Testing Due to Protective Sys Relay Failure. Replaced Failed Relay Before Plant Startup ML17291A8101995-05-12012 May 1995 LER 95-008-00:on 940125,TS Wording Lead to Potential TS Violation.Caused by Lack of Clarity in Ts.Submitted Improved TS for Plant to Provide Addl clarity.W/950512 Ltr ML17291A7841995-05-0505 May 1995 LER 95-007-00:on 950222,emergency Diesel Start Occurred Due to Voltage Transient on BPA Grid.Confirmation Was Received at 17:51 H That Disturbance Had Originated in BPA Grid ML17291A7801995-05-0404 May 1995 LER 95-006-00:on 950405,main Turbine Trip Occurred During Performance of Surveillance Test Due to Protective Sys Relay Failed.Replaced Failed Relay Before Plant startup.W/950504 Ltr ML17291A7851995-05-0303 May 1995 LER 95-005-00:on 950222,inoperable IRM Had Been Relied Upon to Meet TS Requirements During Reactor Startup.Caused by Lack of Neutron Source to Test Instrumentation. Sys Knowledge Gained Will Be incorporated.W/950503 Ltr ML17291A7071995-03-25025 March 1995 LER 95-004-00:on 950226,malfunction in Main Turbine DEH Control Sys Caused All Four High Pressure Turbine Governor Valves to Rapidly Close.Caused by Blown Fuse.Suspected Faulty Circuit Card replaced.W/950325 Ltr ML17291A7011995-03-20020 March 1995 LER 95-002-00:on 950218,automatic Reactor Scram Occurred. Caused by Erroneous Positioning of Control During Performance of Scheduled Periodic Functional Test.Control repositioned.W/950320 Ltr 1999-07-20
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17284A9001999-10-31031 October 1999 Rev 0 to COLR 99-15, WNP-2 Cycle 15,COLR GO2-99-177, LER 99-S01-00:on 990903,failure to Take Compensatory Measure within Required Time Upon Failure of Isolation Zone Microwave Unit,Was Noted.Caused by Personnel Error.Provided Refresher Training on Compensatory Measures.With1999-10-0101 October 1999 LER 99-S01-00:on 990903,failure to Take Compensatory Measure within Required Time Upon Failure of Isolation Zone Microwave Unit,Was Noted.Caused by Personnel Error.Provided Refresher Training on Compensatory Measures.With ML17284A8941999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for WNP-2.With 991012 Ltr ML17284A8801999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for WNP-2.With 990910 Ltr ML17284A8691999-07-31031 July 1999 Monthly Operating Rept for July 1999 for WNP-2.With 990813 Ltr ML17292B7421999-07-20020 July 1999 LER 99-001-00:on 990628,ESF Signal Closed All Eight MSIVs While Plant Was Shutdown.Caused by Failure of Relay RPS-RLY-K10D.Subject Relay Was Replaced & Tested on 990630. with 990720 Ltr ML17292B7271999-06-30030 June 1999 Monthly Operating Rept for June 1999 for WNP-2.With 990707 Ltr ML17292B6961999-05-31031 May 1999 Monthly Operating Repts for May 1999 for WNP-2.With 990608 Ltr ML17292B6641999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for WNP-2.With 990507 Ltr ML17292B6391999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for WNP-2.With 990413 Ltr ML17292B5871999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for WNP-2.With 990311 Ltr ML17292B5571999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for WNP-2.With 990210 Ltr ML17292B5621999-01-31031 January 1999 Rev 1 to COLR 98-14, WNP-2 Cycle 14 Colr. ML17292B5341999-01-15015 January 1999 Part 21 Rept Re Incorrect Modeling of BWR Lower Plenum Vol in Bison.Defect Applies Only to Reload Fuel Assemblies Currently in Operation at WNP-2.BISON Code Model for WNP-2 Has Been Revised to Correct Error ML17292B5331999-01-15015 January 1999 Part 21 Rept Re XL-S96 CPR Correlation for SVEA-96 Fuel. Defect Applies Only to WNP-2,during Cycles 12,13 & 14 Operation.Evaluations of Defect Performed by ABB-CE ML17292B4791998-12-31031 December 1998 Washington Public Power Supply Sys 1998 Annual Rept. with 981215 Ltr ML17292B5351998-12-31031 December 1998 Monthly Operating Rept for Dec 1998 for WNP-2.With 990112 Ltr ML17292B5741998-12-31031 December 1998 WNP-2 1998 Annual Operating Rept. with 990225 Ltr ML17284A8231998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for WNP-2.With 981207 Ltr ML17284A8081998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for WNP-2.With 981110 Ltr ML17292B4451998-10-27027 October 1998 LER 98-012-01:on 980715,failure to Comply with Requirements of TS SR 3.8.4.7 Was Noted.Caused by Inadequate Work Practices.Training Session Was Held with Personnel.With 981027 Ltr ML17284A7831998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for WNP-2.With 981007 Ltr ML17284A7491998-09-10010 September 1998 WNP-2 Inservice Insp Summary Rept for Refueling Outage RF13 Spring,1998. ML17284A7561998-09-0303 September 1998 LER 98-013-00:on 980805,ESF Actuations Were Noted Due to Deenergization of Vital Electrical Bus SM-8.Caused by Inadequate Direction in Troubleshooting Plan.Will Conduct Training for Engineering Personnel.With 980903 Ltr ML17284A7571998-09-0202 September 1998 LER 98-014-00:on 980807,completion of TS 3.8.1.F Required Shutdown Due to Inoperability of EDG-2 Was Noted.Caused by Degraded Voltage Regulator for DG-2.Replaced Voltage Regulator & Associated Scrs.With 980902 Ltr ML17284A7551998-09-0202 September 1998 LER 98-015-00:on 980808,discovered Reactor Coolant Pressure Boundary Leak During Shutdown Conditions.Caused by Leakage from Socket Weld (Fwb 63) on Elbow Connection.Failed Piping Connection Was Replaced.With 980902 Ltr ML17284A7681998-08-31031 August 1998 Monthly Operating Rept for Aug 1998 for WNP-2.With 980915 Ltr ML17284A7311998-08-17017 August 1998 LER 98-012-00:on 980716,determined That 24-month SR 3.8.4.7 Had Not Been Fulfilled within Specified Frequency.Caused by Inadequate Work Practices.License Requested & Received Enforcement Discretion Re Battery Svc test.W/980817 Ltr ML17284A7261998-07-31031 July 1998 Monthly Operating Rept for July 1998 for WNP-2.W/980810 Ltr ML17284A7121998-07-23023 July 1998 LER 98-006-01:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of 10CFR50,App R Calculations for High Impedance Faults.Caused by Inadequate Work Practices.Implemented Procedural Changes ML17284A6951998-07-17017 July 1998 LER 98-011-00:on 980617,ECCS Pump Room Flooding Was Noted Due to FP Sys Pipe Break.Caused by Inadequate Design of FP Sys.Detailed Review of FP Sys Design Was Conducted. W/980717 Ltr ML17284A6961998-07-15015 July 1998 LER 98-010-00:on 980615,TS Required Shutdown Due to Inoperability of TIP Sys Isolation Valve Was Noted.Caused by Improper Installation of TIP Tubing.Reattached Affected Tubing & Inspected Other TIP tubing.W/980715 Ltr ML17284A6731998-07-0101 July 1998 LER 98-009-00:on 980606,nuclear Steam Supply Shutoff Sys Group 3 & 4 Isolations During Testing Was Noted.Caused by Procedural Deficiency.Counseled Individuals Involved in preparation.W/980701 Ltr ML17284A6751998-06-30030 June 1998 Ro:On 980617,flooding of RB Northeast Stairwell with Consequential Flooding of Two ECCS Pump Rooms.Caused by Inadequate Fire Protection Sys Design.Pumped Out Water from Affected Areas to Point Below Berm Areas of Pump Rooms ML17284A6641998-06-24024 June 1998 LER 98-008-00:on 980531,inadvertent Full Scram During RPV Leak Testing in Mode 4 Was Noted.Caused by Change in Mgt Techniques.Revised Procedures to Take Into Account Addl Water Head in Pressure Sensing lines.W/980624 Ltr ML17284A6651998-06-24024 June 1998 LER 98-007-00:on 980530,inadvertent Full Scram & Division 1 ECCS Injection Was Noted.Caused by Failure to Meet Mgt Work Practice Expectation When Encountering Deficient Procedure. Incident Review Board Convened to Review event.W/980624 Ltr ML17284A6631998-06-19019 June 1998 LER 98-006-00:on 980520,discovered Discrepancies in Low Voltage Bus Calculations During Review of App R Calculations for High Impedance Fault Analysis.Caused Indeterminate. Implemented Procedural Changes Involving Operator Action ML17284A6551998-06-0404 June 1998 LER 98-005-00:on 980506,potential for Failure of RHR Sys Valve to Close on Isolation Signal Was Noted.Caused by Design Deficiency.Caution Tag Was Placed on RHR-V-40 Control Switch to Inform Plant Operators of limitation.W/980604 Ltr ML17284A6421998-06-0101 June 1998 LER 98-004-00:on 980502,determined That Primary Containment Penetration Overcurrent Protection Does Not Meet Reg Guide 1.63 Requirements.Caused by Inadequate Design Changes. Installed Addl Fuse in RHR-MO-9 circuit.W/980601 Ltr ML17284A6491998-05-31031 May 1998 Rev 0 to COLR 98-14, WNP-2,Cycle 14 Colr. ML17292B4031998-05-31031 May 1998 Monthly Operating Rept for May 1998 for WNP-2.W/980608 Ltr ML17292B3921998-04-30030 April 1998 Monthly Operating Rept for Apr 1998 for WNP-2.W/980513 Ltr ML17292B3291998-04-0909 April 1998 LER 98-003-00:on 980311,WNP-2 Experienced SCRAM Signal as Result of Low Rpv.Caused by Less than post-SCRAM Operational Strategy for Resetting SCRAM Signal in Conditions.Changes in post-SCRAM Operational Strategy implemented.W/980409 Ltr ML17292B3281998-04-0909 April 1998 LER 98-002-00:on 980311,reactor Scram & Plant Transient Occurred,Due to Failed Closed Main Steam Isolation Valve. Caused by Loss of Pneumatic Actuating Supply Pressure. Problem Evaluation Request Written for Failure of MS-V-22D ML17292B3371998-03-31031 March 1998 Monthly Operating Rept for Mar 1998 for WNP-2.W/980409 Ltr ML17292B2641998-03-0404 March 1998 Performance Self Assessment,WNP-2. ML17292B2661998-03-0404 March 1998 LER 98-001-00:on 980203,automatic Start of HPCS EDG Was Noted.Caused by Operator Error.Operations Crew Stabilized Plant at Approximately 75% Reactor Power & Investigation of Event Was initiated.W/980304 Ltr ML17292B2911998-02-28028 February 1998 Monthly Operating Rept for Feb 1998 for WNP-2.W/980313 Ltr ML17284A7971998-02-17017 February 1998 Rev 28 to Operational QA Program Description, WPPSS-QA-004.With Proposed Rev 29 ML17292B3591998-02-12012 February 1998 WNP-2 Cycle 14 Reload Design Rept. 1999-09-30
[Table view] |
Text
CATEGORY 1 REGULRZf INFORMATION DZSTRIBUTZOlYSTEM (RIDE)
~
ACCESSION NBR:9701140016 DOC.DATE: 97/01/06 NOTARIZED: NO DOCKET FACIL:50-397
~ WPPSS Nuclear Project, Unit 2, Washington Public Powe 05000397 AUTH. NAME AUTHOR AFFILIATION PFITZER,W.A. Washington Public Power Supply System WEBRING,R.L. Washington Public Power Supply System RECIP.NAME RECIPIENT AFFILIATION
SUBJECT:
LER 96-008-00:on 961205,failure to comply with TS action requirement for emergency core cooling sys actuation instrumentation occurred due to unidentified inoperability condition.PMR will be conducted.W/970106 ltr.
DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR i ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.
4 SIZE:
E NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD4-2 PD 1 1 COLBURNFT 1 1 INTERNAL: ACRS 1 1 PD RAB 2 2 AEOD/SPD/RRAB 2 2 FILE C NTE 1 1 NRR/DE/ECGB 1 1 ELB 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DET/EIB 1 1 RGN4 FILE 01 1 D EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCEFJ H 1 1 0 NOAC MURPHYFG.A 1 1 NOAC POOREFW 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 N
T NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
FULL TEXT .ONVERSION REQUIRED TOTAL NUMBEF. OF COPIES REQUIRED: LTTR 26 ENCL 26
WASHINGTON PUBLIC POWER SUPPLY SYSTEM P.O. Box 968 ~ Richlaufi, washington 99352-0968 January 6, 1997 G02-97-003 Docket No. 50-397 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:
Subject:
NUCLEAR PLANT WNP-2, OPERATING LICENSE NPF-21 LICENSEE EV1PIT REPORT NO. 96-008-00 Transmitted herewith is Licensee Event Report No. 96-008-00 for WNP-2. This report is submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) and discusses the items of reportability, corrective action taken, and action to preclude recurrence.
Should you have any questions or desire additional information regarding this matter, please call me or Ms. L. C. Fernandez at (509) 377-4147.
Respectfully,
. L. Webring ~
Vice President, Operations Support/PIO Mail Drop PE08 RLW/CDM Enclosure
(!i CC: LJ Callan - NRC RIV JE Dyer - NRC RIV KE Perkins, Jr. - NRC RIV, Walnut Creek Field Office NS Reynolds - Winston & Strawn TG Colburn - NRR DL Williams - BPA/399 NRC Sr. Resident Inspector - 927N
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il/ il A R ~
9'70ii400i6 '770i06 PDR ADOCK 05000397 S PDR
LICENSEE EVEN EPORT (LER)
FACILITY NAHE (1) OCk.c< NUHBER (2) AGE (3)
Nhshin ton /@clear Plant - Lhit 2 0 5 0 0 0 3 9 7 1 F ITLE (4)
FAILURE TO COMPLY WITH A TECHNICAL SPECIFICATION ACTION REQUIREMENT FOR THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO UNIDENTIFIED INOPERABILITYCONDITION EVEHT DATE (5) LER NUHBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)
HOHTH DAY TEAR TEAR EQUENT IAL EV IS IOH DAY YEAR FACILITY NNlES OCKET NNlBERS(S)
UHBER UHBER 50 1 2 0 5 9 6 9 6 0 0 8 0 0 0 1 0 6 9 7 50 ERAT IHG HIS REPORT IS SUBMITTED PURSUAHT TO THE REQUIREHEHTS OF 10 CFR 5: (Check one or more of the foi oN ng) (11)
E (9)
R LEVEL 0.402(b) 0.405(C) 0.73(a)(2)(iv) .71(b)
(10) 0.405(a)(1)(i) 0.36(c)(1) 0.73(a)(2)(v) .71(c) 0.405(a)(1)(ii) 0.36(c)(2) 0.73(a)(2)(vii) THER (Specify in Abstract 0.405(a)(1)(iii) 0.73(a)(2)(i) 0.73(a)(2)(viii)(A) IoH and in Text, NRC 0.405(a)(1)(iv) 0.73(a)(2)(it) 0.73(a)(2)(viii)(B) Form 366A) 0.405(a)(1)(v) 0.73(a)(2)(iii) 0.73(a)(2)(x)
LICEHSEE CONTACT FOR THIS LER (12)
AHE TELEPHONE NNIBER W.A. Pfitzer, Technical Specialist REA C(X)E 0 9 7 7 . 2 4 1 9 COHPLETE OHE LINE FOR EACH COHPOHEHT FAILURE DESCRIBED IH THIS REPORT (13)
CAUSE SYSTEll COHPOHENT HANUFACTURER EPORTABLE heryP CAUSE STSTEH CONPONEHT HANUFACTURER EPORTABLE 0 NPRDS 0 HPRDS ~BR!
$ ;4?;f.
P S 3 8 SUPPLENEHTAL REPORT EXPECTED (14) XPECTED SUSHI SS ION HONTH DAY YEAR ATE (15)
YES (lf yes, complete EXPECTED SUBHISSIOH DATE) NO TRACf UOI On December 5, 1996 with the plant in Mode 1 at 100% reactor power, it was determined that WNP-2 may have failed to comply with a Technical Specification action requirement for the Emergency Core Cooling System (ECCS) Actuation Instrumentation. Based on subsequent analysis, it was determined that the pressure switches designed to initiate the High Pressure Core Spray (HPCS) system on high containment drywell pressure had exceeded their Technical Specification allowable values on several occasions during the period from June 10, 1996 through November 24, 1996. Contrary to ECCS Actuation Instrumentation Technical Specification 3.3.3.b, action was not taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform Action 30 of Table 3.3.3-1 because the inoperability condition had not been identified. Action 30 requires that the inoperable instrumentation channel(s) be placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or the HPCS system be declared inoperable. In accordance with Technical Specification 3.5.1, the HPCS system can be inoperable for up to 14 days before additional action is required.
The root cause of this event was a program deficiency in that no warehouse controls were placed on the issuance of the pressure switches. A Plant Modification Request (PMR) was initiated which would have prompted the appropriate engineering analysis prior to installation of the switches, but the PMR was later canceled in 1989 and no process tie existed between the PMR and the switches. This resulted in the replacement pressure switches not being installed in the vented configuration as required.
To ensure operability of the HPCS system high drywell pressure trip function, immediate corrective action was taken to vent the associated drywell pressure switches to the reactor building atmosphere and verify the setpoints in accordance with the Channel Functional Test (CFT) surveillance procedures, Further corrective actions have been completed to establish a limitation on use for the affected pressure switches and requiring an engineering evaluation be performed prior to use in other applications to ensure the replacement pressure switches are correct for the application.
This event posed no threat to the health and safety of either the public or plant personnel,
I+I LICENSEE EVENT REPORT )
TEXT CONTINUATION FACILITY NAKE (1) OCKET NUKSER (2) LER KUKSER (8) AGE (3) ear wher ev. Now Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 6 08 0 2 F 5 ITLE (4)
PAILCM!TQ COMPLY WITH A TECHNICAL APECIPICATIQN ACTION IIHQCECIIQMT PQE THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO AN UNIDERI'IFIED INOPERABILITYCONDITION vn D On December 5, 1996 with the plant in Mode 1 at 100% reactor power, it was determined that WNP-2 may have failed to comply with a Technical Specification action requirement for the Emergency Core Cooling System (ECCS) Actuation Instrumentation.
During an investigation of a setpoint drift problem related to pressure switches MS-PS-47B and 47C tPS), it was discovered that the pressure switch cases were not vented to the reactor building atmosphere as assumed in their setpoint calculation. These pressure switches are the Channel B and C sensors, respectively, for High Pressure Core Spray (HPCS) system [BG] initiation on high containment drywell pressure. Based on further investigation, it was determined that the pressure switch cases for MS-PS<7A and 47D P'S], which are the Channel A and D sensors, respectively, were also not vented to the reactor building atmosphere as assumed in the setpoint calculation. The Technical Specification trip setpoint for these pressure switches is ~ 1.65 psig and the allowable value is 6 1A85 psig.
An unvented and sealed pressure switch case is subject to pressure changes within the case due to ambient temperature variations and, because the trip setpoint and allowable value are close to atmospheric pressure, these temperature variations can create internal pressure changes which could affect the switch setpoint. At low setpoint pressures, an increase in internal case temperature will cause an increase in internal case pressure.
This increased case pressure acts against the sensed drywell pressure to shift the setpoint in the nonconservative direction such that a higher drywell pressure would be required for HPCS system initiation. Moreover, the switch setpoint can also be affected by changes in atmospheric pressure. Following each pressure switch calibration, an unvented case is effectively sealed by installation of the cover plate. With the pressure switch case unvented and sealed, a change in atmospheric pressure between calibrations will be evident by a setpoint drift observed at the next calibration. A high atmospheric pressure at the time the pressure switch case is sealed following calibration willresult in a shift in the setpoint in the nonconservative direction (higher drywell pressure required to initiate HPCS) as the atmospheric pressure within the drywell and reactor building decreases between calibrations.
On December 15, 1996, an analysis of the temperature and atmospheric pressure effects described above determined that the pressure switches for the HPCS system high drywell pressure trip function had exceeded their Technical Specification allowable values on several occasions during the period from June 10, 1996 through November 24, 1996. Contrary to ECCS Actuation Instrumentation Technical Specification 3.3.3.b, action was not taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to perform Action 30 of Table 3.3.3-1 because the inoperable condition had not been identified. Action 30 requires that the inoperable instrumentation channel(s) be placed in the, tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or the HPCS system be declared inoperable. In accordance with Technical Specification 3.5.1, the HPCS system can be inoperable for up to 14 days before additional action is required.
Imm i rr iv Acti n To ensure operability of the HPCS system high drywell pressure trip function, immediate action was taken on December 5, 1996 to vent pressure switches MS-PS-47A, 47B, 47C, and 47D to the reactor building atmosphere and verify the setpoints in accordance with the Channel Functional Test (CFT) surveillance procedures 7.4.3.3.1.53 and 7.4.3.3.1.54. The pressure switches were vented by removing the vendor installed case vent caps.
I
LICENSEE EVENT REPORT )
TEXT CONTINUATION FACILITY RAIIE (I) OCKET RLNBER (2) LER NUIIBER (8) AGE (3) eer urber ev. Ho.
Washington Nuclear Plant - Unit 2 0 0 3 9 7 0 5 0 6 08 0 3 F 5 ITLE (4)
~ AILCRE I'C COMPLY WITE A TBCIINICALPPECIPICATICN ACIICN IE!Q IBI!MENT PCR THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO AN UNIDENTIFIED INOPERABILITYCONDITION F rE 'n rr iv A in F rEv n Pursuant to 10 CFR 50.73(a)(2)(i)(B), this event is being reported as a condition prohibited by the WNP-2 Technical Specifications.
- 2. Pressure switches MS-PS-47A, 47B, 47C, and 47D were replaced during the Spring 1996 (R-11) maintenance and refueling outage as they were approaching their qualified life. The original Static-0-Ring (vendor) supplied pressure switch model (12NAASX10TI', or TT) was replaced with a new model (12N6BB4NXC1AJJTIX12, or X12) because the original model could no longer be procured Quality Class 1. The new model pressure switch differed from the original model in that the diaphragm material was changed from kapton to stainless steel and the new model included a vendor supplied integral air-tight conduit seal and case vent port. The case vent port was capped but, at the customer's option, the cap could be removed to vent the pressure switch case. The vendor provided the option to cap (unvent) the pressure switch case for those applications where the pressure switch is expected to remain functional in extreme environments. The original pressure switch did not include an integral air-tight conduit seal.
Revision 0 of the setpoint calculation for MS-PSQ7A, 47B, 47C, and 47D established the setpoint limits for the original pressure switch model TT based on the device being vented to the reactor building atmosphere. Revision 1 of the calculation addressed the changes resulting from replacement of the original pressure switch model with the new model X12. The revised calculation assumed that the new pressure switch model would be vented. However, the new pressure switch model was not installed in a vented configuration because of the presence of the integral air-tight conduit seal and the failure to uncap the vent port, Hence, the temperature and atmospheric pressure effects on the new pressure switches were larger than assumed in the setpoint calculation. Based on an analysis of these effects, it was determined that pressure switches MS-PS-47A, 47B, 47C, and 47D had exceeded their Technical Specification allowable values on several occasions during the period from June 10, 1996 through November 24, 1996.
As discussed above, the effects from changes in atmospheric pressure relative to the atmospheric pressure present at the time of calibration were introduced by the failure to uncap the vent port.
The setpoint drift problem observed following installation of the new pressure switch model has been attributed to the failure to uncap the vent port and the effects from changes in atmospheric pressure. Thus, it is believed that proper venting of the pressure switches will resolve the setpoint drift phenomenon and restore the pressure switches to reliable operation. To validate this conclusion and ensure continued operability, the setpoints for pressure switches MS-PS-47A, 47B, 47C, and 47D will be verified weekly until the pressure switches exhibit a pattern of acceptable setpoint drift in accordance with the administrative limits of the CFT surveillance procedure.
- 3. The new pressure switch model X12 was used to replace pressure switches MS-PS-47A, 47B, 47C, and 47D during the R-11 outage. The new pressure switches were installed in the field in an unvented configuration because the work order for installation and calibration did not include instructions to remove the vent cap. Furthermore, there was no explicit design document requirement to remove the vent cap because there was no engineering evaluation (substitution
LICENSEE EVENT REPORT (L )
TEXT CONTINUATION FAC!LITT HARE (1) OCKET HUHBER (2) LER HUHBER (8) AGE (3) mher ev. Ho.
Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 6 08 0 4 F 5 iTLE (4)
FAILURE TO COMPLY WITH A TECHNICAL SPECIFICATION ACTION REQUIREMENT FOR THE EMERGENCY CORE COOLING SYSTEM ACTUATIONINSTRUMENTATIONDUE TO AN UNIDENTIFIEDINOPERABILITYCONDITION evaluation, design change evaluation, or equivalent change evaluation) performed which assessed replacement of the original pressure switch model TT with the new model. This evaluation was not performed because no warehouse controls were placed on the issuance of the pressure switches.
In 1984, Plant Modification Request (PMR) 84-1125-0 was issued to replace Static-0-Ring pressure switch model Tt'ith model 12N6BB4NXC1AJJTIX6 (X6) as the original model could no longer be procured Quality Class 1. In 1988, before any of the new model pressure switches were installed, Static-0-Ring issued a 10 CFR Part 21 notification against the X6 model pressure switch because of setpoint drift due to process permeation through the diaphragm.
Static-0-Ring subsequently replaced model X6 with model X12 to resolve the 10 CFR Part 21 concern. The only change was that the diaphragm material was changed from kapton to stainless steel. On March 9, 1989, Supply System Substitution Evaluation 567, Revision 0, was prepared to authorize model X12 as a replacement for model X6. The substitution'evaluation identified that a PMR was required for installation of the new model pressure switch. However, this requirement was not entered into the Material Management System (MMS) as a "Limitation on Use" because at the time there was no procedural requirement to do so. As an unrelated action, the substitution evaluation procedure (SPES-1, Section 7.47) was revised approximately two years later, on June 15, 1991, requiring a "Limitation on Use" (includes entry m the MMS) for items where a PMR is required for installation.
On August 3, 1989, PMR 84-1125-0 was canceled for unknown reasons. This effectively eliminated the requirement for an engineering evaluation of the differences between the original pressure switch model and model X6. Substitution Evaluation 567, Revision 0, only addressed the differences between pressure switch model X6 and model X12 (i.e., the change from a kapton to a stainless steel diaphragm) ~ Because the PMR was canceled and no warehouse controls were placed on issuance of the pressure switches, no engineering evaluation was performed which authorized the use of either models X6 or X12 as a replacement for pressure switches MS-PS-47A, 47B, 47C, and 47D. If the MMS had contained a useage limitation against the pressure switches requiring a PMR for installation, this limitation would have ensured that the appropriate engineering evaluation (i.e., substitution evaluation, design change evaluation, or equivalent change evaluation) was performed and adequate instructions were provided for installation. A "Limitation on Use" has been entered into the MMS to ensure that Static-0-Ring pressure switch model X12 is installed in the plant only after a proper engineering evaluation has been performed. Additionally, an engineering evaluation was completed on November 27, 1996 to verify that Static-0-Ring pressure switch model X12 is the correct model for the MS-PS-47A, 47B, 47C, and 47D application.
The revision to the substitution evaluation procedure provides assurance that since June 15, 1991 the MMS has been updated with a "Limitation on Use" whenever a substitution requires a PMR for installation. However, there could be other cases where material was procured for a PMR prior to the procedure revision such that a PMR useage limitation was not entered into the MMS, the material was stored in the warehouse (not installed in the plant), and then the PMR was canceled. To address this possibility, a search of the Plant Tracking Log (PTL) was conducted for similar cases. No similar cases were found. A review of open and canceled PMRs will also be performed to ensure there is no material ordered for a PMR which does not have a limitation on use.
LICENSEE EVENT REPORT ( R)
TEXT CONTINUATION l FACILITY RNIE (1) OCKET RNIBER (2) LER RNIBER (8) AGE (3) ear saber ev. Ho.
Washington Nuclear Plant - Unit 2 0 5 0 0 0 3 9 7 6 08 0 5 F 5 ITLE (4)
PAILQIIE PQ CQANLP IHIH A TECHNICAL EPEE ICAIIQNACQQN Q W THE EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION DUE TO AN UNIDENTIFIED INOPERABILITYCONDITION EQQkXQ11M The root cause of this event was a program deficiency in that no warehouse controls were placed on the issuance of the pressure switches. This resulted in no engineering analysis being done to address changes between pressure switch model TI'nd X12 designs. This resulted in pressure switches MS-PS-47A, 47B, 47C, and 47D not being installed in the vented configuration as required.
hr rr in A review of the procurement and PMR processes will be conducted and process improvements will be made as necessary to assure that disposition of materials procured for PMRs which are later canceled is addressed.
f i nifI This event had minimal safety significance and posed no threat to the health and safety of either the public or plant personnel.
Pressure switches MS-PS-47A, 47B, 47C, and 47D are the sensors for HPCS system initiation on high drywell ressure. The primary purpose of the HPCS system is to maintain reactor vessel inventory following small
~
reak loss of coolant accidents (LOCAs) that do not depressurize the reactor vessel. The HPCS system also provides spray cooling heat transfer during LOCAs where the core becomes uncovered. No credit is taken for
~
the HPCS system high drywell pressure initiation function in the design basis accident (DBA) or transient analyses. The high drywell pressure initiation function is retained for overall redundancy and diversity of the HPCS function. The HPCS system is assumed to be initiated on low reactor vessel water level in the DBA and transient analyses. Furthermore, based on analysis, during the time the pressure switch vents were capped, the HPCS system would have initiated on high drywell pressure at a pressure ~2.50 psig. The Technical Specification allowable value for HPCS system initiation on high drywell pressure is ~ 1.85 psig and the design basis analytical value is ~2.00 psig. Both the allowable value and the analytical limit provide significant margin to the primary containment design pressure.
imil r Ev n There have not been any previous similar reportable events at WNP-2 involving improper use of materials procured for a PMR which was subsequently cancelled.