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| | number = ML17264B127 | | | number = ML17264B127 |
| | issue date = 12/01/1997 | | | issue date = 12/01/1997 |
| | title = LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & Cvi Signals Were Reset | | | title = LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset |
| | author name = MARTIN J T | | | author name = Martin J |
| | author affiliation = ROCHESTER GAS & ELECTRIC CORP. | | | author affiliation = ROCHESTER GAS & ELECTRIC CORP. |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:NRCFORM366(4-95)U.S.CLEARREGULATORYCOMMISSIOLICENSEEEVENTREPORT(LER)(Seereverseforrequirednumberofdigits/charactersforeachblock)I'PROVEDBYOMBNO.3150-0104EXPIRES04/30/96ESTIMATEDBURDENPERRESPONSEToCOMPLYWITHTHISMANDATORYINFORMATIONCOLLECTIONREQUEST:50.0HRS.REPORTEDLESSONSLEARNEDAREINCORPORATEDINTOTHELICENSINGPROCESSANDFEDBACKToINDUSTRY.FORWARDCOMMENTSREGARDINGBURDENESTIMATETo.THEINFORMATIONANDRECORDSMANAGEMENTBRANCHIT-6F33),U.s.NUCLEARREGULATORYCOMMISSION,WASHINGTON,DC20555-0001,ANDToTHEPAPERWORKREDUCTIONPROJECTFACILITYNAME(1)TITLE(4IR.E.GinnaNuclearPowerPlantDOCKEfNUMBER(2I05000244PAGE(3I1OF7UndetectedUnblockingofSafetyInjectionActuationSignalWhileatLowPressureCondition,DuetoFaultyBistable,ResultedinInadvertentSafetyInjectionActuationSignalMONTHDAYYEAR103197EVENTDATE(5)LERNUMBER(6)SEQUENTIALREVISIONNUMBERNUMBER97-005-00REPORTDATE(7)MONTHDAYYEAR120197FACILITYNAMEFACIL(ryNAMEOTHERFACILITIESINVOLVED(6)DOCKETNUMBERDOCKETkUMSEROPERATINGMODE(9)POWERLEVEL(10)000SUANTTOTHEREQUIREMENTSOF10CFRE:(CheckoneorTHISREPORTISSUBMITTEDPUR20.2201(b)20.2203(a)(1)20.2203(a)(2)(i)20.2203(a)(2)(ii)20.2203(a)(2)(nr)20.2203(a)(2)(iv)20.2203(a)(2)(v)20.2203(a)(3)(i)20.2203(a)(3)(ii)20.2203(a)(4)50.36(c)(1)50.36(c)(2)50.73(a)(2)(i)50.73(a)(2)(ii)50.73(a)(2)(iii)X50.73(a)(2)(iv)50.73(a)(2)(v)50.73(a)(2)(vii)more)(11)50.73(a)(2)(viii)50.73(a)(2)(x)73.71OTHERSpecifyInAbstractbeloworinNRCForm366ANAMELICENSEECONTACTFOR'THISLER(12)TELEPHONENUM8ER(IncrudeAreaCode)JohnT.St.Martin-TechnicalAssistantCOMPLETEONELINEFOREACHCOMPONENTFAILUREDES(716)771-3641CRIBEDIN'THISREPORT(13)CAUSESYSTEMCOMPONENTMANUFACTURERREPORTABLEToNPRDSCAUSESYSTEMCOMPONENTMANUFACTURERREPORTABLETONPRDSJEJSF180NOSUPPLEMENTALREPORTEXPECTED(14)YES(Ifyes,completeEXPECTEDSUBMISSIONDATE).XNOEXPECTEDSUBMISSIONDATE(15)MONTHDAYYEARABSTRACT(Limitto1400spaces,i.e.,approximately15single-spacedtypewrittenlines)(16)OnOctober31,1997,atapproximately1640EST,theplantwasinMode6withthereactorcoolantsystembeingmaintainedatatemperatureof80degreesF,andthereactorcavityfilledtogreaterthan23feet.Theplantwasshutdownforrefueling.AninadvertentautomaticSafetyInjectionActuationoccurred.Immediatecorrectiveactionwastomonitortheautomaticstartofengineeredsafetyfeaturescomponentsandsecureunneededequipment.TheunderlyingcauseoftheinadvertentautomaticSafetyInjectionActuationwasanundetectedfaultybistablecircuitboard.Duringtheperformanceofaperiodictest,thisfaultybistablecausedtheSafetyInjectionActuationsignaItobeinadvertentlyunblockedwithpressurizerpressurelessthanthesetpointforSafetyInjectionActuation.CorrectiveactiontoprecluderepetitionisoutlinedinSectionV.B.97i2050i3697i20iPDRADQCK05000244SPDRNRCFORM366(4-95) | | {{#Wiki_filter:NRC FORM 366 U.S. CLEAR REGULATORY COMMISSIO I'PROVED BY OMB NO. 3150-0104 (4-95) EXPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS. |
| NRCFORM366A(4-95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATIONU.s.NUCLEARREGULATORYCOMMISSIONFACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBER(6)YFARSEQUENTIALREVISIONNUMBERNUMBER97-005-00PAGE(3)2OF7TEXTIifmorespeceisrequired,useedditionelcopiesofNRCForm366AI(17)PRE-EVENTPLANTCONDITIONS:OnOctober31,1997,theplantwasinMode6withthereactorcoolantsystem(RCS)beingmaintainedatatemperatureofapproximately80degreesF,andthereactorcavityfilledtogreaterthan23feet.Bothresidualheatremoval(RHR)pumpswereoperating,andfuelmovementwasonhold.Twoservicewater(SW)pumpswereoperating.Theplantwasshutdownforrefueling.PerformanceMonitoringtechnicianswereconductingperiodictestprocedurePT-32.1(PlantSafeguardLogicTestAorBTrain).Testingofthe"B"trainwasstartedatapproximately1606EST.AspartoftheconductofprocedurePT-32.1,simulatedsignalsareinsertedtoprovidenormalrelaystatesformiscellaneoussignals.Priortoinsertingthesesignals,individualsfromPerformanceMonitoringandInstrumentandControl(ILC)verifiedthatindicationswereappropriateontheMainControlBoard,insafeguardsracksintheRelayRoom,andinprotectiveracksintheControlRoom.IRCtechniciansthenperformedthesesimulationsinaccordancewithprocedurePT-32.1.IRCconnectednon-poweredtransmittersimulatorstothreepressurizer(PRZR)pressurechanneltestinjectionjacks(forchannelsP-429,P-430andP-431).TestinjectionswitchesforP-429andP-430weretakentothetestpositionandsimulatoroutputwascheckedforthedesiredoutput.DESCRIPTIONOFEVENT:A.DATESANDAPPROXIMATETIMESOFMAJOROCCURRENCES'ctober31,1997,1606EST:PerformanceMonitoringstartsprocedurePT-32.1for"B"train.October31,1997,1640EST:Eventdateandtime.October31,1997,1640EST:Discoverydateandtime.October31,1997,1643EST:SafetyInjectionActuationandContainmentIsolationsignalsarereset.October31,1997,1649EST:ContainmentVentilationIsolationsignalisreset.October31,1997,1700EST:Pre-eventrefuelingshutdownconditionsarerestored.B.EVENT:OnOctober31,1997,theplantwasinMode6withtheRCSbeingmaintainedatatemperatureofapproximately80degreesF,andthereactorcavityfilledtogreaterthan23feet.Theplantwasshutdownforrefueling.TheSafetyInjection(Sl)ActuationSignal(SIAS)is,providedwithablocksignalwhichpreventsaSIASfromoccurringforLowPRZRPressure.SIASwasblocked,whichisthenormalconfigurationwhenPRZRpressureisintentionallyreducedbelowtheSIASsetpointduringplantshutdownconditions.NRCFORM366A(4.95)
| | LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK To INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE To. THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33), |
| | digits/characters for each block) U.s. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND To THE PAPERWORK REDUCTION PROJECT FACILITYNAME (1) DOCKEf NUMBER(2I PAGE (3I R.E. Ginna Nuclear Power Plant 05000244 1 OF 7 TITLE (4I Undetected Unblocking of Safety Injection Actuation Signal While at Low Pressure Condition, Due to Faulty Bistable, Resulted in Inadvertent Safety Injection Actuation Signal EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6) |
| | SEQUENTIAL FACILITY NAME DOCKET NUMBER REVISION MONTH DAY YEAR NUMBER MONTH DAY YEAR NUMBER FACIL(ry NAME DOCKET kUMSER 10 31 97 97 005 00 12 01 97 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR E: (Check one or more) (11) |
| | MODE (9) 20.2201(b) 20.2203(a)(2) (v) 50.73(a) (2) (i) 50.73(a)(2) (viii) |
| | POWER 20.2203(a)(1) 20.2203(a)(3) (i) 50.73(a) (2) (ii) 50.73(a) (2) (x) |
| | LEVEL (10) 000 20.2203(a)(2)(i) 20.2203(a) (3) (ii) 50.73(a)(2) (iii) 73.71 20.2203(a)(2)(ii) 20.2203(a) (4) X 50.73(a)(2)(iv) OTHER 20.2203(a) (2)(nr) 50.36(c)(1) 50.73(a) (2) (v) Specify In Abstract below or in NRC Form 366A 20.2203(a) (2)(iv) 50.36(c) (2) 50.73(a) (2) (vii) |
| | LICENSEE CONTACT FOR'THIS LER (12) |
| | NAME TELEPHONE NUM8ER (Incrude Area Code) |
| | John T. St. Martin - Technical Assistant (716) 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN'THIS REPORT (13) |
| | SYSTEM COMPONENT REPORTABLE REPORTABLE CAUSE MANUFACTURER To NPRDS CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS JE JS F180 NO SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSIONDATE). X NO DATE (15) |
| | ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16) |
| | On October 31, 1997, at approximately 1640 EST, the plant was in Mode 6 with the reactor coolant system being maintained at a temperature of 80 degrees F, and the reactor cavity filled to greater than 23 feet. The plant was shut down for refueling. An inadvertent automatic Safety Injection Actuation occurred. |
| | Immediate corrective action was to monitor the automatic start of engineered safety features components and secure unneeded equipment. |
| | The underlying cause of the inadvertent automatic Safety Injection Actuation was an undetected faulty bistable circuit board. During the performance of a periodic test, this faulty bistable caused the Safety Injection Actuation signaI to be inadvertently unblocked with pressurizer pressure less than the setpoint for Safety Injection Actuation. |
| | Corrective action to preclude repetition is outlined in Section V.B. |
| | 97i2050i36 97i20i PDR ADQCK 05000244 S PDR NRC FORM 366 (4-95) |
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| NRCFORM366AI4-95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATIONU.S.NUCLEARREGULATORYCOMMISSIONFACILITYNAMEI1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBERI6)YEARSEQUENTIALREVISIONNUMBERNUMBER97-005-00PAGEI3)3OF7TEXT(Ifmorespaceisrequired,useadditionalcopiesofNRCForm366A/(17)Atapproximately1640EST,I&Ctechnicianstookthethirdchannel(P-431)testinjectionswitchtothetestposition.Immediatelyaftertheswitchwasinthetestposition,simulatoroutputwasobservedtobeapproximately30milliamperes,whichsimulatespressureabovetheunblocksetpointof1992PSIG.ThisresultedinunblockingSIASfromLowPRZRPressure.ThisalsoresultedinSlactuationonLowPRZRPressure.TheControlRoomoperatorsimmediatelyrespondedtotheinadvertentSlactuation.TheyrespondedtoMainControlBoardannunciatorD-19(PressurizerLoPressSl1750PSIG).TheyconfirmedthatautomaticSlactuationhadoccurredandverifiedthatalloperableengineeredsafetyfeatures(ESF)componentsfunctionedproperly.NoimmediateactionswererequiredfortheRCS,sincethereactorcavitylevelremainedstable.TheControlRoomoperatorsreferredtoemergencyoperatingproceduresE-0(ReactorTriporSafetyInjection)andES-1.1(SlTermination)forguidanceinsecuringandrestartingequipment.BothRHRpumpscontinuedtooperate,andthetwoselectedSWpumpsstarted.The"A"emergencydieselgenerator(EDG)startedanddidnotenergizeanybusses,sinceallsafeguardsbussesremainedenergizedfromoff-sitepower.AllotheroperableESFcomponentswerealsoobservedtofunctionproperly,withtheexceptionofvalvepositionforair-operatedvalveAOV-5392(instrumentairtocontainmentisolationvalve).Thevalvepositionindicatedbothopenandclosed.However,subsequentinvestigationconfirmedthatthevalvedid,infact,traveltothefullyclosedposition.Duetoplantconditions,manyESFcomponentswerenotoperableatthestartofthisevent.The"B"EDGwasinoperableforperiodicmanufacturerinspectionandoverhaul.AllthreeSIpumpswererenderedinoperable.Therefore,the"B"EDGdidnotstartandnoSlpumpsstarted.NoinjectionflowfromtheSlpumpstotheRCSoccurred.I&CremovedthesafeguardstrainDCsupplyfusestopreventare-occurrenceofSlactuation.I&Cthencommencedtrouble-shootingforthecauseoftheSlactuation.Subsequentrootcauseanalysisrevealedthat,priortothisactivity,therehadbeenafaultinbistablecircuitboardPC-430E/F.Thisundetectedfaultresultedinthebistablede-energizing.Ade-energizedoutputresultsinbotha1/3UnblockSlsignalanda1/3LowPRZRPressureSlsignal.TheSIASforLowPRZRPressurewasautomaticallyunblockedassoonasasecondchannel(theP-431simulatedsignal)wassimulatedabovetheunblocksetpoint,sinceafirstchannel(failedchannelP-430)wasalreadyinsertinganunblocksignal.SimulatedPRZRpressure(asmeasuredbychannelP-429)wasbelowthesetpointforautomaticSIAS(thusinsertinga1/3LowPRZRPressureSlsignal).ThefaultedchannelP-430wasalsoinsertinga1/3LowPRZRPressureSlsignal.This2/3SlsignalwasblockeduntiltheP-431simulatoroutputwasinserted.Atthatinstant,the2/3signalwasunblocked,resultinginautomaticSIactuationfrom2/3LogicforPRZRpressurelessthan1750PSIG.AutomaticSlactuationcausedautomaticactuationofCNMTIsolation(Cl)andCNMTVentilationIsolation(CVI).NRCFORM366A(4.95)
| | NRC FORM 366A U.s. NUCLEAR REGULATORY COMMISSION (4-95) |
| I NRCFORM366AI4.95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATIONU.S.NUCLEARREGULATORYCOMMISSIONFACILITYNAMEI1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBERIB)YEARSEQUENTIALREVISIONNUMBERNUMBER97-005-00PAGEI3)4OF7TEXTllfmorespaceisrequired,useadditionalcopiesofIVRCForm366A/(17)INOPERABLESTRUCTURES,COMPONENTS,ORSYSTEMSTHATCONTRIBUTEDTOTHEEVENT:Trouble-shootingbyl&CidentifiedthatbistablePC-430E/Fhadablownfuse,whichcauseditsoutputstobede-energized.Nooutputfromthisbistable(de-energizedstate)providesa1/3UnblockSland1/3LowPRZRPressureSlsignal.D.OTHERSYSTEMSORSECONDARYFUNCTIONSAFFECTED:NoneE.METHODOFDISCOVERY:ThiseventwasimmediatelyapparentduetonumerousMainControlBoardAnnunciatoralarmsintheControlRoom.F.OPERATORACTION:TheControlRoomoperatorsrespondedtotheAnnunciatoralarms.TheydiagnosedthatautomaticSlactuationhadoccurred,andverifiedthatalloperableESFcomponentsfunctionedproperly.TheyobservedthedualpositionindicationforAOV-5392andconfirmedthatInstrumentAirtoCNMThadbeenproperlyisolated.TheControlRoomoperatorsresettheSIASandClandCVIsignals.Unneededequipmentwassecured.AllClvalvesandCVIcomponents(whichchangedpositionduetotheSIactuation)werereturnedtotheirpositionspriortotheevent.Pre-eventconditionswererestored.Subsequently,theControlRoomoperatorsnotifiedhighersupervisionandtheNRC.TheShiftSupervisornotifiedtheNRCper10CFR50.72(b)(2)(ii),non-emergencyfourhournotification,atapproximately1946ESTonOctober31,1997.G.SAFETYSYSTEMRESPONSES:Allsafetysystemsandcomponentsthatwereoperablerespondedasdesigned,exceptthatthevalvepositionindicationforAOV-5392indicatedbothopenandclosed.AllotheroperableESFcomponentswereobservedtofunctionproperlyaftertheautomaticSlactuation.ThisincludedautostartoftwoSWpumpsandthe"A"EDG.NRCFORM366AI4.95) | | LICENSEE EVENT REPORT (LER) |
| NRCI'ORM366A(4.95)LICENSEEEVENTREPORT(LER)TEXTCONTINUATIONU.S.NUCLEARREGULATORYCOMMISSIONFACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKET05000244LERNUMBER(6)YEARSEQUENTIALREVISIONNUMBERNUMBER97-005-00PAGE(3)5OF7TEXTIffmorespaceisrequired,useeddidonalcopiesofNRCForm386A/(17)III.CAUSEOFEVENT:A.IMMEDIATECAUSE:TheimmediatecauseoftheinadvertentautomaticSlactuationwasfrom2/3logicfromLowPRZRPressurewithSIASunblocked.B.INTERMEDIATECAUSE:TheintermediatecauseofautomaticSIASbeingunblockedwassimulatingthesignalfor'PRZRchannelP-431SlUnblocktobeabovetheunblocksetpointwithaconcurrentundetectedfaultinchannelP-430.ROOTCAUSE:TheunderlyingcausewasfaultybistablecircuitboardPC-430E/FonchannelP-430,whichfunctionstoprovide1/3UnblockSlsignalsand1/3LowPRZRPressureSlsignals.Inbothcases,thebistablede'-energizestoprovidethisfunction.Thebistable'ssupplyfusehadblownandcausedbothsignaloutputstogotozero.Thiscausedthesafetysystemtoseea1/3signalforunblockingandforinitiatingSl.WhenchannelP-431wassimulated,theoutputwasgreaterthantheunblocksetpoint,soSIASwasunblockedfrombothP-430(faulted)andP-431(simulated).ThiseventisNUREG-1022CauseCode(B),"Design,Manufacturing,Construction/Installation".IV.ANALYSISOFEVENT:Thiseventisreportableinaccordancewith10CFR50.73,LicenseeEventReportSystem,item(a)(2)(iv),whichrequiresareportof,"Anyeventorconditionthatresultedinamanualorautomaticactuationofanyengineeredsafetyfeature(ESF)".TheinadvertentautomaticSIactuationisanautomaticactuationofanESF.Anassessmentwasperformedconsideringboththesafetyconsequencesandimplicationsofthiseventwiththefollowingresultsandconclusions:TherewerenooperationalorsafetyconsequencesorimplicationsattributedtotheinadvertentautomaticSlactuationbecause:~TheplantwasinMode6(refuelingshutdownmode)withhighheadSlpumpsrenderedinoperable.TheRHRsystemwasalignedfordecayheatremoval,andsteamgenerator(SG)nozzledamswereinplace.NRCFORM366A(4-95)
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) |
| NRCFORM366A(4-95)U.S.NUCLEARREGULATORYCOMMISSIONFACILITYNAMEI1)LICENSEEEVENTREPORT(LER)TEXTCONTINUATIONDOCKETLERNUMBERI6)PAGEI3)R.E.GinnaNuclearPowerPlant05000244YEARSEQUENTIALREVISIONNUMBERNUMBER6OF797-005-00TEXT(Ifmorespaceisrequired,useadditionalcopiesofNRCForm366A/(17)~Plantconditionsprecludedanyover-pressurecondition.Theheadwasremovedfromthereactorandthereactorcavitywasfilledtogreaterthan23feet.AllSlpumpswereinoperableandtheRefuelingWaterStorageTankwasnotalignedtothesuctionoftheRHRpumps.TherewasnoadditionofwaterinventorytotheRCS.RCStemperaturecontinuedtobemaintainedstableviatheRHRsystem.Therefore,reactivitywasnotaffectedbyRCStemperaturechanges.~TheSWsystemandComponentCoolingWatersystemremainedinserviceduringthisevent,providingadequatecoolingtotheRCSandSpentFuelPool.~TheRHRcoolingcapabilityremainedinserviceandRCSintegritywasmaintained.Therefore,heatremovalfromthereactorwasassured.Basedontheabove,itcanbeconcludedthatthepublic'shealthandsafetywasassuredatalltimes.V.CORRECTIVEACTION:A.ACTIONTAKENTORETURNAFFECTEDSYSTEMSTOPRE-EVENTNORMALSTATUS:IRCdecreasedthesimulatedoutputforP-431tobelowtheSlunblocksetpoint,thusterminatingtheSIUnblocklogic,andallowingtheControlRoomoperatorstoagainblockSIAS.TheSIAS,Cl,andCVIsignalswerereset.Unneededequipmentwassecured.AllClvalvesandCVIcomponents(whichchangedpositionduetotheSlactuation)werereturnedtotheirpositionspriortotheevent.Pre-eventconditionswererestored.B.ACTIONTAKENORPLANNEDTOPREVENTRECURRENCE:AmethodtovisuallyindicateSlUnblockstatusfromPRZRpressurewillbeevaluated.Thefaultybistablecircuitboardwasreplacedwithalikeforlikereplacement,andcalibratedandtestedsatisfactorily.CPI-TRIPTEST-5.20(ReactorProtectionSystemBistableTripTest/CalibrationforChannel2(White)BistableAlarms)wascompletedforallChannel2,Rack1bistablestocheckforproperoperationandtoverifytherewerenootherbistableswithsimilarfaults.Nootherfaultybistableswereidentified.NRCFORM366AI495)
| | YFAR SEQUENTIAL REVISION NUMBER NUMBER 2 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT Iifmore speceis required, use edditionel copies of NRC Form 366AI (17) |
| NRCFORM366A(4.95)U.S.NUCLEARREGULATORYCOMMISSIONFACILITYNAME(1)LICENSEEEVENTREPORT(LER)TEXTCONTINUATIONDOCKETLERNUMBER(6)PAGE(3)R.E.GinnaNuclearPowerPlant05000244YEARSEQUENTIALREVISIONNUMBERNUMBER97-005-007OF7TEXTllfmorespaceisrequired,useadditionalcopiesofNRCForm366Al(17)VI.ADDITIONALINFORMATION:A.FAILEDCOMPONENTS:NonePREVIOUSLERsONSIMILAREVENTS'similarLEReventhistoricalsearchwasconductedwiththefollowingresults:nodocumentationofsimilarLEReventswiththesamerootcauseatGinnaStationcouldbeidentified.However,thefollowingLERsweresimilareventswithdifferentrootcauses:~LER84-006~LER85-004~LER89-003~LER95-003C.SPECIALCOMMENTS'oneNRCFORM366A(495)
| | PRE-EVENT PLANT CONDITIONS: |
| }} | | On October 31, 1997, the plant was in Mode 6 with the reactor coolant system (RCS) being maintained at a temperature of approximately 80 degrees F, and the reactor cavity filled to greater than 23 feet. Both residual heat removal (RHR) pumps were operating, and fuel movement was on hold. Two service water (SW) pumps were operating. The plant was shut down for refueling. Performance Monitoring technicians were conducting periodic test procedure PT-32.1 (Plant Safeguard Logic Test A or B Train). Testing of the "B" train was started at approximately 1606 EST. |
| | As part of the conduct of procedure PT-32.1, simulated signals are inserted to provide normal relay states for miscellaneous signals. Prior to inserting these signals, individuals from Performance Monitoring and Instrument and Control (ILC) verified that indications were appropriate on the Main Control Board, in safeguards racks in the Relay Room, and in protective racks in the Control Room. IRC technicians then performed these simulations in accordance with procedure PT-32.1. IRC connected non-powered transmitter simulators to three pressurizer (PRZR) pressure channel test injection jacks (for channels P-429, P-430 and P-431). Test injection switches for P-429 and P-430 were taken to the test position and simulator output was checked for the desired output. |
| | DESCRIPTION OF EVENT: |
| | A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES'ctober 31, 1997, 1606 EST: Performance Monitoring starts procedure PT-32.1 for "B" train. |
| | October 31, 1997, 1640 EST: Event date and time. |
| | October 31, 1997, 1640 EST: Discovery date and time. |
| | October 31, 1997, 1643 EST: Safety Injection Actuation and Containment Isolation signals are reset. |
| | October 31, 1997, 1649 EST: Containment Ventilation Isolation signal is reset. |
| | October 31, 1997, 1700 EST: Pre-event refueling shutdown conditions are restored. |
| | B. EVENT: |
| | On October 31, 1997, the plant was in Mode 6 with the RCS being maintained at a temperature of approximately 80 degrees F, and the reactor cavity filled to greater than 23 feet. The plant was shut down for refueling. |
| | The Safety Injection (Sl) Actuation Signal (SIAS) is, provided with a block signal which prevents a SIAS from occurring for Low PRZR Pressure. SIAS was blocked, which is the normal configuration when PRZR pressure is intentionally reduced below the SIAS setpoint during plant shutdown conditions. |
| | NRC FORM 366A (4.95) |
| | |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4-95) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER I6) PAGE I3) |
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 3 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT (Ifmore space is required, use additional copies of NRC Form 366A/ (17) |
| | At approximately 1640 EST, I&C technicians took the third channel (P-431) test injection switch to the test position. Immediately after the switch was in the test position, simulator output was observed to be approximately 30 milliamperes, which simulates pressure above the unblock setpoint of 1992 PSIG. This resulted in unblocking SIAS from Low PRZR Pressure. This also resulted in Sl actuation on Low PRZR Pressure. |
| | The Control Room operators immediately responded to the inadvertent Sl actuation. They responded to Main Control Board annunciator D-19 (Pressurizer Lo Press Sl 1750 PSIG). They confirmed that automatic Sl actuation had occurred and verified that all operable engineered safety features (ESF) components functioned properly. No immediate actions were required for the RCS, since the reactor cavity level remained stable. The Control Room operators referred to emergency operating procedures E-0 (Reactor Trip or Safety Injection) and ES-1.1 (Sl Termination) for guidance in securing and restarting equipment. |
| | Both RHR pumps continued to operate, and the two selected SW pumps started. The "A" emergency diesel generator (EDG) started and did not energize any busses, since all safeguards busses remained energized from off-site power. All other operable ESF components were also observed to function properly, with the exception of valve position for air-operated valve AOV-5392 (instrument air to containment isolation valve). The valve position indicated both open and closed. However, subsequent investigation confirmed that the valve did, in fact, travel to the fully closed position. |
| | Due to plant conditions, many ESF components were not operable at the start of this event. The "B" EDG was inoperable for periodic manufacturer inspection and overhaul. All three SI pumps were rendered inoperable. Therefore, the "B" EDG did not start and no Sl pumps started. No injection flow from the Sl pumps to the RCS occurred. |
| | I&C removed the safeguards train DC supply fuses to prevent a re-occurrence of Sl actuation. I&C then commenced trouble-shooting for the cause of the Sl actuation. |
| | Subsequent root cause analysis revealed that, prior to this activity, there had been a fault in bistable circuit board PC-430 E/F. This undetected fault resulted in the bistable de-energizing. A de-energized output results in both a 1/3 Unblock Sl signal and a 1/3 Low PRZR Pressure Sl signal. |
| | The SIAS for Low PRZR Pressure was automatically unblocked as soon as a second channel (the P-431 simulated signal) was simulated above the unblock setpoint, since a first channel (failed channel P-430) was already inserting an unblock signal. Simulated PRZR pressure (as measured by channel P-429) was below the setpoint for automatic SIAS (thus inserting a 1/3 Low PRZR Pressure Sl signal). The faulted channel P-430 was also inserting a 1/3 Low PRZR Pressure Sl signal. This 2/3 Sl signal was blocked until the P-431 simulator output was inserted. At that instant, the 2/3 signal was unblocked, resulting in automatic SI actuation from 2/3 Logic for PRZR pressure less than 1750 PSIG. Automatic Sl actuation caused automatic actuation of CNMT Isolation (Cl) and CNMT Ventilation Isolation (CVI). |
| | NRC FORM 366A (4.95) |
| | |
| | I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4.95) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER IB) PAGE I3) |
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT llfmore spaceis required, use additional copies of IVRC Form 366A/ (17) |
| | INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: |
| | Trouble-shooting by l&C identified that bistable PC-430 E/F had a blown fuse, which caused its outputs to be de-energized. No output from this bistable (de-energized state) provides a 1/3 Unblock Sl and 1/3 Low PRZR Pressure Sl signal. |
| | D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: |
| | None E. METHOD OF DISCOVERY: |
| | This event was immediately apparent due to numerous Main Control Board Annunciator alarms in the Control Room. |
| | F. OPERATOR ACTION: |
| | The Control Room operators responded to the Annunciator alarms. They diagnosed that automatic Sl actuation had occurred, and verified that all operable ESF components functioned properly. |
| | They observed the dual position indication for AOV-5392 and confirmed that Instrument Air to CNMT had been properly isolated. |
| | The Control Room operators reset the SIAS and Cl and CVI signals. Unneeded equipment was secured. All Cl valves and CVI components (which changed position due to the SI actuation) were returned to their positions prior to the event. Pre-event conditions were restored. |
| | Subsequently, the Control Room operators notified higher supervision and the NRC. The Shift Supervisor notified the NRC per 10CFR50.72 (b) (2) (ii), non-emergency four hour notification, at approximately 1946 EST on October 31, 1997. |
| | G. SAFETY SYSTEM RESPONSES: |
| | All safety systems and components that were operable responded as designed, except that the valve position indication for AOV-5392 indicated both open and closed. All other operable ESF components were observed to function properly after the automatic Sl actuation. This included autostart of two SW pumps and the "A" EDG. |
| | NRC FORM 366A I4.95) |
| | |
| | NRC I'ORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) |
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 5 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT Iffmore spaceis required, use eddidonal copies of NRC Form 386A/ (17) |
| | III. CAUSE OF EVENT: |
| | A. IMMEDIATECAUSE: |
| | The immediate cause of the inadvertent automatic Sl actuation was from 2/3 logic from Low PRZR Pressure with SIAS unblocked. |
| | B. INTERMEDIATE CAUSE: |
| | The intermediate cause of automatic SIAS being unblocked was simulating the signal for'PRZR channel P-431 Sl Unblock to be above the unblock setpoint with a concurrent undetected fault in channel P-430. |
| | ROOT CAUSE: |
| | The underlying cause was faulty bistable circuit board PC-430 E/F on channel P-430, which functions to provide 1/3 Unblock Sl signals and 1/3 Low PRZR Pressure Sl signals. In both cases, the bistable de'-energizes to provide this function. The bistable's supply fuse had blown and caused both signal outputs to go to zero. This caused the safety system to see a 1/3 signal for unblocking and for initiating Sl. When channel P-431 was simulated, the output was greater than the unblock setpoint, so SIAS was unblocked from both P-430 (faulted) and P-431 (simulated). |
| | This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction/Installation". |
| | IV. ANALYSIS OF EVENT: |
| | This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (iv), |
| | which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF)". The inadvertent automatic SI actuation is an automatic actuation of an ESF. |
| | An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions: |
| | There were no operational or safety consequences or implications attributed to the inadvertent automatic Sl actuation because: |
| | ~ The plant was in Mode 6 (refueling shutdown mode) with high head Sl pumps rendered inoperable. The RHR system was aligned for decay heat removal, and steam generator (SG) nozzle dams were in place. |
| | NRC FORM 366A (4-95) |
| | |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER I6) PAGE I3) |
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 6 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT (Ifmore spaceis required, use additional copies of NRC Form 366A/ (17) |
| | ~ Plant conditions precluded any over-pressure condition. The head was removed from the reactor and the reactor cavity was filled to greater than 23 feet. All Sl pumps were inoperable and the Refueling Water Storage Tank was not aligned to the suction of the RHR pumps. There was no addition of water inventory to the RCS. RCS temperature continued to be maintained stable via the RHR system. Therefore, reactivity was not affected by RCS temperature changes. |
| | ~ The SW system and Component Cooling Water system remained in service during this event, providing adequate cooling to the RCS and Spent Fuel Pool. |
| | ~ The RHR cooling capability remained in service and RCS integrity was maintained. |
| | Therefore, heat removal from the reactor was assured. |
| | Based on the above, it can be concluded that the public's health and safety was assured at all times. |
| | V. CORRECTIVE ACTION: |
| | A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: |
| | IRC decreased the simulated output for P-431 to below the Sl unblock setpoint, thus terminating the SI Unblock logic, and allowing the Control Room operators to again block SIAS. |
| | The SIAS, Cl, and CVI signals were reset. Unneeded equipment was secured. |
| | All Cl valves and CVI components (which changed position due to the Sl actuation) were returned to their positions prior to the event. Pre-event conditions were restored. |
| | B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE: |
| | A method to visually indicate Sl Unblock status from PRZR pressure will be evaluated. |
| | The faulty bistable circuit board was replaced with a like for like replacement, and calibrated and tested satisfactorily. |
| | CPI-TRIP TEST-5.20 (Reactor Protection System Bistable Trip Test/Calibration for Channel 2 (White) Bistable Alarms) was completed for all Channel 2, Rack 1 bistables to check for proper operation and to verify there were no other bistables with similar faults. No other faulty bistables were identified. |
| | NRC FORM 366A I4 95) |
| | |
| | NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95) |
| | LICENSEE EVENT REPORT (LER) |
| | TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3) |
| | YEAR SEQUENTIAL REVISION NUMBER NUMBER 7 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT llfmore space is required, use additional copies of NRC Form 366Al (17) |
| | VI. ADDITIONALINFORMATION: |
| | A. FAILED COMPONENTS: |
| | None PREVIOUS LERs ON SIMILAR EVENTS' similar LER event historical search was conducted with the following results: no documentation of similar LER events with the same root cause at Ginna Station could be identified. However, the following LERs were similar events with different root causes: |
| | ~ LER 84-006 |
| | ~ LER 85-004 |
| | ~ LER 89-003 |
| | ~ LER 95-003 C. SPECIAL COMMENTS'one NRC FORM 366A (4 95)}} |
LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were ResetML17264B127 |
Person / Time |
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Site: |
Ginna |
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Issue date: |
12/01/1997 |
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From: |
Martin J ROCHESTER GAS & ELECTRIC CORP. |
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To: |
|
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Shared Package |
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ML17264B126 |
List: |
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References |
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LER-97-005, LER-97-5, NUDOCS 9712050136 |
Download: ML17264B127 (9) |
|
Similar Documents at Ginna |
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Category:LICENSEE EVENT REPORT (SEE ALSO AO
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[Table view] Category:RO)
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Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A3671998-07-14014 July 1998 LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr ML17265A1921998-03-11011 March 1998 LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells ML17265A1641998-02-0606 February 1998 LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified ML17265A1601998-02-0606 February 1998 LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status ML17264B1441997-12-17017 December 1997 LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr ML17264B1291997-12-0303 December 1997 LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr ML17264B1271997-12-0101 December 1997 LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset ML17264B1211997-11-24024 November 1997 LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage ML17264B0461997-09-29029 September 1997 LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr ML17264B0111997-08-27027 August 1997 LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr ML17264A9941997-08-19019 August 1997 LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr ML17264A9911997-08-11011 August 1997 LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe ML17264A8271997-03-0303 March 1997 LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr ML17264A8071997-01-22022 January 1997 LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr ML17264A7471996-11-27027 November 1996 LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr ML17264A6051996-09-19019 September 1996 LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr ML17264A6061996-09-19019 September 1996 LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr ML17264A5911996-09-0505 September 1996 LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr ML17264A5921996-09-0505 September 1996 LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr ML17264A5891996-08-22022 August 1996 LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr ML17264A5781996-08-0606 August 1996 LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr ML17264A5561996-07-12012 July 1996 LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr ML17264A5421996-06-20020 June 1996 LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr ML17264A5411996-06-17017 June 1996 LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr ML17264A5051996-05-17017 May 1996 LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed ML17264A4481996-04-0808 April 1996 LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr ML17264A4471996-04-0808 April 1996 LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr ML17264A4101996-03-18018 March 1996 LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr ML17264A2971995-12-14014 December 1995 LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr ML17264A1711995-09-25025 September 1995 LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr 1999-09-22
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17265A7601999-10-0505 October 1999 Part 21 Rept Re W2 Switch Supplied by W Drawn from Stock, Did Not Operate Properly After Being Installed on 990409. Switch Returned to W on 990514 for Evaluation & Root Cause Analysis ML17265A7621999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Re Ginna Npp.With 991008 Ltr ML17265A7531999-09-23023 September 1999 Part 21 Rept Re Corrective Action & Closeout of 10CFR21 Rept of Noncompliance Re Unacceptable Part for 30-4 Connector. Unacceptable Parts Removed from Stock & Scrapped ML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7471999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Re Ginna Npp.With 990909 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Re Ginna Npp.With 990806 Ltr ML17265A7291999-07-29029 July 1999 Interim Part 21 Rept Re safety-related DB-25 Breaker Mechanism Procured from W Did Not Pas Degradatin Checks When Drawn from Stock to Be Installed Into BUS15/03A.Holes Did Not line-up & Tripper Pan Bent ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7131999-07-22022 July 1999 Special Rept:On 990407,radiation Monitor RM-14A Was Declared Inoperable.Caused by Failed Communication Link from TSC to Plant Process Computer Sys.Communication Link Was re-established & RM-14A Was Declaed Operable on 990521 ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7211999-07-19019 July 1999 ISI Rept for Third Interval (1990-1999) Third Period, Second Outage (1999) at Re Ginna Npp. ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A7661999-06-30030 June 1999 1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr ML17265A7011999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Re Ginna Npp.With 990712 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6761999-06-16016 June 1999 Part 21 Rept Re Defects & noncompliances,10CFR21(d)(3)(ii), Which Requires Written Notification to NRC on Identification of Defect or Failure to Comply. Relays Were Returned to Eaton for Evaluation & Root Cause Analysis ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17265A6681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Re Ginna Nuclear Power Plant.With 990608 Ltr ML17265A6651999-05-27027 May 1999 Interim Rept Re W2 Control Switch,Procured from W,Did Not Operate Satisfactorily When Drawn from Stock to Be Installed in Main Control Board for 1C2 Safety Injection Pump. Estimated That Evaluation Will Be Completed by 991001 ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6591999-05-17017 May 1999 Part 21 Rept Re Relay Deficiency Detected During pre-installation Testing.Caused by Incorrectly Wired Relay Coil.Relays Were Returned to Eaton Corp for Investigation. Relays Were Repaired & Retested ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6381999-05-0707 May 1999 Part 21 Rept Re Replacement Turbocharger Exhaust Turbine Side Drain Port Not Functioning as Design Intended.Caused by Manufacturing Deficiency.Turbocharger Was Reaasembled & Reinstalled on B EDG ML17265A6391999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Re Ginna Nuclear Power Plant.With 990510 Ltr ML17265A6361999-04-23023 April 1999 Part 21 Rept Re Power Supply That Did Not Work Properly When Drawn from Stock & Installed in -25 Vdc Slot.Power Supply Will Be Sent to Vendor to Perform Failure Mode Assessment.Evaluation Will Be Completed by 991001 ML17265A6301999-04-18018 April 1999 Rev 1 to Cycle 28 COLR for Re Ginna Npp. ML17265A6251999-04-15015 April 1999 Special Rept:On 990309,halon Systems Were Removed from Svc & Fire Door F502 Was Blocked Open.Caused by Mods Being Made to CR Emergency Air Treatment Sys.Continuous Fire Watch Was Established with Backup Fire Suppression Equipment ML17265A6551999-04-0909 April 1999 Initial Part 21 Rept Re Mfg Deficiency in Replacement Turbocharger for B EDG Supplied by Coltec Industries. Deficiency Consisted of Missing Drain Port in Intermediate Casing.Required Oil Drain Port Machined Open ML17265A6291999-03-31031 March 1999 Rev 0 to Cycle 28 COLR for Re Ginna Npp. ML17265A6241999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ginna Station.With 990409 Ltr ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A5661999-03-0101 March 1999 Rev 26 to QA Program for Station Operation. ML17265A5961999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ginna Nuclear Power Plant.With 990310 Ltr ML17265A5371999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Re Ginna Nuclear Power Plant.With 990205 Ltr ML17265A5951998-12-31031 December 1998 Rg&E 1998 Annual Rept. ML17265A5001998-12-21021 December 1998 Rev 26 to QA Program for Station Operation. ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4761998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Re Ginna Nuclear Power Plant.With 981210 Ltr ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4531998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Re Ginna Nuclear Power Plant.With 981110 Ltr ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A4291998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Re Ginna Nuclear Power Plant.With 981009 Ltr 1999-09-30
[Table view] |
Text
NRC FORM 366 U.S. CLEAR REGULATORY COMMISSIO I'PROVED BY OMB NO. 3150-0104 (4-95) EXPIRES 04/30/96 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.
LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK To INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE To. THE (See reverse for required number of INFORMATION AND RECORDS MANAGEMENT BRANCH IT-6 F33),
digits/characters for each block) U.s. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND To THE PAPERWORK REDUCTION PROJECT FACILITYNAME (1) DOCKEf NUMBER(2I PAGE (3I R.E. Ginna Nuclear Power Plant 05000244 1 OF 7 TITLE (4I Undetected Unblocking of Safety Injection Actuation Signal While at Low Pressure Condition, Due to Faulty Bistable, Resulted in Inadvertent Safety Injection Actuation Signal EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (6)
SEQUENTIAL FACILITY NAME DOCKET NUMBER REVISION MONTH DAY YEAR NUMBER MONTH DAY YEAR NUMBER FACIL(ry NAME DOCKET kUMSER 10 31 97 97 005 00 12 01 97 OPERATING THIS REPORT IS SUBMITTED PUR SUANT TO THE REQUIREMENTS OF 10 CFR E: (Check one or more) (11)
MODE (9) 20.2201(b) 20.2203(a)(2) (v) 50.73(a) (2) (i) 50.73(a)(2) (viii)
POWER 20.2203(a)(1) 20.2203(a)(3) (i) 50.73(a) (2) (ii) 50.73(a) (2) (x)
LEVEL (10) 000 20.2203(a)(2)(i) 20.2203(a) (3) (ii) 50.73(a)(2) (iii) 73.71 20.2203(a)(2)(ii) 20.2203(a) (4) X 50.73(a)(2)(iv) OTHER 20.2203(a) (2)(nr) 50.36(c)(1) 50.73(a) (2) (v) Specify In Abstract below or in NRC Form 366A 20.2203(a) (2)(iv) 50.36(c) (2) 50.73(a) (2) (vii)
LICENSEE CONTACT FOR'THIS LER (12)
NAME TELEPHONE NUM8ER (Incrude Area Code)
John T. St. Martin - Technical Assistant (716) 771-3641 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN'THIS REPORT (13)
SYSTEM COMPONENT REPORTABLE REPORTABLE CAUSE MANUFACTURER To NPRDS CAUSE SYSTEM COMPONENT MANUFACTURER TO NPRDS JE JS F180 NO SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSIONDATE). X NO DATE (15)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On October 31, 1997, at approximately 1640 EST, the plant was in Mode 6 with the reactor coolant system being maintained at a temperature of 80 degrees F, and the reactor cavity filled to greater than 23 feet. The plant was shut down for refueling. An inadvertent automatic Safety Injection Actuation occurred.
Immediate corrective action was to monitor the automatic start of engineered safety features components and secure unneeded equipment.
The underlying cause of the inadvertent automatic Safety Injection Actuation was an undetected faulty bistable circuit board. During the performance of a periodic test, this faulty bistable caused the Safety Injection Actuation signaI to be inadvertently unblocked with pressurizer pressure less than the setpoint for Safety Injection Actuation.
Corrective action to preclude repetition is outlined in Section V.B.
97i2050i36 97i20i PDR ADQCK 05000244 S PDR NRC FORM 366 (4-95)
NRC FORM 366A U.s. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
YFAR SEQUENTIAL REVISION NUMBER NUMBER 2 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT Iifmore speceis required, use edditionel copies of NRC Form 366AI (17)
PRE-EVENT PLANT CONDITIONS:
On October 31, 1997, the plant was in Mode 6 with the reactor coolant system (RCS) being maintained at a temperature of approximately 80 degrees F, and the reactor cavity filled to greater than 23 feet. Both residual heat removal (RHR) pumps were operating, and fuel movement was on hold. Two service water (SW) pumps were operating. The plant was shut down for refueling. Performance Monitoring technicians were conducting periodic test procedure PT-32.1 (Plant Safeguard Logic Test A or B Train). Testing of the "B" train was started at approximately 1606 EST.
As part of the conduct of procedure PT-32.1, simulated signals are inserted to provide normal relay states for miscellaneous signals. Prior to inserting these signals, individuals from Performance Monitoring and Instrument and Control (ILC) verified that indications were appropriate on the Main Control Board, in safeguards racks in the Relay Room, and in protective racks in the Control Room. IRC technicians then performed these simulations in accordance with procedure PT-32.1. IRC connected non-powered transmitter simulators to three pressurizer (PRZR) pressure channel test injection jacks (for channels P-429, P-430 and P-431). Test injection switches for P-429 and P-430 were taken to the test position and simulator output was checked for the desired output.
DESCRIPTION OF EVENT:
A. DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES'ctober 31, 1997, 1606 EST: Performance Monitoring starts procedure PT-32.1 for "B" train.
October 31, 1997, 1640 EST: Event date and time.
October 31, 1997, 1640 EST: Discovery date and time.
October 31, 1997, 1643 EST: Safety Injection Actuation and Containment Isolation signals are reset.
October 31, 1997, 1649 EST: Containment Ventilation Isolation signal is reset.
October 31, 1997, 1700 EST: Pre-event refueling shutdown conditions are restored.
B. EVENT:
On October 31, 1997, the plant was in Mode 6 with the RCS being maintained at a temperature of approximately 80 degrees F, and the reactor cavity filled to greater than 23 feet. The plant was shut down for refueling.
The Safety Injection (Sl) Actuation Signal (SIAS) is, provided with a block signal which prevents a SIAS from occurring for Low PRZR Pressure. SIAS was blocked, which is the normal configuration when PRZR pressure is intentionally reduced below the SIAS setpoint during plant shutdown conditions.
NRC FORM 366A (4.95)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER I6) PAGE I3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 3 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT (Ifmore space is required, use additional copies of NRC Form 366A/ (17)
At approximately 1640 EST, I&C technicians took the third channel (P-431) test injection switch to the test position. Immediately after the switch was in the test position, simulator output was observed to be approximately 30 milliamperes, which simulates pressure above the unblock setpoint of 1992 PSIG. This resulted in unblocking SIAS from Low PRZR Pressure. This also resulted in Sl actuation on Low PRZR Pressure.
The Control Room operators immediately responded to the inadvertent Sl actuation. They responded to Main Control Board annunciator D-19 (Pressurizer Lo Press Sl 1750 PSIG). They confirmed that automatic Sl actuation had occurred and verified that all operable engineered safety features (ESF) components functioned properly. No immediate actions were required for the RCS, since the reactor cavity level remained stable. The Control Room operators referred to emergency operating procedures E-0 (Reactor Trip or Safety Injection) and ES-1.1 (Sl Termination) for guidance in securing and restarting equipment.
Both RHR pumps continued to operate, and the two selected SW pumps started. The "A" emergency diesel generator (EDG) started and did not energize any busses, since all safeguards busses remained energized from off-site power. All other operable ESF components were also observed to function properly, with the exception of valve position for air-operated valve AOV-5392 (instrument air to containment isolation valve). The valve position indicated both open and closed. However, subsequent investigation confirmed that the valve did, in fact, travel to the fully closed position.
Due to plant conditions, many ESF components were not operable at the start of this event. The "B" EDG was inoperable for periodic manufacturer inspection and overhaul. All three SI pumps were rendered inoperable. Therefore, the "B" EDG did not start and no Sl pumps started. No injection flow from the Sl pumps to the RCS occurred.
I&C removed the safeguards train DC supply fuses to prevent a re-occurrence of Sl actuation. I&C then commenced trouble-shooting for the cause of the Sl actuation.
Subsequent root cause analysis revealed that, prior to this activity, there had been a fault in bistable circuit board PC-430 E/F. This undetected fault resulted in the bistable de-energizing. A de-energized output results in both a 1/3 Unblock Sl signal and a 1/3 Low PRZR Pressure Sl signal.
The SIAS for Low PRZR Pressure was automatically unblocked as soon as a second channel (the P-431 simulated signal) was simulated above the unblock setpoint, since a first channel (failed channel P-430) was already inserting an unblock signal. Simulated PRZR pressure (as measured by channel P-429) was below the setpoint for automatic SIAS (thus inserting a 1/3 Low PRZR Pressure Sl signal). The faulted channel P-430 was also inserting a 1/3 Low PRZR Pressure Sl signal. This 2/3 Sl signal was blocked until the P-431 simulator output was inserted. At that instant, the 2/3 signal was unblocked, resulting in automatic SI actuation from 2/3 Logic for PRZR pressure less than 1750 PSIG. Automatic Sl actuation caused automatic actuation of CNMT Isolation (Cl) and CNMT Ventilation Isolation (CVI).
NRC FORM 366A (4.95)
I NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION I4.95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER IB) PAGE I3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 4 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT llfmore spaceis required, use additional copies of IVRC Form 366A/ (17)
INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:
Trouble-shooting by l&C identified that bistable PC-430 E/F had a blown fuse, which caused its outputs to be de-energized. No output from this bistable (de-energized state) provides a 1/3 Unblock Sl and 1/3 Low PRZR Pressure Sl signal.
D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:
None E. METHOD OF DISCOVERY:
This event was immediately apparent due to numerous Main Control Board Annunciator alarms in the Control Room.
F. OPERATOR ACTION:
The Control Room operators responded to the Annunciator alarms. They diagnosed that automatic Sl actuation had occurred, and verified that all operable ESF components functioned properly.
They observed the dual position indication for AOV-5392 and confirmed that Instrument Air to CNMT had been properly isolated.
The Control Room operators reset the SIAS and Cl and CVI signals. Unneeded equipment was secured. All Cl valves and CVI components (which changed position due to the SI actuation) were returned to their positions prior to the event. Pre-event conditions were restored.
Subsequently, the Control Room operators notified higher supervision and the NRC. The Shift Supervisor notified the NRC per 10CFR50.72 (b) (2) (ii), non-emergency four hour notification, at approximately 1946 EST on October 31, 1997.
G. SAFETY SYSTEM RESPONSES:
All safety systems and components that were operable responded as designed, except that the valve position indication for AOV-5392 indicated both open and closed. All other operable ESF components were observed to function properly after the automatic Sl actuation. This included autostart of two SW pumps and the "A" EDG.
NRC FORM 366A I4.95)
NRC I'ORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 5 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT Iffmore spaceis required, use eddidonal copies of NRC Form 386A/ (17)
III. CAUSE OF EVENT:
A. IMMEDIATECAUSE:
The immediate cause of the inadvertent automatic Sl actuation was from 2/3 logic from Low PRZR Pressure with SIAS unblocked.
B. INTERMEDIATE CAUSE:
The intermediate cause of automatic SIAS being unblocked was simulating the signal for'PRZR channel P-431 Sl Unblock to be above the unblock setpoint with a concurrent undetected fault in channel P-430.
ROOT CAUSE:
The underlying cause was faulty bistable circuit board PC-430 E/F on channel P-430, which functions to provide 1/3 Unblock Sl signals and 1/3 Low PRZR Pressure Sl signals. In both cases, the bistable de'-energizes to provide this function. The bistable's supply fuse had blown and caused both signal outputs to go to zero. This caused the safety system to see a 1/3 signal for unblocking and for initiating Sl. When channel P-431 was simulated, the output was greater than the unblock setpoint, so SIAS was unblocked from both P-430 (faulted) and P-431 (simulated).
This event is NUREG-1022 Cause Code (B), "Design, Manufacturing, Construction/Installation".
IV. ANALYSIS OF EVENT:
This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2) (iv),
which requires a report of, "Any event or condition that resulted in a manual or automatic actuation of any engineered safety feature (ESF)". The inadvertent automatic SI actuation is an automatic actuation of an ESF.
An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:
There were no operational or safety consequences or implications attributed to the inadvertent automatic Sl actuation because:
~ The plant was in Mode 6 (refueling shutdown mode) with high head Sl pumps rendered inoperable. The RHR system was aligned for decay heat removal, and steam generator (SG) nozzle dams were in place.
NRC FORM 366A (4-95)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME I1) DOCKET LER NUMBER I6) PAGE I3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 6 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT (Ifmore spaceis required, use additional copies of NRC Form 366A/ (17)
~ Plant conditions precluded any over-pressure condition. The head was removed from the reactor and the reactor cavity was filled to greater than 23 feet. All Sl pumps were inoperable and the Refueling Water Storage Tank was not aligned to the suction of the RHR pumps. There was no addition of water inventory to the RCS. RCS temperature continued to be maintained stable via the RHR system. Therefore, reactivity was not affected by RCS temperature changes.
~ The SW system and Component Cooling Water system remained in service during this event, providing adequate cooling to the RCS and Spent Fuel Pool.
~ The RHR cooling capability remained in service and RCS integrity was maintained.
Therefore, heat removal from the reactor was assured.
Based on the above, it can be concluded that the public's health and safety was assured at all times.
V. CORRECTIVE ACTION:
A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:
IRC decreased the simulated output for P-431 to below the Sl unblock setpoint, thus terminating the SI Unblock logic, and allowing the Control Room operators to again block SIAS.
The SIAS, Cl, and CVI signals were reset. Unneeded equipment was secured.
All Cl valves and CVI components (which changed position due to the Sl actuation) were returned to their positions prior to the event. Pre-event conditions were restored.
B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
A method to visually indicate Sl Unblock status from PRZR pressure will be evaluated.
The faulty bistable circuit board was replaced with a like for like replacement, and calibrated and tested satisfactorily.
CPI-TRIP TEST-5.20 (Reactor Protection System Bistable Trip Test/Calibration for Channel 2 (White) Bistable Alarms) was completed for all Channel 2, Rack 1 bistables to check for proper operation and to verify there were no other bistables with similar faults. No other faulty bistables were identified.
NRC FORM 366A I4 95)
NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4.95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)
YEAR SEQUENTIAL REVISION NUMBER NUMBER 7 OF 7 R.E. Ginna Nuclear Power Plant 05000244 97 005 00 TEXT llfmore space is required, use additional copies of NRC Form 366Al (17)
VI. ADDITIONALINFORMATION:
A. FAILED COMPONENTS:
None PREVIOUS LERs ON SIMILAR EVENTS' similar LER event historical search was conducted with the following results: no documentation of similar LER events with the same root cause at Ginna Station could be identified. However, the following LERs were similar events with different root causes:
~ LER 84-006
~ LER 85-004
~ LER 89-003
~ LER 95-003 C. SPECIAL COMMENTS'one NRC FORM 366A (4 95)