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| | issue date = 08/13/1987 | | | issue date = 08/13/1987 |
| | title = LER 87-007-00:on 870714,ESF Reactor Trip Occurred Due to Undervoltage of Reactor Coolant Buses.Caused by Failure of Main Generator Voltage Control Sys.Failed & Suspected Components of Sys replaced.W/870813 Ltr | | | title = LER 87-007-00:on 870714,ESF Reactor Trip Occurred Due to Undervoltage of Reactor Coolant Buses.Caused by Failure of Main Generator Voltage Control Sys.Failed & Suspected Components of Sys replaced.W/870813 Ltr |
| | author name = SAMPSON J R, SMITH W G | | | author name = Sampson J, Smith W |
| | author affiliation = AMERICAN ELECTRIC POWER SERVICE CORP., INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG | | | author affiliation = AMERICAN ELECTRIC POWER SERVICE CORP., INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| | addressee name = | | | addressee name = |
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| =Text= | | =Text= |
| {{#Wiki_filter:REGULAYINFORNATION DISTRIBUTIO SYSTEM(RIDS)ACCESSION NBR:8708180184 DOC.DATE:87/08/13NOTARIZED: | | {{#Wiki_filter:REGULA Y INFORNATION DISTRIBUTIO SYSTEM (RIDS) |
| NODOCKETFAGIL:50-316DonaldC.CookNuclearPoeerPlant'nit2iIndiana&05000316AUTH.MANEAUTHORAFFILIATION SAMPSONeJ.R.Indiana&MichiganElectricCo.SMITHS'.G.Indiana&MichiganElectricCo.RECIP.NAMERECIPIENTAFFILIATION
| | ACCESSION NBR: 8708180184 DOC. DATE: 87/08/13 NOTARIZED: NO DOCKET FAG IL: 50-316 Donald C. Cook Nuc lear Poeer AFFILIATION P lant'nit 2i Indiana & 05000316 AUTH. MANE AUTHOR SAMPSONe J. R. Indiana & Michigan Electric Co. |
| | SMITHS'. G. Indiana & Michigan Electric Co. |
| | REC IP. NAME REC IP IENT AFFILIATION |
|
| |
|
| ==SUBJECT:== | | ==SUBJECT:== |
| LER87-007-00:
| | LER 87-007-00: on 870714'SF reactor trip occurred due to undervoltage of reactor coolant busses. Caused bg failure of main generator voltage control sos. Failed & suspected components of sos replaced. W/870813 ltr. |
| on870714'SF reactortripoccurredduetoundervoltage ofreactorcoolantbusses.Causedbgfailureofmaingenerator voltagecontrolsos.Failed&suspected components ofsosreplaced.
| | DISTRIBUTION CODE: IE22D COPIES RECEIVED: LTR ENCL SIZE: |
| W/870813ltr.DISTRIBUTION CODE:IE22DCOPIESRECEIVED: | | TITLE: 50. 73 Licensee Event Report (LER)> Incident Rpti etc. |
| LTRENCLSIZE:TITLE:50.73LicenseeEventReport(LER)>IncidentRptietc.NOTES:RECIPIENT IDCODE/NANE PD3-3LAWIGGINGTON DCOPIESLTTRENCL1111RECIPIENT IDCODE/NANE PD3-3PDCOPIESLTTRENCL11INTERNAL:
| | NOTES: |
| ACRSNICHELSON AEOD/DOAAEOD/DSP/PQAB DEDRONRR/DEST/ADS NRR/DEST/ELB NRR/DEST/MEB NRR/DEST/PSB NRR/DEST/SGB NRR/DLPG/GAB NRR/DREP/RAB NRR/PMAS/ILRB RESDEPYGIRES/DE/EIB EXTERNAL:
| | RECIPIENT COPIES RECIPIENT COPIES ID CODE/NANE LTTR ENCL ID CODE/NANE LTTR ENCL PD3-3 LA 1 1 PD3-3 PD 1 1 WIGGINGTON D 1 1 INTERNAL: ACRS NICHELSON 1 ACRS NOELLER 2 2 AEOD/DOA 1 AEOD/DSP/NAS 1 1 AEOD/DSP/PQAB 2 2 AEOD/DSP/TPAB 1 DEDRO 1 1 NRR/DEST/ADE 1 0 NRR/DEST/ADS 1 0 NRR/DEST/CEB 1 1 NRR/DEST/ELB 1 NRR/DEST/ ICSB 1 NRR/DEST/MEB 1 1 NRR/DEST/NTB '1 1 NRR/DEST/PSB 1 1 NRR/DEBT/RSB 1 NRR/DEST/SGB 1 1 NRR/DLPG/HFB 1-NRR/DLPG/GAB 1 1 NRR/DOEA/EAB 1 NRR/DREP/RAB 1 1 RlhhDRE PB 2 2 NRR/PMAS/ILRB 1 G FILE 02 1 RES DEPY GI 1 1 R FORDS J 1 1 RES/DE/EIB 1 1 RGN3 FILE 01 1 1 EXTERNAL: EG&G GROf<i N 5 5 H ST LOBBY WARD 1 LPDR 1 NRC PDR 1 NSIC HARRIS' 1 1 NSIC MAYST' 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 44 ENCL 42 |
| EG&GGROf<iNLPDRNSICHARRIS'11221110111111111111111155111ACRSNOELLERAEOD/DSP/NAS AEOD/DSP/TPAB NRR/DEST/ADE NRR/DEST/CEB NRR/DEST/
| | : r. ~ |
| ICSBNRR/DEST/NTB NRR/DEBT/RSB NRR/DLPG/HFB NRR/DOEA/EAB RlhhDREPBGFILE02RFORDSJRGN3FILE01HSTLOBBYWARDNRCPDRNSICMAYST'2211110111'1111-122111111111TOTALNUMBEROFCOPIESREQUIRED:
| | NRC FOrm 3dd U.S. NUCLEAR REOULATORY COMMISSION (943) |
| LTTR44ENCL42 r.~NRCFOrm3dd(943)LICENSEEEVENTREPORT(LER)U.S.NUCLEARREOULATORY COMMISSION APPROVEDOMBNO.31504104EXPIRES:SI31lddFACILITYNAMEPID.C.CookNuclearPlant,Unit2DOCKETNUMBER(2)PAE3l050003161oFESFActuation (ReactorTrip)DuetoUndervoltage oftheReactorCoolantPumBussesasaResultofCEVENTDATE(5)LERNUMBER(5)REPORTDATE(7)OTHERFACILITIES INVOLVED(5)MONTHDAY'9EARYEARSEOUENTrAI NUMBEtlf<<r<<<<REVISION
| | APPROVED OMB NO. 31504104 EXPIRES: SI31ldd LICENSEE EVENT REPORT (LER) |
| .>>TdNUMBERMONTHOAYYEARFACILITYNAMESDOCKETNUMBER(El 050000714878700700138705000OPERATINO MODE(9)POWERLEYEL080$<<@$.PgP.Exnan)pr)Q<<R; | | FACILITY NAME PI DOCKET NUMBER (2) PA E 3l D. C. Cook Nuclear Plant, Unit 2 0 5 0 0 0 3 1 61 oF ESF Actuation (Reactor Trip) Due to Undervoltage of the Reactor Coolant Pum Busses as a Result of C EVENT DATE (5) LER NUMBER (5) REPORT DATE (7) OTHER FACILITIES INVOLVED (5) |
| <<r<<rr20.402(B) 20.406(s)
| | MONTH DAY '9EAR YEAR SEOUENTrAI f<<r<<<<REVISION MONTH OAY YEAR FACILITYNAMES DOCKET NUMBER(El NUMBEtl .>>Td NUMBER 0 5 0 0 0 0 7 1 4 8 7 8 7 0 0 7 0 0 1 3 8 7 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE AEOUIREMENTS OF 10 CFR II: IChecfr one or more of the follory'np) (11 OPERATINO MODE (9) 20.402(B) 20.405(c) 60.73(s) (2) (iv) 73.71(B) |
| (1)(i)20.405(sI(1I(6)20.405(s) | | POWER 20.406(s) (1)(i) 50M(c)(1) 50.73(s)(2 I (v) 73.71(c) |
| (I)(ill)20.405(sI(I)IIv)20.405(s) | | LEYEL 0 8 0 20.405(s I (1 I (6) 50.35 (c) (2) 50.73(s) (2) (rd) OTHER ISpeclfy ln Aprtrect INfow end In Text, NIIC form |
| (1I(v)20.405(c) 50M(c)(1) 50.35(c)(2)60.73(s)(2)(I)50.73(s)(2)(Q) 60.73(s)(2)(ill)LICENSEECONTACTFORTHISLER(12)60.73(s)(2)(iv)50.73(s)(2 I(v)50.73(s)(2)(rd)50,73(s)(2)(vill)IA)50.73(s)(2)(rlii)(BI 60.73(s)(2)(x) | | $<<@$ .PgP |
| THISREPORTISSUBMITTED PURSUANTT0THEAEOUIREMENTS OF10CFRII:IChecfroneormoreofthefollory'np)
| | <<r . E xnan)pr)Q<<R; 20.405(s) (I ) (ill) 60.73(s) (2) (I) 50,73(s) (2) (vill)IA) 366AI |
| (1173.71(B)73.71(c)OTHERISpeclfylnAprtrectINfowendInText,NIICform366AINAMEJ.R.Sampson-Safet6Assessment SuerinteTELEPHONE NUMBERAREACODE616465-590 COMPLETEONELINEFOREACHCOMPONENT FAILUREOESCRIBEO INTHISREPORT(13ICAUSESYSTEMCOMPONENT MANUFAC.TURERREPORTABLE
| | << 20.405(s I (I) I Iv) 50.73(s)(2)(Q) 50.73(s) (2)(rlii)(BI rr 20.405(s) (1I(v) 60.7 3(s) (2) (ill) 60.73(s)(2)(x) |
| '@'P,CAUSESYSTEMx>gjcQ.yg<<>S>>rTONPROSr.',<<,rrrrCOMPONENT MANUFAC.TUREREPORTABLETONPADSXEXELUJX34553%@$i"'6"4CR3455NrrMArXEL773455N3455N'.je~P~Nx4+>>~~KKM SUPPLEMENTAL REPORTEXPFCTED(14)YESIlfyer,complereEXPECTEDS(fphtlSSION OATEIXNoABSTRACTILlmittote00rpecer,Ie.,epproxlmerely llfteentlnplespecstypevrritren finer)(15)EXPECTEDSUBMISSIONDATE(15)MONTHDAYYEAROnJuly14,1987,at0707hours,anEngineered SafetyFeaturesActuation (ReactorTrip)occurredduetoundervoltage ofthereactorcoolantpumpbusses.Theundervoltage condition wastheresultofthefailureofthemaingenerator voltagecontrolsystem.TheUnitwasstabilized inMode3(HotStandby)atapproximately 0755hours,July14,1987.Noabnormalreactortripsequenceresponses werenoted.TheNRCwasnotifiedoftheeventviatheENSat0815hours,July14,1987.Post-event testingofthevoltagecontrolsystemcomponents revealedtwofailedpowersupplies(automatic andmanual),andsixfailedSCRModules.Allfailed/suspected components werereplaced.
| | LICENSEE CONTACT FOR THIS LER (12) |
| Investigation regarding industryexperience withBrownBoverivoltagecontrolsystemsindicates thatthefailureswhichoccurredareconsistent withthoseexperienced atotherplantsandisnotaresultofoperatorerrororlackofmaintenance. | | NAME TELEPHONE NUMBER AREA CODE J. R. Sampson Safet 6 Assessment Su erinte COMPLETE ONE LINE FOR EACH COMPONENT FAILURE OESCRIBEO IN THIS REPORT (13I 616465-590 x>gjc P, |
| Evaluation astothenecessity fortheredesignofthevoltagecontrolsystem,viathePlantModification progra'm, continues however,nospecificpreventive measureshavebeenidentified todate.8708180184 870813PDRADOCI(r0500031(EI 8PDRNRCForm3dd(9d3) | | Q.yg <<>S>>r CAUSE SYSTEM MANUFAC. REPORTABLE '@ MANUFAC. EPORTAB LE COMPONENT TO NPROS CAUSE SYSTEM COMPONENT TURER r TURER TO NPADS |
| NRCForm366A(()4)3)LICENSEEETREPORT(LER)TEXTCONTINUAU.S.NUCLEARREGULATORY COMMISSION APPROVEDOMBNO.3150-0104 EXPIRES:8/31/88FACILITYNAME(1)DOCKETNUMBER(2)LERNUMBER(6)YEAR8'jets'EGUENTIAL 4.NUMSE4REVISIONNOMEE4PAGE(3)D.C.CookNuclearPlant,Unit2ppppp31TEXT/FIINvoEPEcoirroqoiaEoooaANPenelHRCForm355AB/((1)870070002oF0Conditions PriortoOccurrence Unit2inMode1(poweroperations) at80percentReactorThermalPower.DescritionofEventOnJuly14,1987,at0707hours,anEngineered Actuation (ReactorTripSequence) occurreddueofthereactorcoolantpumpbusses(EIIS/EA).
| | <<, rrrr.', |
| condition wastheresultofthefailureofthevoltage.controlsystem(EIIS/EL)
| | X E U J X 3 4 5 5 rr MAr i "'6 "4 X E L 7 7 3 4 5 5 N je~P~ |
| .SafetyFeaturestoundervoltage Theundervoltage maingenerator Atapproximately 0637hours,July14,1987,theincomingUnit,Supervisor (Utility-Licensed Operator) noticedthattheoutputfromthemaingenerator automatic voltageregulator (EIIS/EL-RG) wasnotnulledwiththemanualvoltageregulator (EIIS/EL-RG).
| | Nx4+>>~~KKM X E L CR3455N 3%@$ |
| Theautomatic regulator wascurrently inserviceandhadbeenoperating atthelowestvalueofthevoltagebandthroughout thepreviousshift.WhentheUnitSupervisor adjustedthemanualregulator tomatchtheautomatic
| | SUPPLEMENTAL REPORT EXPFCTED (14) 3 4 5 5 N MONTH DAY YEAR EXPECTED SUB M I SS ION DATE (15) |
| : setpoint, themaingenerator voltageandexcitercurrentindication dippedmomentarily andreturnedtonormal[theadjustment.
| | YES Ilfyer, complere EXPECTED S(fphtlSSION OATEI X No ABSTRACT ILlmit to te00 rpecer, I e., epproxlmerely llfteen tlnple specs typevrritren finer) (15) |
| shouldhavehadnoeffectonthemaingenerator (EIIS/EL-GEN)].
| | On July 14, 1987, at 0707 hours, an Engineered Safety Features Actuation (Reactor Trip) occurred due to undervoltage of the reactor coolant pump busses. The undervoltage condition was the result of the failure of the main generator voltage control system. |
| Hethenrequested thataReactorOperator(RO)(Utility-Licensed Operator) bepostedat.theregulator tomonitortheindication.
| | The Unit was stabilized in Mode 3 (Hot Standby) at approximately 0755 hours, July 14, 1987. No abnormal reactor trip sequence responses were noted. The NRC was notified of the event via the ENS at 0815 hours, July 14, 1987. |
| TheROreviewedthegenerator panelstatuswiththeincomingUnit,Supervisor andnoticedthat.thegenerator outputvoltagewasslightlybelowtheminimumoftheoperating band.TheROincreased theautomatic regulator torestorethegenerator outputvoltagetothebottomoftheoperating band.Fortheremaining periodpriortothetripsequence, thevolt-amps
| | Post-event testing of the voltage control system components revealed two failed power supplies (automatic and manual), and six failed SCR Modules. All failed/suspected components were replaced. Investigation regarding industry experience with Brown Boveri voltage control systems indicates that the failures which occurred are consistent with those experienced at other plants and is not a result of operator error or lack of maintenance. |
| : reactive, generator fieldtemperature, andtoalesserdegreethegenerator outputvoltageindication showedunstablebehavior.
| | Evaluation as to the necessity for the redesign of the voltage control system, via the Plant Modification progra'm, continues however, no specific preventive measures have been identified to date. |
| Noalarmsoccurreduntilsecondsbeforethetripsequence.
| | 8708180184 870813 PDR ADOCI(r 0500031(EI 8 PDR NRC Form 3dd (9d3) |
| Following thetripsequence[openingofthereactortripbreakers(EIIS/JE-BKR),
| |
| insertion ofthereactorcontrolrods(EIIS/AA-ROD),
| |
| feedwater isolation (EIIS/JB),
| |
| automatic startingofthemotor-drivenandturbine-driven auxiliary feedwater pumps(EIIS/BA-P)]
| |
| operations personnel immediately implemented thespecialEmergency Operating Procedure, 1OHP-4023.E-O, toverifyproperresponseoftheautomatic protection system(EIIS/JC) andtoassessplantconditions forinitiating appropriate recoveryactions.Therewasnoautomatic ormanualactuation oftheintermediate headsafetyinjection system(EIIS/BQ)
| |
| .NACFOAM366A(84)3)oU.S.GPO:1988 0.824538/455
| |
|
| |
|
| NRCForm366A(94)31LICENSEEETREPORT(LER)TEXTCONTINUAU.S.NUCLEARREGULATORY COMMISSION APPROVEDOMBNO.3150&104EXPIRES:8/31/88FACILITYNAME(I)DOCKETNUMBER(2)YEARLERNUMBER(6)WSEOUENTIAL NUMEERNUMBERPAGE(3)D.C.CookNuclearPlant,Unit20s00031687TEXT/SmoteEPooo)rnrqrr)or/,
| | NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION |
| rroo/Ar/orro/HRC Forrrr35549/(IT)-007-003oF05Thevoltagetransient wasalsodetectedbytheT21Dsafeguards bus(train"A")undervoltage relays(EIIS/EA-27) whichstartedtheCDemergency dieselgenerator (EIIS/EK-DG) andinitiated loadsheddingofthe"A"trainsafeguards busses.TheRB"trainbusseswereapproximately 70to100voltshigherthanthe"A"trainbussesimmediately priortothetrip-consequently theloadsheddingactuation logicforthe"BRtrainwasnotsatisfied, andtheABemergency dieselgenerator didnot,automatically start.TheEastcentrifugal chargingpump(EIIS/CB-P) hadbeenoperating andwasnotrestoredduringblackoutloadsequencing (asperdesign).Operators manuallystartedtheWestcentrifugal chargingpumptorestorereactorcoolantpumpsealinjection (EIIS/CB) andopenedIM0-911,chargingpumpsuctionfromtherefueling waterstoragetank(EIIS/CB-TK), | | (()4)3) |
| torestorepressurizer levelandtocleartheletdownisolation (whichoccurred, asdesigned, duetothelossoftheEastcentrifugal chargingpump).Theloadshedalsode-energized thelightingtransformers (EIIS/FF-XFMR),
| | LICENSEE E T REPORT (LER) TEXT CONTINUA APPROVED OMB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) |
| isolating plantlightingandoutlets(EIIS/FF-OUT). | | YEAR 8'jets'EGUENTIAL REVISION 4 . NUMSE4 NOME E 4 D.C. Cook Nuclear Plant, Unit 2 p p p p p 3 1 8 7 0 0 7 0 0 0 2oF0 TEXT /F IINvo EPEco ir roqoiaE ooo aANPenel HRC Form 355AB/ ((1) |
| Otherlightingtransformer loadslostincluded; theanalogrodpositionindication (EIIS/AA-XF) andtherodbottomlights(EIIS/AA-IL),
| | Conditions Prior to Occurrence Unit 2 in Mode 1 (power operations) at 80 percent Reactor Thermal Power. |
| chartdrivesforthenarrowrangeandwiderangereactorcoolanttemperature recorder(EIIS/JC-TR),
| | Descri tion of Event On July 14, 1987, at 0707 hours, an Engineered Safety Features Actuation (Reactor Trip Sequence) occurred due to undervoltage of the reactor coolant pump busses (EIIS/EA). The undervoltage condition was the result of the failure of the main generator voltage. control system (EIIS/EL) . |
| maingenerator megawattrecorderchartdrive(EIIS/EL-XR),
| | At approximately 0637 hours, July 14, 1987, the incoming Unit, Supervisor (Utility-Licensed Operator) noticed that the output from the main generator automatic voltage regulator (EIIS/EL-RG) was not nulled with the manual voltage regulator (EIIS/EL-RG). |
| normalcontrolroomlighting(EIIS/FF)E andtheABemergency dieselgenerator recorderchartdrive(EIIS/EK-XR).
| | The automatic regulator was currently in service and had been operating at the lowest value of the voltage band throughout the previous shift. When the Unit Supervisor adjusted the manual regulator to match the automatic setpoint, the main generator voltage and exciter current indication dipped momentarily and returned to normal [the adjustment. should have had no effect on the main generator (EIIS/EL-GEN)]. He then requested that a Reactor Operator (RO) (Utility-Licensed Operator) be posted at. |
| Theloadsheddingsequenceexperienced wasinaccordance withundervoltage protection system(EIIS/EA)design.
| | the regulator to monitor the indication. The RO reviewed the generator panel status with the incoming Unit, Supervisor and noticed that. the generator output voltage was slightly below the minimum of the operating band. The RO increased the automatic regulator to restore the generator output voltage to the bottom of the operating band. For the remaining period prior to the trip sequence, the volt-amps reactive, generator field temperature, and to a lesser degree the generator output voltage indication showed unstable behavior. No alarms occurred until seconds before the trip sequence. |
| Powerwasmanuallly restoredtothelightingtransformers within30minutes.TheABemergency dieselgenerator wasmanuallystarted45secondsafterthetrip,butwasneverloaded.Bothemergency dieselsweresecuredafteroffsitepowerwasestablished tothe"A"trainsafeguards busses.TheUnitwasstabilized inMode3(hotstandby)atapproximately 0755hours,July14,1987.TheNRCwasnotifiedoftheeventviatheENSat0815hours,July14,1987.Withtheexception ofthefailureofthemaingenerator voltagecontrolsystem,therewerenoinoperative structures, components, orsystemsthatcontributed tothisevent.CauseofEventThecauseoftheeventwasdetermined tobethefailureofthemaingenerator automatic voltageregulator powersupplyconcurrent withthefailureofthemanualvoltageregulator powersupply.Post-event testingofvoltagecontrolsystemcomponents revealedsixfailedSCRmodules,inadditiontothetwofailedpowersupplies(automatic and.manual).Intermittent failuresofthemaingenerator automatic voltageregulator pulseamplifier andpulsedrivercircuitboardsaretheprobablecauseoftheSCRmodulefailures.
| | Following the trip sequence [opening of the reactor trip breakers (EIIS/JE-BKR), insertion of the reactor control rods (EIIS/AA-ROD), |
| NRCFORM366A(94)3)*U.S.GPO:1986-0.824 538/455 NRCForm3BBA(94)3)LICENSEEETREPORT(LER)TEXTCONTINUAT JU.S.NUCLEARREGULATORY COMMISSION APPROVEOOMBNO.3150-0104 EXPIRES:8/31/88FACILITYNAME(1)DOCKETNUMBER(2)LERNUMBER(8)SEQUENTIAL NUMBERREVISIONNUMBERPAGE(3)D.C.CookNuclearPlant,Unit2TEXT//Fmoro<<>>oo/4Er)U)or/,
| | feedwater isolation (EIIS/JB), automatic starting of the motor-driven and turbine-driven auxiliary feedwater pumps (EIIS/BA-P)] |
| I>>o////or>>/NRC | | operations personnel immediately implemented the special Emergency Operating Procedure, 1 OHP -4023.E-O, to verify proper response of the automatic protection system (EIIS/JC) and to assess plant conditions for initiating appropriate recovery actions. There was no automatic or manual actuation of the intermediate head safety injection system (EIIS/BQ) . |
| %%drm3654'4/(17)050003180070004"05AnalsisofEventThisEngineered SafetyFeaturesActuation, whichresultedinareactortripsequence, isreportable pursuant. | | NAC FOAM 366A o U.S.GPO:1988 0.824 538/455 (84)3) |
| to10CFR50.73(a)(2)(iv).TheOperations SequenceMonitorfunctioned asdesigned.
| | |
| Atimestudyofparameters monitored concluded thatallautomatic protection systemresponses; reactortrip,andresulting actuations, functioned properlyasaresultoftheEngineered SafetyFeaturesactuation.
| | NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION (94)31 LICENSEE E T REPORT (LER) TEXT CONTINUA APPROVED OMB NO. 3150&104 EXPIRES: 8/31/88 FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) |
| Basedontheabove,itisconcluded thattheeventdidnotconstitute anunreviewed safetyquestionasdefinedin10CFR50.59(a)(2),nordiditadversely impacthealthandsafety.Corrective ActionsImmediate corrective actioninvolvedoperations personnel implementing plantprocedures toverifyproperresponseoftheautomatic protection systemandtoassessplantconditions forinitiating appropriate recoveryactions.Allfailed/suspected components werereplaced(automatic andmanualvoltageregulator powersupplies, sixSCRmodules,twopulsegenerator circuitboardsandtwopulseamplifier circuit.boards).Investigation regarding industryexperience withBrownboverivoltagecontrolsystemsindicates thatthefailureswhichoccurredareconsistent withthoseexperienced atotherplantsandisnot.theresultofoperatorerrororlackofmaintenance.
| | YEAR W SEOUENTIAL NUMEER NUMBER D.C. Cook Nuclear Plant, Unit 2 0 s 0 0 0 3 1 6 8 7 0 0 7 0 0 3oF 0 5 TEXT /Smote EPooo )r nrqrr)or/, rroo/Ar/orro/HRC Forrrr 35549/ (IT) |
| Evaluation astothenecessity fortheredesignofthevoltagecontrolsystem,viathePlantModification program,continues | | The voltage transient was also detected by the T21D safeguards bus (train "A") undervoltage relays (EIIS/EA-27) which started the CD emergency diesel generator (EIIS/EK-DG) and initiated load shedding of the "A" train safeguards busses. The RB" train busses were approximately 70 to 100 volts higher than the "A" train busses immediately prior to the trip consequently the load shedding actuation logic for the "BR train was not satisfied, and the AB emergency diesel generator did not, automatically start. The East centrifugal charging pump (EIIS/CB-P) had been operating and was not restored during blackout load sequencing (as per design). |
| -however,nospecificpreventive measureshavebeenidentified todate.NRCFORMSSEA(94)3)rrU.S.GPO;198B 0-B24538/455 NRCForm3BSA(843)LICENSEETREPORT(LER)TEXTCONTINUAU.S.NUCLEARREGULATORY COMMISSION APPROVEOOM8NO.3150-0I04 EXPIRES:8/31/88FACILITYNAMEll)DOCKETNUMBER(2)YEARLFRNUMBER(8)SEOUENTIAL NUMBER'.RSREVISIONNUMBERPAGE(3)D.C.CookNuclearPlant,Unit2TExT///'oroBPoco/Br/v/rorEIrroo//I/mr/NRc%%drrrr3()343/lIT)050003]870070005OF05FailedComonentIdentification PlantDescription:
| | Operators manually started the West centrifugal charging pump to restore reactor coolant pump seal injection (EIIS/CB) and opened IM0-911, charging pump suction from the refueling water storage tank (EIIS/CB-TK), to restore pressurizer level and to clear the letdown isolation (which occurred, as designed, due to the loss of the East centrifugal charging pump). The load shed also de-energized the lighting transformers (EIIS/FF-XFMR), isolating plant lighting and outlets (EIIS/FF-OUT). Other lighting transformer loads lost included; the analog rod position indication (EIIS/AA-XF) and the rod bottom lights (EIIS/AA-IL), |
| Manufacturer: | | chart drives for the narrow range and wide range reactor coolant temperature recorder (EIIS/JC-TR), main generator megawatt recorder chart drive (EIIS/EL-XR), normal control room lighting (EIIS/FF)E and the AB emergency diesel generator recorder chart drive (EIIS/EK-XR). The load shedding sequence experienced was in accordance with undervoltage protection system (EIIS/EA)design. |
| Manufacturer IDNumber:EIISCode:Numberreplaced: | | Power was manuallly restored to the lighting transformers within 30 minutes. The AB emergency diesel generator was manually started 45 seconds after the trip, but was never loaded. Both emergency diesels were secured after offsite power was established to the "A" train safeguards busses. The Unit was stabilized in Mode 3 (hot standby) at approximately 0755 hours, July 14, 1987. |
| MainGenerator VoltageRegulator PowerSuppliesBrownBoveriVRCE1R39287116 AR103963R1 ND-'501BEL-UJX2PlantDescription:
| | The NRC was notified of the event via the ENS at 0815 hours, July 14, 1987. |
| Manufacturer:
| | With the exception of the failure of the main generator voltage control system, there were no inoperative structures, components, or systems that contributed to this event. |
| Manufacturer IDnumber:EIISCode:NumberReplaced:
| | Cause of Event The cause of the event was determined to be the failure of the main generator automatic voltage regulator power supply concurrent with the failure of the manual voltage regulator power supply. Post-event testing of voltage control system components revealed six failed SCR modules, in addition to the two failed power supplies (automatic and. |
| MainGenerator VoltageRegulator SCR(Thyristor Insert)ModuleBrownBoveri07060GR90075/4 GR0053P1EL-SCR6PlantDescription:
| | manual). Intermittent failures of the main generator automatic voltage regulator pulse amplifier and pulse driver circuit boards are the probable cause of the SCR module failures. |
| Manufacturer:
| | NRC FORM 366A *U.S.GPO:1986-0.824 538/455 (94)3) |
| Manufacturer IDNumber:EIISCode:NumberReplaced: | | |
| MainGenerator VoltageRegulator PulseGenerator CircuitBoardBrownBoveriGT032ALGV455007P13 EL-772PlantDescription:
| | NRC Form 3BBA U.S. NUCLEAR REGULATORY COMMISSION (94)3) |
| Manufacturer:
| | LICENSEE E T REPORT (LER) TEXT CONTINUAT J APPROVEO OMB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) PAGE (3) |
| Manufacturer IDNumber:EIISCode:NumberReplaced: | | SEQUENTIAL REVISION NUMBER NUMBER D.C. Cook Nuclear TEXT //F moro <<>>oo /4 Er)U)or/, I>>o Plant, Unit |
| MainGenerator VoltageRegulator PulseAmplifier CircuitBoardBrownBoveri07102A139046615RUT/RU10LGV455011P EL-AMP2PreviousSimilarEventsNoneNRCFORMSBBA(()43)*U.S.GPO.10854 524538/455
| | ////or>>/NRC %%drm 3654'4/ (17) 2 0 5 0 0 0 3 1 8 0 0 7 0 0 0 4 " 0 5 Anal sis of Event This Engineered Safety Features Actuation, which resulted in a reactor trip sequence, is reportable pursuant. to 10 CFR 50.73 (a) |
| ~eileenDIAMICNIGANEIECTDICCDAIEANYDonaldC.CookNuclearPlantP.O.8ox456,8ridgman, Michigan49166August13,1987UnitedStatesNuclearRegulatory Commission DocumentControlDeskWashington, D.C.20555Operating LicenseDPR-74DocketNo.50-316Document. | | (2) (iv) . |
| ControlManager:Inaccordance withthecriteriaestablished by10CFR50.73entitledLicenseeEventReortinSstem,thefollowing reportisbeingsubmitted:
| | The Operations Sequence Monitor functioned as designed. A time study of parameters monitored concluded that all automatic protection system responses; reactor trip, and resulting actuations, functioned properly as a result of the Engineered Safety Features actuation. |
| 87-007-00 Sincerely, W.G.Smith,Jr.PlantManager/afhAttachment cc:JohnE.DolanA.B.Davis,RegionIIIM.P.AlexichR.F.KroegerH.B.BruggerR.W.Jurgensen NRCResidentInspector R.C.CallenG.Charnoff, Esq.D.HahnINPOD.Wigginton, NRCPNSRCA.A.BlindDottieSherman,ANILibraryJ.G.Feinstein/B. | | Based on the above, it is concluded that the event did not constitute an unreviewed safety question as defined in 10 CFR 50.59 (a) (2), nor did it adversely impact health and safety. |
| P.LauzauFileW(PI'll}} | | Corrective Actions Immediate corrective action involved operations personnel implementing plant procedures to verify proper response of the automatic protection system and to assess plant conditions for initiating appropriate recovery actions. All failed/suspected components were replaced (automatic and manual voltage regulator power supplies, six SCR modules, two pulse generator circuit boards and two pulse amplifier circuit. boards). Investigation regarding industry experience with Brown boveri voltage control systems indicates that the failures which occurred are consistent with those experienced at other plants and is not. the result of operator error or lack of maintenance. Evaluation as to the necessity for the redesign of the voltage control system, via the Plant Modification program, continues however, no specific preventive measures have been identified to date. |
| | NRC FORM SSEA rrU.S.GPO;198B 0-B24 538/455 (94)3) |
| | |
| | NRC Form 3BSA U.S. NUCLEAR REGULATORY COMMISSION (843) |
| | LICENSEE T REPORT (LER) TEXT CONTINUA APPROVEO OM8 NO. 3150-0I04 EXPIRES: 8/31/88 FACILITYNAME ll) DOCKET NUMBER (2) LFR NUMBER (8) PAGE (3) |
| | YEAR SEOUENTIAL REVISION NUMBER '.RS NUMBER D.C. Cook Nuclear Plant, Unit 2 0 5 0 0 0 3 ] 8 7 0 0 7 0 0 0 5 OF 0 5 TExT ///'oro BPoco /B r/v/rorE Irro o//I/mr/ NRc %%drrrr 3()343/ l IT) |
| | Failed Com onent Identification Plant |
| | |
| | == Description:== |
| | Main Generator Voltage Regulator Power Supplies Manufacturer: Brown Boveri Manufacturer ID Number: VRCE1R39287116 AR103963R1 ND-'501B EIIS Code: EL-UJX Number replaced: 2 Plant |
| | |
| | == Description:== |
| | Main Generator Voltage Regulator SCR (Thyristor Insert) Module Manufacturer: Brown Boveri Manufacturer ID number: 07060 GR90075/4 GR0053P1 EIIS Code: EL-SCR Number Replaced: 6 Plant |
| | |
| | == Description:== |
| | Main Generator Voltage Regulator Pulse Generator Circuit Board Manufacturer: Brown Boveri Manufacturer ID Number: GT032A LGV455007P13 EIIS Code: EL-77 Number Replaced: 2 Plant |
| | |
| | == Description:== |
| | Main Generator Voltage Regulator Pulse Amplifier Circuit Board Manufacturer: Brown Boveri Manufacturer ID Number: 07102 A1390466 15 RUT/RU10 LGV455011P EIIS Code: EL-AMP Number Replaced: 2 Previous Similar Events None NRC FORM SBBA *U.S.GPO.10854 524 538/455 |
| | (()43) |
| | |
| | ~eileen DIA MICNIGAN EIECTDIC CDAIEANY Donald C. Cook Nuclear Plant P.O. 8ox 456, 8ridgman, Michigan 49166 August 13, 1987 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Operating License DPR-74 Docket No. 50-316 Document. Control Manager: |
| | In accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Re ortin S stem, the following report is being submitted: |
| | 87-007-00 Sincerely, W. G. Smith, Jr. |
| | Plant Manager |
| | /afh Attachment cc: John E. Dolan A. B. Davis, Region M. P. Alexich III R. F. Kroeger H. B. Brugger R. W. Jurgensen NRC Resident Inspector R. C. Callen G. Charnoff, Esq. |
| | D. Hahn INPO D. Wigginton, NRC PNSRC A. A. Blind Dottie Sherman, ANI Library J. G. Feinstein/B. P. Lauzau File W(PI |
| | 'll}} |
Similar Documents at Cook |
|---|
Category:LICENSEE EVENT REPORT (SEE ALSO AO
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
ML17335A5171999-02-11011 February 1999
LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted
ML17335A5141999-02-10010 February 1999
LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405
ML17335A5011999-02-0101 February 1999
LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr
ML17335A4951999-01-29029 January 1999
LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr
ML17335A4961999-01-27027 January 1999
LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr
ML17335A4921999-01-19019 January 1999
LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components
ML17335A4721999-01-0606 January 1999
LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing
ML17335A4691999-01-0606 January 1999
LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr
ML17335A4661999-01-0505 January 1999
LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing
ML17335A4631999-01-0404 January 1999
LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr
ML17335A4481998-12-30030 December 1998
LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr
ML17335A4581998-12-28028 December 1998
LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process
ML17335A4281998-12-22022 December 1998
LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event
ML17335A4111998-12-17017 December 1998
LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr
ML17335A4141998-12-16016 December 1998
LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122
ML17335A4181998-12-16016 December 1998
LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria
ML17335A3871998-12-11011 December 1998
LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing
ML17335A3821998-12-0808 December 1998
LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr
ML17335A3781998-12-0707 December 1998
LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr
ML17335A3771998-12-0303 December 1998
LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned
ML17335A3741998-12-0202 December 1998
LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3
1999-09-17
[Table view]
Category:RO)
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
ML17335A5171999-02-11011 February 1999
LER 99-002-00:on 990112,determined That RCS Pressurizer PORVs Had Not Been Tested,Per Ts.Caused by Inadequate Scheduling Controls Allowing Personnel Error.Surveillance Procedure Was Completed & Updated LER Will Be Submitted
ML17335A5141999-02-10010 February 1999
LER 99-001-00:on 990106,noted That GE Hfa Relays Installed in EDGs May Not Meet Seismic Qualification.Caused by Operating Experience Info Incorrectly Dispositioned in 1985. Updated LER Will Be Submitted by 990405
ML17335A5011999-02-0101 February 1999
LER 98-060-00:on 981231,identified That Rt Sys Response Time Testing Did Not Comply with TS Definition.Caused by Inadequate Procedures.Corrective Actions Will Be Developed & Update to LER Will Be Submitted by 990415.With 990201 Ltr
ML17335A4951999-01-29029 January 1999
LER 98-059-00:on 981230,interim LER -single Failure in Containment Spray Sys Could Result in Containment Spray Ph Outside Design Occurred.Investigation Into Condition Continuing.Update Will Be Submitted by 990514 Ltr
ML17335A4961999-01-27027 January 1999
LER 98-057-00:on 981228,discovered That AFW Valves Were Not Tested IAW Inservice Testing Program.Caused by Failure to Recognize Design Bases Features Re Afws by Personnel. Updated LER Will Be Submittted by 990415.With 990127 Ltr
ML17335A4921999-01-19019 January 1999
LER 98-052-01:on 981128,no Analysis for NSR Sc Manual Loader for Tdafwp Could Be Found in Original Design.Cause Due to All Failure Modes Not Considered When Compressed Air Sys Originally Designed.Performed Review of Components
ML17335A4721999-01-0606 January 1999
LER 98-055-00:on 981207,potential for Condition Outside of Design Bases for Rod Control Sys Was Noted.Caused by Calibration Error Coupled with Single Rod Failure.Condition Rept Investigation Is Ongoing
ML17335A4691999-01-0606 January 1999
LER 98-056-00:on 981211,hot Leg Nozzle Gaps Resulted in Plant Being in Unanalyzed Condition.Analyses Are Being Performed by W to Resolve Problem.Updated LER Will Be Submitted by 990211.With 990106 Ltr
ML17335A4661999-01-0505 January 1999
LER 98-049-00:on 981020,emergency Boron Injection Flow Path Was Inoperable.Caused by Original Design Deficiency. Engineering Evaluation of Event Is Continuing
ML17335A4631999-01-0404 January 1999
LER 98-054-00:on 981202,discovered That at Least One MSSV Had Not Been Reset as Required by Ts.Engineering Is Continuing Review of Extent of Condition for Event.Updated LER Will Be Submitted by 990129.With 990104 Ltr
ML17335A4481998-12-30030 December 1998
LER 98-053-00:on 981130,discovered Use of Inoperable Substitute Subcooling Margin Monitor.Caused by Condition Existing Since Installation of Plant Process Computer in 1992.Updated LER Will Be Submitted.With 981230 Ltr
ML17335A4581998-12-28028 December 1998
LER 98-052-00:on 981128,turbine Driven AFW Pump Speed Controller Failure Mode Occurred.Caused Because Not All Failure Modes Were Considered When Compressed Air Sys Was Originally Designed.Verified Current Design Change Process
ML17335A4281998-12-22022 December 1998
LER 98-051-00:on 981122,reactor Trip Signal from Manual Safety Injection Not Verified as Required by TS Surveillance,Was Discovered.Maintenance Currently Evaluating Significance & Cause of Event
ML17335A4111998-12-17017 December 1998
LER 98-047-00:on 981117,potential for Increase Leakage from Reactor Coolant Pump Seals Was Identified.Util Is Working with W to Resolve Issue.Current Expectations Are to Submit Update to LER by 990215.With 981217 Ltr
ML17335A4141998-12-16016 December 1998
LER 98-058-00:on 981216,postulated High Line Break Could Result in Condition Outside Design Bases for AF Occurred. Caused by Deficiencies Associated with Administration of HELB Program.Analysis of AF Will Be Completed by 990122
ML17335A4181998-12-16016 December 1998
LER 98-050-00:on 980814,ancillary Equipment Installed in Ice Condenser Was Not Designed to Withstand Design Basis Accident/Earthquake Loads.Caused by Lack of Established Design Criteria.Developed Design Criteria
ML17335A3871998-12-11011 December 1998
LER 98-031-01:on 980610,potential Common Mode Failure of RHR Pumps Were Noted.Caused by Inaccurate Values.Accurate Miniflow Numbers Have Been Determined by Flow Testing
ML17335A3821998-12-0808 December 1998
LER 98-039-01 Re EOP Step Conflicts with Small Break LOCA Analysis.Ler 98-039-00 Has Been Canceled.With 981208 Ltr
ML17335A3781998-12-0707 December 1998
LER 98-007-00:on 981106,high Energy Line Break Effects in Auxiliary FW Sys Was Noted.Cause of Event Is Under Investigation & Will Be Completed by 990220.Updated LER Will Be Submitted by 990310.With 981207 Ltr
ML17335A3771998-12-0303 December 1998
LER 98-046-00:on 981103,determined That Afs Was Unable to Meet Design Flow Requirements During Special Test.Caused by Failure to Consider All Aspects of Sys Operation in Design of Suction Basket Strainers.Sys Will Be Redesigned
ML17335A3741998-12-0202 December 1998
LER 97-011-02:on 970822,operation Was Noted Outside Design Bases for ECCS & CSP for Switchover to Recirculation Sump Suction.Caused by Ineffective Change Mgt.Revised Procedure for Switchover 01(02) Ohp 4023.ES-1.3
1999-09-17
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17335A5631999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 1.With 991012 Ltr
ML17335A5621999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 2.With 991012 Ltr
ML17335A5481999-09-30030 September 1999
Non-proprietary DC Cook Nuclear Plant Units 1 & 2 Mods to Containment Sys W SE (Secl 99-076,Rev 3).
ML17335A5451999-09-28028 September 1999
Rev 1 to Containment Sump Level Design Condition & Failure Effects Analysis for Potential Draindown Scenarios.
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1481999-09-17017 September 1999
Independent Review of Control Rod Insertion Following Cold Leg Lbloca,Dc Cook,Units 1 & 2.
ML17326A1211999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 2.With 990915 Ltr
ML17326A1201999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 1.With 990915 Ltr
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17335A5461999-08-0202 August 1999
Rev 0 to Evaluation of Cook Recirculation Sump Level for Reduced Pump Flow Rates.
ML17326A0871999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Unit 1.With 990812 Ltr
ML17326A0861999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Units 2.With 990812 Ltr
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0511999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 2.With 990709 Ltr
ML17326A0501999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 1.With 990709 Ltr
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17326A0061999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Dcp.With 990609 Ltr
ML17326A0071999-05-31031 May 1999
Monthly Operating Rept for May 1999 for DC Cook Nuclear Plant,Unit 2.With 990609 Ltr
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17335A5301999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 1.With 990508 Ltr
ML17335A5291999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 2.With 990508 Ltr
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5491999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant Unit 2.With 990408 Ltr
ML17325B5441999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant,Unit 1.With 990408 Ltr
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B5671999-03-0202 March 1999
Summary of Unit 1 Steam Generator Layup Chemistry from 980101 to 990218.
ML17325B4631999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Power Station,Unit 2.With 990308 Ltr
ML17325B4621999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Plant,Unit 1.With 990308 Ltr
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
1999-09-30
[Table view] |
Text
REGULA Y INFORNATION DISTRIBUTIO SYSTEM (RIDS)
ACCESSION NBR: 8708180184 DOC. DATE: 87/08/13 NOTARIZED: NO DOCKET FAG IL: 50-316 Donald C. Cook Nuc lear Poeer AFFILIATION P lant'nit 2i Indiana & 05000316 AUTH. MANE AUTHOR SAMPSONe J. R. Indiana & Michigan Electric Co.
SMITHS'. G. Indiana & Michigan Electric Co.
REC IP. NAME REC IP IENT AFFILIATION
SUBJECT:
LER 87-007-00: on 870714'SF reactor trip occurred due to undervoltage of reactor coolant busses. Caused bg failure of main generator voltage control sos. Failed & suspected components of sos replaced. W/870813 ltr.
DISTRIBUTION CODE: IE22D COPIES RECEIVED: LTR ENCL SIZE:
TITLE: 50. 73 Licensee Event Report (LER)> Incident Rpti etc.
NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NANE LTTR ENCL ID CODE/NANE LTTR ENCL PD3-3 LA 1 1 PD3-3 PD 1 1 WIGGINGTON D 1 1 INTERNAL: ACRS NICHELSON 1 ACRS NOELLER 2 2 AEOD/DOA 1 AEOD/DSP/NAS 1 1 AEOD/DSP/PQAB 2 2 AEOD/DSP/TPAB 1 DEDRO 1 1 NRR/DEST/ADE 1 0 NRR/DEST/ADS 1 0 NRR/DEST/CEB 1 1 NRR/DEST/ELB 1 NRR/DEST/ ICSB 1 NRR/DEST/MEB 1 1 NRR/DEST/NTB '1 1 NRR/DEST/PSB 1 1 NRR/DEBT/RSB 1 NRR/DEST/SGB 1 1 NRR/DLPG/HFB 1-NRR/DLPG/GAB 1 1 NRR/DOEA/EAB 1 NRR/DREP/RAB 1 1 RlhhDRE PB 2 2 NRR/PMAS/ILRB 1 G FILE 02 1 RES DEPY GI 1 1 R FORDS J 1 1 RES/DE/EIB 1 1 RGN3 FILE 01 1 1 EXTERNAL: EG&G GROf<i N 5 5 H ST LOBBY WARD 1 LPDR 1 NRC PDR 1 NSIC HARRIS' 1 1 NSIC MAYST' 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 44 ENCL 42
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NRC FOrm 3dd U.S. NUCLEAR REOULATORY COMMISSION (943)
APPROVED OMB NO. 31504104 EXPIRES: SI31ldd LICENSEE EVENT REPORT (LER)
FACILITY NAME PI DOCKET NUMBER (2) PA E 3l D. C. Cook Nuclear Plant, Unit 2 0 5 0 0 0 3 1 61 oF ESF Actuation (Reactor Trip) Due to Undervoltage of the Reactor Coolant Pum Busses as a Result of C EVENT DATE (5) LER NUMBER (5) REPORT DATE (7) OTHER FACILITIES INVOLVED (5)
MONTH DAY '9EAR YEAR SEOUENTrAI f<<r<<<<REVISION MONTH OAY YEAR FACILITYNAMES DOCKET NUMBER(El NUMBEtl .>>Td NUMBER 0 5 0 0 0 0 7 1 4 8 7 8 7 0 0 7 0 0 1 3 8 7 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE AEOUIREMENTS OF 10 CFR II: IChecfr one or more of the follory'np) (11 OPERATINO MODE (9) 20.402(B) 20.405(c) 60.73(s) (2) (iv) 73.71(B)
POWER 20.406(s) (1)(i) 50M(c)(1) 50.73(s)(2 I (v) 73.71(c)
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LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER AREA CODE J. R. Sampson Safet 6 Assessment Su erinte COMPLETE ONE LINE FOR EACH COMPONENT FAILURE OESCRIBEO IN THIS REPORT (13I 616465-590 x>gjc P,
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SUPPLEMENTAL REPORT EXPFCTED (14) 3 4 5 5 N MONTH DAY YEAR EXPECTED SUB M I SS ION DATE (15)
YES Ilfyer, complere EXPECTED S(fphtlSSION OATEI X No ABSTRACT ILlmit to te00 rpecer, I e., epproxlmerely llfteen tlnple specs typevrritren finer) (15)
On July 14, 1987, at 0707 hours0.00818 days <br />0.196 hours <br />0.00117 weeks <br />2.690135e-4 months <br />, an Engineered Safety Features Actuation (Reactor Trip) occurred due to undervoltage of the reactor coolant pump busses. The undervoltage condition was the result of the failure of the main generator voltage control system.
The Unit was stabilized in Mode 3 (Hot Standby) at approximately 0755 hours0.00874 days <br />0.21 hours <br />0.00125 weeks <br />2.872775e-4 months <br />, July 14, 1987. No abnormal reactor trip sequence responses were noted. The NRC was notified of the event via the ENS at 0815 hours0.00943 days <br />0.226 hours <br />0.00135 weeks <br />3.101075e-4 months <br />, July 14, 1987.
Post-event testing of the voltage control system components revealed two failed power supplies (automatic and manual), and six failed SCR Modules. All failed/suspected components were replaced. Investigation regarding industry experience with Brown Boveri voltage control systems indicates that the failures which occurred are consistent with those experienced at other plants and is not a result of operator error or lack of maintenance.
Evaluation as to the necessity for the redesign of the voltage control system, via the Plant Modification progra'm, continues however, no specific preventive measures have been identified to date.
8708180184 870813 PDR ADOCI(r 0500031(EI 8 PDR NRC Form 3dd (9d3)
NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION
(()4)3)
LICENSEE E T REPORT (LER) TEXT CONTINUA APPROVED OMB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR 8'jets'EGUENTIAL REVISION 4 . NUMSE4 NOME E 4 D.C. Cook Nuclear Plant, Unit 2 p p p p p 3 1 8 7 0 0 7 0 0 0 2oF0 TEXT /F IINvo EPEco ir roqoiaE ooo aANPenel HRC Form 355AB/ ((1)
Conditions Prior to Occurrence Unit 2 in Mode 1 (power operations) at 80 percent Reactor Thermal Power.
Descri tion of Event On July 14, 1987, at 0707 hours0.00818 days <br />0.196 hours <br />0.00117 weeks <br />2.690135e-4 months <br />, an Engineered Safety Features Actuation (Reactor Trip Sequence) occurred due to undervoltage of the reactor coolant pump busses (EIIS/EA). The undervoltage condition was the result of the failure of the main generator voltage. control system (EIIS/EL) .
At approximately 0637 hours0.00737 days <br />0.177 hours <br />0.00105 weeks <br />2.423785e-4 months <br />, July 14, 1987, the incoming Unit, Supervisor (Utility-Licensed Operator) noticed that the output from the main generator automatic voltage regulator (EIIS/EL-RG) was not nulled with the manual voltage regulator (EIIS/EL-RG).
The automatic regulator was currently in service and had been operating at the lowest value of the voltage band throughout the previous shift. When the Unit Supervisor adjusted the manual regulator to match the automatic setpoint, the main generator voltage and exciter current indication dipped momentarily and returned to normal [the adjustment. should have had no effect on the main generator (EIIS/EL-GEN)]. He then requested that a Reactor Operator (RO) (Utility-Licensed Operator) be posted at.
the regulator to monitor the indication. The RO reviewed the generator panel status with the incoming Unit, Supervisor and noticed that. the generator output voltage was slightly below the minimum of the operating band. The RO increased the automatic regulator to restore the generator output voltage to the bottom of the operating band. For the remaining period prior to the trip sequence, the volt-amps reactive, generator field temperature, and to a lesser degree the generator output voltage indication showed unstable behavior. No alarms occurred until seconds before the trip sequence.
Following the trip sequence [opening of the reactor trip breakers (EIIS/JE-BKR), insertion of the reactor control rods (EIIS/AA-ROD),
feedwater isolation (EIIS/JB), automatic starting of the motor-driven and turbine-driven auxiliary feedwater pumps (EIIS/BA-P)]
operations personnel immediately implemented the special Emergency Operating Procedure, 1 OHP -4023.E-O, to verify proper response of the automatic protection system (EIIS/JC) and to assess plant conditions for initiating appropriate recovery actions. There was no automatic or manual actuation of the intermediate head safety injection system (EIIS/BQ) .
NAC FOAM 366A o U.S.GPO:1988 0.824 538/455 (84)3)
NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION (94)31 LICENSEE E T REPORT (LER) TEXT CONTINUA APPROVED OMB NO. 3150&104 EXPIRES: 8/31/88 FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)
YEAR W SEOUENTIAL NUMEER NUMBER D.C. Cook Nuclear Plant, Unit 2 0 s 0 0 0 3 1 6 8 7 0 0 7 0 0 3oF 0 5 TEXT /Smote EPooo )r nrqrr)or/, rroo/Ar/orro/HRC Forrrr 35549/ (IT)
The voltage transient was also detected by the T21D safeguards bus (train "A") undervoltage relays (EIIS/EA-27) which started the CD emergency diesel generator (EIIS/EK-DG) and initiated load shedding of the "A" train safeguards busses. The RB" train busses were approximately 70 to 100 volts higher than the "A" train busses immediately prior to the trip consequently the load shedding actuation logic for the "BR train was not satisfied, and the AB emergency diesel generator did not, automatically start. The East centrifugal charging pump (EIIS/CB-P) had been operating and was not restored during blackout load sequencing (as per design).
Operators manually started the West centrifugal charging pump to restore reactor coolant pump seal injection (EIIS/CB) and opened IM0-911, charging pump suction from the refueling water storage tank (EIIS/CB-TK), to restore pressurizer level and to clear the letdown isolation (which occurred, as designed, due to the loss of the East centrifugal charging pump). The load shed also de-energized the lighting transformers (EIIS/FF-XFMR), isolating plant lighting and outlets (EIIS/FF-OUT). Other lighting transformer loads lost included; the analog rod position indication (EIIS/AA-XF) and the rod bottom lights (EIIS/AA-IL),
chart drives for the narrow range and wide range reactor coolant temperature recorder (EIIS/JC-TR), main generator megawatt recorder chart drive (EIIS/EL-XR), normal control room lighting (EIIS/FF)E and the AB emergency diesel generator recorder chart drive (EIIS/EK-XR). The load shedding sequence experienced was in accordance with undervoltage protection system (EIIS/EA)design.
Power was manuallly restored to the lighting transformers within 30 minutes. The AB emergency diesel generator was manually started 45 seconds after the trip, but was never loaded. Both emergency diesels were secured after offsite power was established to the "A" train safeguards busses. The Unit was stabilized in Mode 3 (hot standby) at approximately 0755 hours0.00874 days <br />0.21 hours <br />0.00125 weeks <br />2.872775e-4 months <br />, July 14, 1987.
The NRC was notified of the event via the ENS at 0815 hours0.00943 days <br />0.226 hours <br />0.00135 weeks <br />3.101075e-4 months <br />, July 14, 1987.
With the exception of the failure of the main generator voltage control system, there were no inoperative structures, components, or systems that contributed to this event.
Cause of Event The cause of the event was determined to be the failure of the main generator automatic voltage regulator power supply concurrent with the failure of the manual voltage regulator power supply. Post-event testing of voltage control system components revealed six failed SCR modules, in addition to the two failed power supplies (automatic and.
manual). Intermittent failures of the main generator automatic voltage regulator pulse amplifier and pulse driver circuit boards are the probable cause of the SCR module failures.
NRC FORM 366A *U.S.GPO:1986-0.824 538/455 (94)3)
NRC Form 3BBA U.S. NUCLEAR REGULATORY COMMISSION (94)3)
LICENSEE E T REPORT (LER) TEXT CONTINUAT J APPROVEO OMB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (8) PAGE (3)
SEQUENTIAL REVISION NUMBER NUMBER D.C. Cook Nuclear TEXT //F moro <<>>oo /4 Er)U)or/, I>>o Plant, Unit
////or>>/NRC %%drm 3654'4/ (17) 2 0 5 0 0 0 3 1 8 0 0 7 0 0 0 4 " 0 5 Anal sis of Event This Engineered Safety Features Actuation, which resulted in a reactor trip sequence, is reportable pursuant. to 10 CFR 50.73 (a)
(2) (iv) .
The Operations Sequence Monitor functioned as designed. A time study of parameters monitored concluded that all automatic protection system responses; reactor trip, and resulting actuations, functioned properly as a result of the Engineered Safety Features actuation.
Based on the above, it is concluded that the event did not constitute an unreviewed safety question as defined in 10 CFR 50.59 (a) (2), nor did it adversely impact health and safety.
Corrective Actions Immediate corrective action involved operations personnel implementing plant procedures to verify proper response of the automatic protection system and to assess plant conditions for initiating appropriate recovery actions. All failed/suspected components were replaced (automatic and manual voltage regulator power supplies, six SCR modules, two pulse generator circuit boards and two pulse amplifier circuit. boards). Investigation regarding industry experience with Brown boveri voltage control systems indicates that the failures which occurred are consistent with those experienced at other plants and is not. the result of operator error or lack of maintenance. Evaluation as to the necessity for the redesign of the voltage control system, via the Plant Modification program, continues however, no specific preventive measures have been identified to date.
NRC FORM SSEA rrU.S.GPO;198B 0-B24 538/455 (94)3)
NRC Form 3BSA U.S. NUCLEAR REGULATORY COMMISSION (843)
LICENSEE T REPORT (LER) TEXT CONTINUA APPROVEO OM8 NO. 3150-0I04 EXPIRES: 8/31/88 FACILITYNAME ll) DOCKET NUMBER (2) LFR NUMBER (8) PAGE (3)
YEAR SEOUENTIAL REVISION NUMBER '.RS NUMBER D.C. Cook Nuclear Plant, Unit 2 0 5 0 0 0 3 ] 8 7 0 0 7 0 0 0 5 OF 0 5 TExT ///'oro BPoco /B r/v/rorE Irro o//I/mr/ NRc %%drrrr 3()343/ l IT)
Failed Com onent Identification Plant
Description:
Main Generator Voltage Regulator Power Supplies Manufacturer: Brown Boveri Manufacturer ID Number: VRCE1R39287116 AR103963R1 ND-'501B EIIS Code: EL-UJX Number replaced: 2 Plant
Description:
Main Generator Voltage Regulator SCR (Thyristor Insert) Module Manufacturer: Brown Boveri Manufacturer ID number: 07060 GR90075/4 GR0053P1 EIIS Code: EL-SCR Number Replaced: 6 Plant
Description:
Main Generator Voltage Regulator Pulse Generator Circuit Board Manufacturer: Brown Boveri Manufacturer ID Number: GT032A LGV455007P13 EIIS Code: EL-77 Number Replaced: 2 Plant
Description:
Main Generator Voltage Regulator Pulse Amplifier Circuit Board Manufacturer: Brown Boveri Manufacturer ID Number: 07102 A1390466 15 RUT/RU10 LGV455011P EIIS Code: EL-AMP Number Replaced: 2 Previous Similar Events None NRC FORM SBBA *U.S.GPO.10854 524 538/455
(()43)
~eileen DIA MICNIGAN EIECTDIC CDAIEANY Donald C. Cook Nuclear Plant P.O. 8ox 456, 8ridgman, Michigan 49166 August 13, 1987 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Operating License DPR-74 Docket No. 50-316 Document. Control Manager:
In accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Re ortin S stem, the following report is being submitted:
87-007-00 Sincerely, W. G. Smith, Jr.
Plant Manager
/afh Attachment cc: John E. Dolan A. B. Davis, Region M. P. Alexich III R. F. Kroeger H. B. Brugger R. W. Jurgensen NRC Resident Inspector R. C. Callen G. Charnoff, Esq.
D. Hahn INPO D. Wigginton, NRC PNSRC A. A. Blind Dottie Sherman, ANI Library J. G. Feinstein/B. P. Lauzau File W(PI
'll
