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| | issue date = 03/19/1990 | | | issue date = 03/19/1990 |
| | title = Forwards Response to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implications of Control Sys in LWR Nuclear Power Plants,' Per 10CFR50.54(f) | | | title = Forwards Response to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implications of Control Sys in LWR Nuclear Power Plants,' Per 10CFR50.54(f) |
| | author name = GOLDBERG J H | | | author name = Goldberg J |
| | author affiliation = FLORIDA POWER & LIGHT CO. | | | author affiliation = FLORIDA POWER & LIGHT CO. |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:ACCELERATEDDIRIBUTIONDEMON~TIONSYSIEMjy'eREGULATORYINFORMATIONDISTRIBUTIONSYSTEM(RIDS)ACCESSIONNBR:9003260394DOC.DATE:90/03/19NOTARIZED:YESFACIL:50-335St.LuciePlant,Unit1,FloridaPower6LightCo.50-389St.LuciePlant,Unit2,FloridaPower6LightCo.AUTH.NAMEAUTHORAFFILIATIONGOLDBERG,J.H.FloridaPower6LightCo.RECIP.NAMERECIPIENTAFFILIATIONDocumentControlBranch(DocumentControlDesk) | | {{#Wiki_filter:ACCELERATED DIRIBUTION DEMON~TION SYSIEM j 'e y |
| | REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) |
| | ACCESSION NBR:9003260394 DOC.DATE: 90/03/19 NOTARIZED: YES DOCKET FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power 6 Light Co. 05000335 50-389 St. Lucie Plant, Unit 2, Florida Power 6 Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION GOLDBERG,J.H. Florida Power 6 Light Co. |
| | RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk) |
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| ==SUBJECT:== | | ==SUBJECT:== |
| RespondstoGenericLtr89-19,"RequestforActionreResolutionofUSIA;47."k'ISTRIBUTIONCODE:AOOIDCOPIESRECEIVED:LTRQENCL(SIZE:TITLE:ORSubmittal:GeneralDistributionNOTES:DOCKET0500033505000389RECIPIENTIDCODE/NAMEPD2-2LANORRIS,JINTERNAL:ACRSNRR/DOEA/OTSB11NRR/DST/SELB8DNRR/DST/SRXB8EOCRFILEEXTERNAL:LPDRNSICCOPIESLTTRENCL11556611111110111111RECIPIENTIDCODE/NAMEPD2-2PDNRR/DET/ECMB9HNRR/DST8E2NRR/DST/SICB7ENUDOCS-ABSTRACTOGC/HDS2RES/DSIR/EIBNRCPDRCOPIESLTTRENCL11111111111011NOTETOALL"RIDS"RECIPIENTS:PLEASEHELP,USTOREDUCEWASTEICONTACTTHE.DOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATEYOURNAMEFROMDISTRIBUTIONLISISFORDOCUMENTSYOUDON'TNEEDtTOTALNUMBEROFCOPIESREQUIRED:LTTR27ENCL25 0
| | Responds to Generic Ltr 89-19, "Request for Action re Resolution of USI A;47." |
| P.O.Box14000,JunoBeach,FL33408-0420MARCH49~99OL-90-10610CFR50.54(f)U.S.NuclearRegulatoryCommissionAttn:DocumentControlDeskWashington,D.C.20555Gentlemen:Re:St.LucieUnits1and2DocketNos.50-335and50-389GenericLetter89-19Request.forActionRelatedtoResolutionofUnresolvedSafetyIssueA-47"SafetyImplicationsofControlSystemsinLWRNuclearPowerPlants"Pursuantto10CFR50.54fAsaresultofthetechnicalresolutionofUSIA-47,"SafetyImplicationsofControlSystemsinLWRNuclearPowerPlants,"the'NRCconcludedthatprotectionshouldbeprovidedforcertaincontrolsystemfailures,and,forcertainplants,thatselectedemergencyproceduresshouldbemodifiedtoassurethatplanttransientsresultingfromcontrolsystemfailuresdonotcompromisepublicsafety.GenericLetter89-19,issuedSeptember20,1989,providedrecommendationsforcontrolsystemdesignandproceduralmodificationsforresolutionofUSIA-47.Specifically,GenericLetter89-19recommendedthatallCombustionEngineeringplantdesignsprovideautomaticsteamgeneratoroverfillprotectiontomitigatemainfeedwater(MFW)overfeedevents,andthatplantproceduresandtechnicalspecificationsincludeprovisionstoperiodicallyverifytheoperabilityoftheMFWoverfillprotectionandensurethattheautomaticoverfillprotectionisoperableduringreactorpoweroperation.Additionally,itwasrecommendedthatallutilitiesthathaveplantsdesignedwithhighpressureinjectionpumpdischargepressureslessthanorequalto1275psireassesstheiremergencyproceduresandoperatortrainingprogramsandmodifythem,asneeded,toensurethattheoperatorscanhandlethefullspectrumofpossiblesmall-breaklossofcoolantaccidentscenarios.FloridaPower6-LightCompany'sresponsetotherecommendationsinGenericLetter89-19forSt.LucieUnits1and2isattached.90032603949003i9PDRADOCK05000335PPD.C,oIanFPLGroupcompany 0t~t~\i U.S.NuclearRegulatoryCommissionL-90-106PagetwoShouldtherebeanyquestionsregardingtheattachedinformation,pleasecontactus.Vrytrulyyours,J.H.GoldbergExecutiveVicePresidentJHG/MSD/gpAttachmentcc:StewartD.Ebneter,RegionalAdministrator,RegionII,USNRCSeniorResidentInspector,USNRC,St.LuciePlant STATEOFFLORIDA))ss.COUNTYOFPALMBEACH)J.H.Goldberbeingfirstdulysworn,deposesandsays:ThatheisExecutiveVicePresident,ofFloridaPower&LightCompany,theLicenseeherein;Thathehasexecutedtheforegoingdocument;thatthestatementsmadeinthisdocumentaretrueandcorrecttothebestofhisknowledge,informationandbelief,andthatheisauthorizedtoexecutethedocumentonbehalfofsaidLicensee.J.H.GoldbeSubscribedandsworntobeforemethis~7dayof'NOTARYPUBLC,in<,.PalmPeach,StateandforthofFloridaountyofNot%,pobrir,StateofHorMaNyt:omm~sonExpreslone11993SoadadrheoTroyFainIasaraacaInc..My,Commissionexpires | | CODE: AOOID k'ISTRIBUTION COPIES RECEIVED:LTR Q ENCL ( SIZE: |
| /<</~~l/j4t~l ATTACHMENTResponsetoGenericLetter89-19,UnresolvedSafetyIssueA-47'~SafetImlementationofControlSstemsinLWRPowerPlants~~BackroundThepurposeofGenericLetter89-19istoensurethatallpowerplantshavereactorvesselorsteamgeneratoroverfillprotection.Additionally,itwasrecommendedthatallutilitiesthathaveplantsdesignedwithhighpressureinjectionpumpdischargepressureslessthanorequalto1275psireassesstheiremergencyproceduresandoperatortrainingprogramsandmodifythem,asneeded,toensurethattheoperatorscanhandlethefullspectrumofpossiblesmall-breaklossofcoolantaccidentscenarios.Enclosure2toGenericLetter89-19discussestherecommendationsforCombustionEngineering(CE)NuclearSteamSupplySystempressurizedwaterreactorslikeSt.Lucie.Therecommendationsare:(1)AllCEplantsprovideautomaticsteamgeneratoroverfillprotection.(2)AllCEplantsprovideplantproceduresandtechnicalspecificationsforperiodicsurveillanceoftheoverfillprotection.(3)CEplantswithhighpressureinjectionpumpdischargepressureslessthan1275psireassesstheiremergencyproceduresandoperatortrainingtoensuresafeshutdownduringanypostulatedsmallbreaklossofcoolantaccident.InEnclosure2theNRCconcludes,"CE-designedplantsdonotprovideautomaticsteamgeneratoroverfillprotectionthatterminatesMFWflow."FPLisofthepositionthat,withrespecttoSt.LucieUnits1and2,thisisincorrectandbothunitshavesteamgeneratoroverfillprotectionsystemstosatisfactorilyresolveUSIA-47.TherecommendationsoutlinedinSection4ofEnclosure2arediscussedbelowforSt.LucieUnits1and2.Recommendation4aItisrecommendedthatallCombustionEngineeringplantsprovideautomatic,steamgeneratoroverfillprotectiontomitigatemainfeedwater(MFW)overfeedevents.Thedesignfortheoverfill-protectionsystemshouldbesufficientlyseparatefromtheMFWcontrolsystemtoensurethattheMFWpumpwilltriponasteamgeneratorhigh-water-levelsignalwhenrequired,evenifalossofpower,alossofventilation,orafireinthecontrolportionoftheMFWcontrolsystemshouldoccur.Commonfailuremodesthatcoulddisableoverfillprotectionandthefeedwatercontrolsystem,but,wouldstillresultinafeedwaterpumptrip,areconsideredacceptablefailuremodes. | | TITLE: OR Submittal: General Distribution NOTES: |
| FPLResonse:SteamGeneratorOverfillSstemDescritionSt.LucieUnits1and2usea2-of-4coincidencelogicforsteamgeneratoroverfillprotectionfromfourReactorProtectiveSystem(RPS)levelloopspersteamgenerator.Theloopsprovidehighandhigh-,highlevelisolationsignals.Thesesignalsresultinfeedwaterregulatingvalveclosu'reonhighsteamgeneratorlevelandfeedwaterpumptripandmainturbinetriponhigh-highsteamgeneratorlevel.Theinitiatingsignalsareusedinnon-safetyrelatedoverfillprotectioncircuits.Thelevelloopswhichinitiatesteamgeneratorisolation(feedwaterregulatingvalveclosure,feedwaterpumptripandmainturbinetrip)aredifferentfromthelevelchannelsusedfornormalfeedwaterregulatingvalvecontrol.ThehighlevelisolationsignalfromtheRPSlevelindicatingcontrollersprovidesafeedwatercontroloverridesignal,whichresultsinclosureofthefeedwaterregulatingvalvefortherespectivetrain.AlthoughtheinitiatingsignalisgeneratedfromtheRPSlevelloops,themainfeedwaterregulatingvalveisolationfunctionisintegraltothefeedwatercontrolsystem.Thehigh-highlevelisolationfunctionisaccomplishedindependentlyofthefeedwatersystemviaaseparatecircuit.Uponreceiptofahigh-highlevelisolationsignal,witha2-of-4coincidencefromeithersteamgenerator,atriprelayisenergized.Thisrelaytripsbothfeedwaterpumpsandenergizesasecondtriprelayandan.autostopsolenoid.Thesecondtriprelayprovidesannunciation.Theautostopsolenoidprovidesforamainturbinetripbyclosingtheturbinestop/controlvalves.Theinitiatinglogicforsteamgeneratoroverfillprotectionisfromeitheroftwosetsoffoursafety-grade,independentRPStransmitterloops.ThesignalsinitiatefromlevelindicatorswhicharemountedontheReactorTurbineGeneratorBoard.TheSt.LucieUnit2signalspassthroughisolationcabinets;theUnit1signalsdonot.Thesignalsthenpassthroughcommon(separatecircuits)"normal/bypass"controlswitches.Thehighlevelsignalthentiesintothefeedwaterregulatingsystemmodules.Thehigh-highlevelsignaltiesintothetwotriprelaysandturbinestopsolenoid.Reviewofthepowersuppliesforthefeedwatercontrolsystemandthesteamgeneratoroverfillprotectioncircuitshowstheyarefromdifferentsources.Thenormalpowersuppliesforasinglefeedwatercontroltrainandtheoverfillprotectionsystemareultimatelyfedfromacommon480voltswitchgear(Unit1)or4160voltbus(Unit2);however,thefeedwatercontrolsystemissupplied120voltACpowerandthesteamgeneratoroverfillprotectionsystemisprovided125voltDC.Thesteamgeneratoroverfillprotectioncircuitisprovidedback-upemergencypowerfromthe2B(1A)stationbatteryintheeventnormalpowerislost.Eachtrainofthefeedwatercontrolsystemisalsoprovidedwithanautomatictransfertoaback-uppowersourcefromthe120voltvitalACcabinets,toensurefurtherreliability.Thissourceis V~tultimatelypoweredfromthe2AB(lAB)swingbus,whichcanbemanuallyalignedtoeithertrain"A"or"B"480voltACswitchgear.Emergencyback-uppowerisalsoprovidedviathe2D(1D)stationbatteryandassociatedinverter.CommonFailureModesDiscussionVariouspotentialcommonmodefailuresbetweenthefeedwatercontrolsystemandthesteamgeneratoroverfillprotectioncircuitswereevaluated.Thelimitingcommonmodefailuresevaluatedwereforalossofpower,acommonfire,andfailureofthesharedoverride/manualcontrolswitch.PowerLossThenormalpowersuppliesforthefeedwatercontrolandsteamgeneratoroverfillprotectioncircuitsarenottotallyindependent(asdiscussedabove);however,theback-upsuppliesaresufficientlyindependenttoensureoperabilityofoneorbothsystemsintheeventacommonnormalbusislost.Thesteamgeneratoroverfillprotectioncircuitryisprovidedback-up125voltDCpowerfromsafety-relatedstationbattery2B(1A),suchthatthefeedwaterpumptripshouldstilloccurifahigh-highsteamgeneratorlevelconditionoccurscoincidentwithalossofnormalpower.Similarly,thefeedwatercontrolsystemisprovidedwithanautomatictransfertothe120voltACvitalbus(fedoffthe"AB"480voltMotorControlCenter)intheeventnormalpowerislost.Thisfeedwatercontrolpowersourceisfurtherbacked-upbythe2D(1D)battery.Furthermore,thedesignofthefeedwatercontrolsystemrequiresthefeedwaterregulatingvalvetofailclosedonalossofpower,suchthatevenintheremoteevent(multiplefailure)thecommonbusandtheback-uppowersuppliesarelost,feedwaterflowwillstillbeisolatedfortheaffectedtrain.FireIntheeventofafireinthecontrolroom,thefeedwatercontrolandoverfillprotectionsystemsarenotsufficientlyseparatetoensuretheoperabilityofatleastonesystem.Thetwosystemshavecircuitry,whichsharecommon"Norm/bypass"handswitchesinthecontrolroom,commoncabletrayroutingandarelocatedwithinadjacentcontrolroompanels.Givenafireofalargemagnitude,whichrendersthecontrolroomuninhabitable,amanualreactor/turbinetripwouldbeinitiatedpriortoleavingthecontrolroomandsafeshutdownwouldbeaccomplishedfromtheremoteshutdownpanel.Mainfeedwaterflowwouldbeisolatedperoperatingproceduresandsteamgeneratorleveliscontrolledviaauxiliaryfeedwater.SharedControlSwitchThefinalcommonmodefailurepostulatedwasthefailureofacommon"normal/bypass"handswitchduetoinadvertentmispositioning,contactfailure"open,"orshorttoground.
| | RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 NORRIS,J 5 5 INTERNAL: ACRS 6 6 NRR/DET/ECMB 9H 1 1 NRR/DOEA/OTSB11 1 1 NRR/DST 8E2 1 1 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 OC 1 0 OGC/HDS2 1 0 R FILE 1 1 RES/DSIR/EIB 1 1 EXTERNAL: LPDR 1 1 NRC PDR NSIC 1 1 NOTE TO ALL "RIDS" RECIPIENTS: |
| Inadvertentmispositioningofthehandswitchesisnotcredible,asannunciationisprovidedtowarnifanyofthefourswitchesismispositioned.Furthermore,ifasinglehandswitchismispositionedoritscontactsfailopen,thehighandhigh-highoverfillprotectionarestilloperable.A2-of-3coincidencewillnowberequiredtoinitiateahigh/high-highsteamgeneratorisolationsignal.Intheeventasingleswitchshortstoground,boththefeedwatercontrolandsteamgeneratoroverfillprotectioncircuitslosepowerandthefeedwaterregulatingvalvesfailclose.Basedontheabove,FPLconcludesthatthesteamgeneratoroverfillprotectionsystemcurrentlyinstalledinSt.LucieUnits1and2providesadequateassurancethatasteamgeneratoroverfilleventwillbeprevented.Recommendation4bItisrecommendedthatplantproceduresandtechnicalspecificationsforallCombustionEngineeringplantsincludeprovisionstoverifyperiodicallytheoperabilityofoverfillprotectionandensurethatautomaticmainfeedwateroverfillprotectionisoperableduringreactorpoweroperation.Theinstrumentationshouldbedemonstratedtobeoperablebytheperformanceofachannelcheck,channelfunctionaltesting,andchannelcalibration,includingsetpointverificationandbyidentifyingtheLCOs.Thesetechnicalspecificationsshouldbecommensuratewithexistingplanttechnicalspecificationrequirementsforchannelsthatinitiateprotectionactions.FPLResonse:St.LuciePlantproceduresincludeprovisionstoperiodicallyverify.steamgeneratoroverfillprotectionandensurethatautomaticmainfeedwateroverfillprotectionisoperableduringreactorpoweroperation.Testingofthesteamgeneratorhighandhigh-highlevelcircuitryiscompletedonan18monthrefuelingbasisbyfunctionallytestingthecircuitryusingthe2-of-4logictoclosethefeedwaterregulatingvalve,tripthefeedwaterpumpandtriptheturbine.TheNRCstaffiscurrentlyreviewingtheNuclearSteamSupplySystem(NSSS)vendorspecificrevisedstandardtechnicalspecificationsdevelopedusingthecriteriaoftheNRCinterimpolicystatementontechnicalspecificationimprovements(FederalRegisterp.3788/Vol.52No.25/Friday,February6,1987).RecommendationsforspecificLimitingConditionsforOperation(LCO)pertainingtosteamgeneratoroverfillprotectionshouldbeaddressedaspartofthereviewoftheimprovedNSSSvendortechnicalspecificationsubmittals.Inaddition,FPLreviewedthecriteriaoftheCommission'sinterimpolicystatementontechnicalspecificationimprovementspertainingtotheidentificationofLCOstobeincludedinthetechnicalspecifications.ItwasconcludedthroughuseofthecriteriaintheinterimpolicystatementthatadditionalLCOsforthesteamgeneratoroverfillprotectionsystem arenotappropriateforinclusionintheSt.LuciePlantTechnicalSpecifications.FPLbelievesthat,therequirementsfortestingthesteamgeneratoroverfillprotectionsystemcanbecontrolledthroughtheplantprocedures.Recommendation4cReassessemergencyproceduresandoperatortrainingprogramsandmodifythem,asneeded,toensurethattheoperatorscanhandlethefullspectrumofpossiblesmall-breakloss-of-coolant-accident(SBLOCA)scenarios.ThismayincludetheneedtodepressurizetheprimarysystemviatheatmosphericdumpvalvesortheturbinebypassvalvesandcooldowntheplantduringsomeSBLOCA.Thereassessmentshouldensurethatasinglefailurewouldnotnegatetheoperabilityofthevalvesneededtoachievesafeshutdown.Theproceduresshouldclearlydescribeanyactionstheoperatorisrequiredtoperformintheeventalossofinstrumentairorelectricpowerpreventsremoteoperationofthevalves.TheuseofthepressurizerPORVstodepressurizetheplantduringaSBLOCA,ifneeded,andthemeanstoensurethattheRT>>,(referencetemperature,nilductilitytransition)limitsarenotcompromised,shouldalsobeclearlydescribed.FPLResonse:St.LuciePlantEmergencyOperatingProcedure,EOP-3,LossofCoolantAccident(LOCA),providesoperatorinstructionsandcontingencyactionsforthefullspectrumofLOCAscenarios.GuidanceisprovidedwithinEOP-3forcooldownanddepressurizationoftheplant.ThisprocedureaddressescooldownanddepressurizationbymeansoftheSteamBypassControlSystem(SBCS).WhentheSBCSisunavailable,guidancecontainedinthecontingencyactionsofEOP-3isgiventousetheAtmosphericDumpValves(ADV)tocooldownanddepressurizetheplant.EOP-3iswrittentoallowcooldownanddepressurizationoftheplantwitheitheraLossofOffsitePower(LOOP)orthelossofasingleemergencyelectricaltrain.LossofinstrumentairoranelectricalcasualtymayrequiremanualoperationoftheADVs.ManualoperationoftheADVsispossibleandisincludedinoperatortrainingclasses.'heinabilitytocooldownanddepressurizetheplantusingtheSBCSortheADVsconstitutesthelossofaSafetyFunction.UponthelossofSafetyFunction,proceduralguidanceandtrainingontheuseofEmergencyOperatingProceduresdirectstheoperatortoEOP-15,FunctionalRecovery.CooldownanddepressurizationoftheplantwouldthenbeaffectedthroughuseofEOP-15,RCSandCoreHeatRemovalSuccessPath4,oncethroughcoolingusingthePilotOperatedReliefValves(PORV).TheRT>>IcriteriaaremetthroughcompliancewithFigureOne(Pressure/TemperatureCurves),whichisprovidedinallEmergencyOperatingProcedures,andPressurizedThermalShock(PTS)CriteriawhichareprovidedintheappropriateEmergencyOperatingProceduresinvolvingcomplicatedreactortrips.Additionally,FPLisparticipatinginaCombustionEngineeringOwner'sGrouptaskofconductinganassessmentofthepotentialforinadequatecorecoolingduringasmallbreakLOCA.
| | PLEASE HELP, US TO REDUCE WASTEI CONTACT THE.DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDt TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 25 |
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| | 0 P.O. Box14000, Juno Beach, FL 33408-0420 MARCH 4 9 ~99O L-90-106 10 CFR 50.54(f) |
| | U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen: |
| | Re: St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Generic Letter 89-19 Request. for Action Related to Resolution of Unresolved Safety Issue A-47 "Safety Implications of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54 f As a result of the technical resolution of USI A-47, "Safety Implications of Control Systems in LWR Nuclear Power Plants," the |
| | 'NRC concluded that protection should be provided for certain control system failures, and, for certain plants, that selected emergency procedures should be modified to assure that plant transients resulting from control system failures do not compromise public safety. |
| | Generic Letter 89-19, issued September 20, 1989, provided recommendations for control system design and procedural modifications for resolution of USI A-47. Specifically, Generic Letter 89-19 recommended that all Combustion Engineering plant designs provide automatic steam generator overfill protection to mitigate main feedwater (MFW) overfeed events, and that plant procedures and technical specifications include provisions to periodically verify the operability of the MFW overfill protection and ensure that the automatic overfill protection is operable during reactor power operation. Additionally, it that all utilities that have plants designed with high pressure was recommended injection pump discharge pressures less than or equal to 1275 psi reassess their emergency procedures and operator training programs and modify them, as needed, to ensure that the operators can handle the full spectrum of possible small-break loss of coolant accident scenarios. |
| | Florida Power 6-Light Company's response to the recommendations in Generic Letter 89-19 for St. Lucie Units 1 and 2 is attached. |
| | 9003260394 9003i9 PDR ADOCK 05000335 P PD.C ,oI an FPL Group company |
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| | U. S. Nuclear Regulatory Commission L-90-106 Page two Should there be any questions regarding the attached information, please contact us. |
| | V ry truly yours, J. H. Goldberg Executive Vice President JHG/MSD/gp Attachment cc: Stewart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie Plant II, USNRC |
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| | STATE OF FLORIDA ) |
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| | COUNTY OF PALM BEACH ) |
| | J. H. Goldber being first duly sworn, deposes and says: |
| | That he is Executive Vice President, of Florida Power & Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee. |
| | J. H. Goldbe Subscribed and sworn to before me this |
| | ~7day of |
| | 'NOTARY PUBL C, in and for th ounty of |
| | <,. Palm Peach, State of Florida Not%, pobrir, State of HorMa on Exp res lone 1 1993 Ny t:omm~ s Iasaraaca Inc. |
| | Soadad rheo Troy Fain |
| | .My,Commission expires |
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| | ATTACHMENT Response to Generic Letter 89-19, Unresolved Safety Issue Im lementation of Control S stems in LWR Power Plants~~ A-47'~Safet Back round The purpose of Generic Letter 89-19 is to ensure that all power plants have reactor vessel or steam generator overfill protection. |
| | Additionally, it was recommended that all utilities that have plants designed with high pressure injection pump discharge pressures less than or equal to 1275 psi reassess their emergency procedures and operator training programs and modify them, as needed, to ensure that the operators can handle the full spectrum of possible small-break loss of coolant accident scenarios. to Generic Letter 89-19 discusses the recommendations for Combustion Engineering (CE) Nuclear Steam Supply System pressurized water reactors like St. Lucie. The recommendations are: |
| | (1) All CE plants provide automatic steam generator overfill protection. |
| | (2) All CE plants provide plant procedures and technical specifications for periodic surveillance of the overfill protection. |
| | (3) CE plants with high pressure injection pump discharge pressures less than 1275 psi reassess their emergency procedures and operator training to ensure safe shutdown during any postulated small break loss of coolant accident. |
| | In Enclosure 2 the NRC concludes, "CE-designed plants do not provide automatic steam generator overfill protection that terminates MFW flow." FPL is of the position that, with respect to St. Lucie Units 1 and 2, this is incorrect and both units have steam generator overfill protection systems to satisfactorily resolve USI A-47. The recommendations outlined in Section 4 of are discussed below for St. Lucie Units 1 and 2. |
| | Recommendation 4a It is recommended automatic, steam that all Combustion Engineering plants provide generator overfill protection to mitigate main feedwater (MFW) overfeed events. The design for the overfill-protection system should be sufficiently separate from the MFW control system to ensure that the MFW pump will trip on a steam generator high-water-level signal when required, even if a loss of power, a loss of ventilation, or a fire in the control portion of the MFW control system should occur. Common failure modes that could disable overfill protection and the feedwater control system, but, would still result in a feedwater pump trip, are considered acceptable failure modes. |
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| | FPL Res onse: |
| | Steam Generator Overfill S stem Descri tion St. Lucie Units 1 and 2 use a 2-of-4 coincidence logic for steam generator overfill protection from four Reactor Protective System (RPS) level loops per steam generator. The loops provide high and high-,high level isolation signals. These signals result in feedwater regulating valve closu're on high steam generator level and feedwater pump trip and main turbine trip on high-high steam generator level. The initiating signals are used in non-safety related overfill protection circuits. The level loops which initiate steam generator isolation (feedwater regulating valve closure, feedwater pump trip and main turbine trip) are different from the level channels used for normal feedwater regulating valve control. |
| | The high level isolation signal from the RPS level indicating controllers provides a feedwater control override signal, which results in closure of the feedwater regulating valve for the respective train. Although the initiating signal is generated from the RPS level loops, the main feedwater regulating valve isolation function is integral to the feedwater control system. The high-high level isolation function is accomplished independently of the feedwater system via a separate circuit. Upon receipt of a high-high level isolation signal, with a 2-of-4 coincidence from either steam generator, a trip relay is energized. This relay trips both feedwater pumps and energizes a second trip relay and an .auto stop solenoid. The second trip relay provides annunciation. The auto stop solenoid provides for a main turbine trip by closing the turbine stop/control valves. |
| | The initiating logic for steam generator overfill protection is from either of two sets of four safety-grade, independent RPS transmitter loops. The signals initiate from level indicators which are mounted on the Reactor Turbine Generator Board. The St. |
| | Lucie Unit 2 signals pass through isolation cabinets; the Unit 1 signals do not. The signals then pass through common (separate circuits) "normal/bypass" control switches. The high level signal then ties into the feedwater regulating system modules. The high-high level signal ties into the two trip relays and turbine stop solenoid. |
| | Review of the power supplies for the feedwater control system and the steam generator overfill protection circuit shows they are from different sources. The normal power supplies for a single feedwater control train and the overfill protection system are ultimately fed from a common 480 volt switchgear (Unit 1) or 4160 volt bus (Unit 2); however, the feedwater control system is supplied 120 volt AC power and the steam generator overfill protection system is provided 125 volt DC. The steam generator overfill protection circuit is provided back-up emergency power from the 2B(1A) station battery in the event normal power is lost. |
| | Each train of the feedwater control system is also provided with an automatic transfer to a back-up power source from the 120 volt vital AC cabinets, to ensure further reliability. This source is |
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| | ~ t ultimately powered from the 2AB(lAB) swing bus, which can be manually aligned to either train "A" or "B" 480 volt AC switchgear. |
| | Emergency back-up power is also provided via the 2D(1D) station battery and associated inverter. |
| | Common Failure Modes Discussion Various potential common mode failures between the feedwater control system and the steam generator overfill protection circuits were evaluated. The limiting common mode failures evaluated were for a loss of power, a common fire, and failure of the shared override/manual control switch. |
| | Power Loss The normal power supplies for the feedwater control and steam generator overfill protection circuits are not totally independent (as discussed above); however, the back-up supplies are sufficiently independent to ensure operability of one or both systems in the event a common normal bus is lost. The steam generator overfill protection circuitry is provided back-up 125 volt DC power from safety-related station battery 2B(1A), such that the feedwater pump trip should still occur if a high-high steam generator level condition occurs coincident with a loss of normal power. Similarly, the feedwater control system is provided with an automatic transfer to the 120 volt AC vital bus (fed off the "AB" 480 volt Motor Control Center) in the event normal power is lost. This feedwater control power source is further backed-up by the 2D(1D) battery. Furthermore, the design of the feedwater control system requires the feedwater regulating valve to fail closed on a loss of power, such that even in the remote event (multiple failure) the common bus and the back-up power supplies are lost, feedwater flow will still be isolated for the affected train. |
| | Fire In the event of a fire in the control room, the feedwater control and overfill protection systems are not sufficiently separate to ensure the operability of at least one system. The two systems have circuitry, which share common "Norm/bypass" handswitches in the control room, common cable tray routing and are located within adjacent control room panels. Given a fire of a large magnitude, which renders the control room uninhabitable, a manual reactor/turbine trip would be initiated prior to leaving the control room and safe shutdown would be accomplished from the remote shutdown panel. Main feedwater flow would be isolated per operating procedures and steam generator level is controlled via auxiliary feedwater. |
| | Shared Control Switch The final common mode failure postulated was the failure of a common "normal/bypass" handswitch due to inadvertent mispositioning, contact failure "open," or short to ground. |
| | |
| | Inadvertent mispositioning of the handswitches is not credible, as annunciation is provided to warn mispositioned. Furthermore, if if any of the four switches is a single handswitch is mispositioned or its contacts fail open, the high and high-high overfill protection are still operable. A 2-of-3 coincidence will now be required to initiate a high/high-high steam generator isolation signal. In the event a single switch shorts to ground, both the feedwater control and steam generator overfill protection circuits lose power and the feedwater regulating valves fail close. |
| | Based on the above, FPL concludes that the steam generator overfill protection system currently installed in St. Lucie Units 1 and 2 provides adequate assurance that a steam generator overfill event will be prevented. |
| | Recommendation 4b It is recommended that plant specifications for all Combustion procedures and technical Engineering plants include provisions to verify periodically the operability of overfill protection and ensure that automatic main feedwater overfill protection is operable during reactor power operation. The instrumentation should be demonstrated to be operable by the performance of a channel check, channel functional testing, and channel calibration, including setpoint verification and by identifying the LCOs. These technical specifications should be commensurate with existing plant technical specification requirements for channels that initiate protection actions. |
| | FPL Res onse: |
| | St. Lucie Plant procedures include provisions to periodically verify. steam generator overfill protection and ensure that automatic main feedwater overfill protection is operable during reactor power operation. Testing of the steam generator high and high-high level circuitry is completed on an 18 month refueling basis by functionally testing the circuitry using the 2-of-4 logic to close the feedwater regulating valve, trip the feedwater pump and trip the turbine. |
| | The NRC staff is currently reviewing the Nuclear Steam Supply System (NSSS) vendor specific revised standard technical specifications developed using the criteria of the NRC interim policy statement on technical specification improvements (Federal Register p. 3788/Vol. 52 No. 25/Friday, February 6, 1987). |
| | Recommendations for specific Limiting Conditions for Operation (LCO) pertaining to steam generator overfill protection should be addressed as part of the review of the improved NSSS vendor technical specification submittals. In addition, FPL reviewed the criteria of the Commission's interim policy statement on technical specification improvements pertaining to the identification of LCOs to be included in the technical specifications. It was concluded through use of the criteria in the interim policy statement that additional LCOs for the steam generator overfill protection system |
| | |
| | are not appropriate for inclusion in the St. Lucie Plant Technical Specifications. FPL believes that, the requirements for testing the steam generator overfill protection system can be controlled through the plant procedures. |
| | Recommendation 4c Reassess emergency procedures and operator training programs and modify them, as needed, to ensure that the operators can handle the full spectrum of possible small-break loss-of-coolant-accident (SBLOCA) scenarios. This may include the need to depressurize the primary system via the atmospheric dump valves or the turbine bypass valves and cool down the plant during some SBLOCA. The reassessment should ensure that a single failure would not negate the operability of the valves needed to achieve safe shutdown. The procedures should clearly describe any actions the operator is required to perform in the event a loss of instrument air or electric power prevents remote operation of the valves. The use of the pressurizer PORVs to depressurize the plant during a SBLOCA, if needed, and the means to ensure that the RT>>, (reference temperature, nil ductility transition) limits are not compromised, should also be clearly described. |
| | FPL Res onse: |
| | St. Lucie Plant Emergency Operating Procedure, EOP-3, Loss of Coolant Accident (LOCA), provides operator instructions and contingency actions for the full spectrum of LOCA scenarios. |
| | Guidance is provided within EOP-3 for cooldown and depressurization of the plant. This procedure addresses cooldown and depressurization by means of the Steam Bypass Control System (SBCS). When the SBCS is unavailable, guidance contained in the contingency actions of EOP-3 is given to use the Atmospheric Dump Valves (ADV) to cooldown and depressurize the plant. EOP-3 is written to allow cooldown and depressurization of the plant with either a Loss of Offsite Power (LOOP) or the loss of a single emergency electrical train. Loss of instrument air or an electrical casualty may require manual operation of the ADVs. |
| | Manual operation of the ADVs is possible and is included in operator training classes. 'he inability to cooldown and depressurize the plant using the SBCS or the ADVs constitutes the loss of a Safety Function. Upon the loss of Safety Function, procedural guidance and training on the use of Emergency Operating Procedures directs the operator to EOP-15, Functional Recovery. |
| | Cooldown and depressurization of the plant would then be affected through use of EOP-15, RCS and Core Heat Removal Success Path 4, once through cooling using the Pilot Operated Relief Valves (PORV). |
| | The RT>>I criteria are met through compliance with Figure One (Pressure/Temperature Curves), which is provided in all Emergency Operating Procedures, and Pressurized Thermal Shock (PTS) Criteria which are provided in the appropriate Emergency Operating Procedures involving complicated reactor trips. |
| | Additionally, FPL is participating in a Combustion Engineering Owner's Group task of conducting an assessment of the potential for inadequate core cooling during a small break LOCA.}} |
|
|---|
Category:CORRESPONDENCE-LETTERS
MONTHYEARML17241A5001999-10-21021 October 1999
Forwards Rev 3 to Emergency Response Data Sys (ERDS) Data Point Library for St Lucie Unit 1.Rev Provides Replacement Pages & Follows Format Recommended by NUREG 1394, ERDS Implementation, Rev 1,App C
ML17309A9981999-10-19019 October 1999
Forwards Revised Epips,Including Rev 3 to EPIP-10 & Rev 25 to HP-202.EPIP-10 Added Onsite Monitoring Points,Made Administrative Changes & Incorporated New Attachments & HP-202 Added Red Team Survey Points
ML20217F6171999-10-0808 October 1999
Forwards Insp Repts 50-335/99-11 & 50-389/99-11 on 990827 & 990907-09.No Violations Identified.Matl Encl Contained Safeguards Info as Defined by 10CFR73.21 & Disclosed to Unauthorized Individuals Prohibited by Section 147 of AEA
ML17241A4811999-10-0101 October 1999
Reports Number of Tubes Plugged During Unit 1 Refueling Outage SL1-16,per TS 4.4.5.5.a
ML20212M1601999-09-28028 September 1999
Refers to 990908 Engineering Meeting Conducted at NRC Region II to Discuss Engineering Issues at Lucie & Turkey Point Facilities.List of Attendees & Copy of Presentation Handout Encl
ML17241A4701999-09-25025 September 1999
Forwards Info Requested by NRC Staff During 990916 Telcon to Complete Staff Review of Request for risk-informed Extension of Action Completion/Aot Specified for Inoperable Train of LPSI Sys at Plant
ML17241A4721999-09-24024 September 1999
Forwards Rev 1 to Plant Change/Mod (PCM) 99016 to St Lucie Unit 1,Cycle 16 COLR, IAW TS 6.9.1.11.d.Refueling Overhaul Activities Are Currently in Progress & Reactor Operations for Cycle 16 Are Scheduled to Commence in Oct 1999
ML17241A4681999-09-22022 September 1999
Requests Restriction Be Added to Senior Operator License SOP-21093 for TE Bolander.Nrc Forms 369,encl.Encl Withheld Per 10CFR2.790(a)(6)
ML17241A4671999-09-20020 September 1999
Forwards Completed NRC Form 536, Operator Licensing Exam Data, for St Lucie Units 1 & 2,as Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams.
ML17241A4581999-09-13013 September 1999
Forwards Info Requested by NRC Staff During 990630 & 0816 Telcons,To Complete Review of Proposed License Amend for Fuel Reload Process Improvement Program
ML17241A4531999-08-31031 August 1999
Informs That No Candidates from St Lucie Plant Will Be Participating in PWR Gfes Being Administered on 991006
ML17241A4521999-08-31031 August 1999
Withdraws Relief Request 16 & Suppl Relief Request 15 with Info Requested During 990526 Telephone Conference Re ISI Insp Plan,Third 10-yr Interval
ML17241A4501999-08-26026 August 1999
Informs That FPL Has Reviewed Reactor Vessel Integrity Database,Called RVID2,re Closure of GL 92-01,rev 1,suppl 1. Requested Corrections & Marked Up Pages from Rvid 2 Database Summary Repts That Correspond to Comments,Attached
ML17241A4371999-08-13013 August 1999
Forwards fitness-for-duty Program Performance Data for six- Month Period Ending 990630,per 10CFR26.71(d)
ML17241A4461999-08-11011 August 1999
Requests That W Rept Entitled, Evaluation of Turbine Missile Ejection Probability Resulting from Extending Test Interval of Interceptor & Reheat Stop Valves at St Lucie Units 1 & 2, Be Withheld from Public Disclosure
L-99-171, Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3)1999-07-29029 July 1999
Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3)
ML17309A9911999-07-26026 July 1999
Forwards Revised EPIPs & Revised Procedures That Implement Emergency Plan as Listed.Procedures Provides Instruction for Operational Support Ctr (OSC) Chemistry Supervisor to Establish Remote Labs at Locations Specified
ML17241A4471999-07-22022 July 1999
Requests That Rev 1 to WCAP-14732 & Rev 1,Add 1 to WCAP-14732 Be Withheld from Public Disclosure
ML17241A4221999-07-22022 July 1999
Forwards List of Proposed Licensing Actions for St Lucie Units 1 & 2,planned During Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates.
ML17241A4151999-07-22022 July 1999
Forwards Revised Relief Request 25 for Second 10-yr ISI Interval for Unit 2
ML17241A4101999-07-16016 July 1999
Forwards FP&L Supplemental Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants
ML17309A9881999-07-0707 July 1999
Forwards Rev 5 to EPIP-03, Emergency Response Organization Notification/Staff Augmentation. Rev 5 to EPIP-03 Was Revised to Transfer EP Responsibilities from Training Manager to Protection Svcs Manager
ML20209F1541999-07-0606 July 1999
Informs That NRC in Process of Conducting Operational Safeguards Response Evaluations at Nuclear Power Reactors. Plant Chosen for Such Review Scheduled for Wk of 990823-26
ML17241A4011999-06-30030 June 1999
Forwards Info Copy of Florida Wastewater Permit (FL0002208) (Formerly NPDES Permit) Mod,Which Was Issued by Florida Dept of Environ Protection on 990604
ML17241A3971999-06-30030 June 1999
Forwards Suppl Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, as Requested in 990317 Ltr
ML17355A3661999-06-30030 June 1999
Forwards Florida Power & Light Topical QA Rept, Dtd June 1999.Encl I Includes Summary of Changes Made to Topical QA Rept Since 1998
ML17241A3951999-06-29029 June 1999
Provides Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants, Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl
ML17241A3731999-06-17017 June 1999
Supplements Relief Requests 4,11 & 13 for Third ten-year ISI Interval with Info Requested During 990526 Telcon.Expedited Review Is Requested by 990730 to Avoid Negatively Impacting Upcoming St Lucie Unit 1 Refueling Outage (SL1-16)
ML17241A3641999-06-14014 June 1999
Submits Supplement to Relief Request 24 with Info Requested by Nrc.In Addition Relief Request 24 Is Identical to St Lucie Unit 1 Relief Request 4 for Third ISI Interval Being Supplemented by FPL Ltr L-99-139
ML20195F3871999-06-11011 June 1999
Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5)
IA-99-247, Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5)1999-06-11011 June 1999
Final Response to FOIA Request for Documents.App a Records Being Withheld in Entirety (Ref FOIA Exemption 5)
L-99-129, Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3)1999-06-0909 June 1999
Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3)
ML17241A3561999-06-0707 June 1999
Forwards Rept Containing Brief Description & Summary of SEs for Changes,Tests & Experiments Which Were Approved for Unit 3 During Period of 970526-981209
ML17241A3601999-06-0707 June 1999
Forwards Correction to Annual Radiological Environ Operating Rept for CY98.Util Has Identified Transcription Error on Last Page of Attachment C of Rept,Results from Interlaboratory Comparison Program 1998
ML20195F3941999-05-27027 May 1999
FOIA Request That Memo from J Calvo to Fl Lebdon Re TIA - St Lucie,Unit 1 Environ Qualification of Woodward Governor Controls Be Placed in PDR
ML17241A3461999-05-24024 May 1999
Forwards Revised Relief Request 22 to Clarify Several Areas of Relief.Nrc Action Is Requested to Be Complete by Aug 1999 to Support Planning for Spring 2000 Unit 2 Refueling Outage
ML17241A3391999-05-20020 May 1999
Forwards Notification of Change to Small Break LOCA ECCS Evaluation Model Used for St Lucie Unit 1.Anomaly Was Discovered & Corrected That Resulted in Reducing Calculated PCT for Limiting SBLOCA by More than 50 F
ML17241A3371999-05-20020 May 1999
Forwards Util Suppl to GL 95-07 Response Re pressure-locking & Thermal Binding of safety-related power-operated Gate Valves,In Response to NRC Second RAI Dtd 990225
ML20207C7531999-05-17017 May 1999
Discusses Issue Identified by FPL in Feb 1998 Involving Potential for Fire to Cause Breach of Rc Sys High/Low Pressure Interface Boundary & NRC Decision for Exercise of Enforcement Discretion
ML17241A3301999-05-17017 May 1999
Forwards LER 99-004-00 Re as Found Cycle 10 Psv Setpoints Outside TS Limits,Which Occurred on 990415.Root Cause Determination Not Yet Complete.Suppl to Include Root Cause & Corrective Actions Will Be Submitted
ML17309A9821999-05-10010 May 1999
Forwards Rev 36 to St Lucie Emergency Plan, Per 10CFR50.54(q).Executive Summary & Summary of Changes Incorporated by Rev,Encl
IR 05000335/19980141999-04-29029 April 1999
Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated
ML17241A3221999-04-29029 April 1999
Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated
ML17229B1071999-04-28028 April 1999
Forwards 1998 Annual Environ Operating Rept for St Lucie Unit 2. Rept Includes Discussions of 5-inch Barrier Net Maint & Taprogge Condenser Tube Cleaning Sys Ball Loss,As Agreed at First Biennial Sea Turtle Meeting Held on 980120
ML17229B1051999-04-22022 April 1999
Requests That Listed Individuals Be Placed on Official Serve List for Nuclear Matl Safety & Safeguards Info Notices
ML17229B1061999-04-21021 April 1999
Notifies NRC of Change in Medical Status of Licensed Operator Pf Farnsworth (Docket 55-21285,license SOP-21094). NRC Form 3996, Medical Exam Certification, Encl.Encl Withheld Per 10CFR2.790(a)(6)
ML17309A9851999-04-15015 April 1999
Requests That NRC Review Denial of Appeal from Assessment of Fees Assessed in 981101 Invoice RS0062-99 & Assessment of Fees in Invoice RS0182-99 Which Was Also Denied in 990305 Ltr.Both Invoices Are for Fees Re Inspector GG Warnick
ML20205M0431999-04-13013 April 1999
Eighth Partial Response to FOIA Request for Records.App Q & R Records Encl & Being Made Available in PDR
ML17229B0951999-04-0808 April 1999
Requests Approval of Encl Revised Relief Request 6,in Response to 990322 Telcon with NRC & 10CFR55.55a(a)(3). Request States That Visual VT-3 Exams Will Be Conducted IAW IWA-2213 & Repairs Will Be IAW Util ASME Section IX Program
ML17229B0821999-04-0707 April 1999
Requests Approval of Interim Relief Request 26 Re Repair Requirements for Class 2 ECCS Piping,Per 10CFR50.55a(a)(3) & 50.55a(g)(iii).Alternative Actions Apply Guidance of GLs 91-18 & 90-05 & ASME Code Case N-513.Evaluation,encl
1999-09-28
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML17241A5001999-10-21021 October 1999
Forwards Rev 3 to Emergency Response Data Sys (ERDS) Data Point Library for St Lucie Unit 1.Rev Provides Replacement Pages & Follows Format Recommended by NUREG 1394, ERDS Implementation, Rev 1,App C
ML17309A9981999-10-19019 October 1999
Forwards Revised Epips,Including Rev 3 to EPIP-10 & Rev 25 to HP-202.EPIP-10 Added Onsite Monitoring Points,Made Administrative Changes & Incorporated New Attachments & HP-202 Added Red Team Survey Points
ML17241A4811999-10-0101 October 1999
Reports Number of Tubes Plugged During Unit 1 Refueling Outage SL1-16,per TS 4.4.5.5.a
ML17241A4701999-09-25025 September 1999
Forwards Info Requested by NRC Staff During 990916 Telcon to Complete Staff Review of Request for risk-informed Extension of Action Completion/Aot Specified for Inoperable Train of LPSI Sys at Plant
ML17241A4721999-09-24024 September 1999
Forwards Rev 1 to Plant Change/Mod (PCM) 99016 to St Lucie Unit 1,Cycle 16 COLR, IAW TS 6.9.1.11.d.Refueling Overhaul Activities Are Currently in Progress & Reactor Operations for Cycle 16 Are Scheduled to Commence in Oct 1999
ML17241A4681999-09-22022 September 1999
Requests Restriction Be Added to Senior Operator License SOP-21093 for TE Bolander.Nrc Forms 369,encl.Encl Withheld Per 10CFR2.790(a)(6)
ML17241A4671999-09-20020 September 1999
Forwards Completed NRC Form 536, Operator Licensing Exam Data, for St Lucie Units 1 & 2,as Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams.
ML17241A4581999-09-13013 September 1999
Forwards Info Requested by NRC Staff During 990630 & 0816 Telcons,To Complete Review of Proposed License Amend for Fuel Reload Process Improvement Program
ML17241A4531999-08-31031 August 1999
Informs That No Candidates from St Lucie Plant Will Be Participating in PWR Gfes Being Administered on 991006
ML17241A4521999-08-31031 August 1999
Withdraws Relief Request 16 & Suppl Relief Request 15 with Info Requested During 990526 Telephone Conference Re ISI Insp Plan,Third 10-yr Interval
ML17241A4501999-08-26026 August 1999
Informs That FPL Has Reviewed Reactor Vessel Integrity Database,Called RVID2,re Closure of GL 92-01,rev 1,suppl 1. Requested Corrections & Marked Up Pages from Rvid 2 Database Summary Repts That Correspond to Comments,Attached
ML17241A4371999-08-13013 August 1999
Forwards fitness-for-duty Program Performance Data for six- Month Period Ending 990630,per 10CFR26.71(d)
ML17241A4461999-08-11011 August 1999
Requests That W Rept Entitled, Evaluation of Turbine Missile Ejection Probability Resulting from Extending Test Interval of Interceptor & Reheat Stop Valves at St Lucie Units 1 & 2, Be Withheld from Public Disclosure
L-99-171, Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3)1999-07-29029 July 1999
Forwards Rev 56 to Physical Security Plan.Summary of Changes & Marked Up Copy of Revised Pages Also Encl.Encls Withheld from Public Disclosure Per 10CFR2.790(a)(3)
ML17309A9911999-07-26026 July 1999
Forwards Revised EPIPs & Revised Procedures That Implement Emergency Plan as Listed.Procedures Provides Instruction for Operational Support Ctr (OSC) Chemistry Supervisor to Establish Remote Labs at Locations Specified
ML17241A4221999-07-22022 July 1999
Forwards List of Proposed Licensing Actions for St Lucie Units 1 & 2,planned During Fys 2000 & 2001,in Response to Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates.
ML17241A4471999-07-22022 July 1999
Requests That Rev 1 to WCAP-14732 & Rev 1,Add 1 to WCAP-14732 Be Withheld from Public Disclosure
ML17241A4151999-07-22022 July 1999
Forwards Revised Relief Request 25 for Second 10-yr ISI Interval for Unit 2
ML17241A4101999-07-16016 July 1999
Forwards FP&L Supplemental Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants
ML17309A9881999-07-0707 July 1999
Forwards Rev 5 to EPIP-03, Emergency Response Organization Notification/Staff Augmentation. Rev 5 to EPIP-03 Was Revised to Transfer EP Responsibilities from Training Manager to Protection Svcs Manager
ML17241A4011999-06-30030 June 1999
Forwards Info Copy of Florida Wastewater Permit (FL0002208) (Formerly NPDES Permit) Mod,Which Was Issued by Florida Dept of Environ Protection on 990604
ML17241A3971999-06-30030 June 1999
Forwards Suppl Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, as Requested in 990317 Ltr
ML17355A3661999-06-30030 June 1999
Forwards Florida Power & Light Topical QA Rept, Dtd June 1999.Encl I Includes Summary of Changes Made to Topical QA Rept Since 1998
ML17241A3951999-06-29029 June 1999
Provides Response to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants, Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl
ML17241A3731999-06-17017 June 1999
Supplements Relief Requests 4,11 & 13 for Third ten-year ISI Interval with Info Requested During 990526 Telcon.Expedited Review Is Requested by 990730 to Avoid Negatively Impacting Upcoming St Lucie Unit 1 Refueling Outage (SL1-16)
ML17241A3641999-06-14014 June 1999
Submits Supplement to Relief Request 24 with Info Requested by Nrc.In Addition Relief Request 24 Is Identical to St Lucie Unit 1 Relief Request 4 for Third ISI Interval Being Supplemented by FPL Ltr L-99-139
L-99-129, Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3)1999-06-0909 June 1999
Forwards Rev 55 to Physical Security Plan,Summary of Changes & Marked Up Copy of Revised Pages.With Directions for Incorporating Rev Into Plan & Copies of Replacement Pages.Rev Withheld,Per 10CFR2.790(a)(3)
ML17241A3601999-06-0707 June 1999
Forwards Correction to Annual Radiological Environ Operating Rept for CY98.Util Has Identified Transcription Error on Last Page of Attachment C of Rept,Results from Interlaboratory Comparison Program 1998
ML17241A3561999-06-0707 June 1999
Forwards Rept Containing Brief Description & Summary of SEs for Changes,Tests & Experiments Which Were Approved for Unit 3 During Period of 970526-981209
ML20195F3941999-05-27027 May 1999
FOIA Request That Memo from J Calvo to Fl Lebdon Re TIA - St Lucie,Unit 1 Environ Qualification of Woodward Governor Controls Be Placed in PDR
ML17241A3461999-05-24024 May 1999
Forwards Revised Relief Request 22 to Clarify Several Areas of Relief.Nrc Action Is Requested to Be Complete by Aug 1999 to Support Planning for Spring 2000 Unit 2 Refueling Outage
ML17241A3371999-05-20020 May 1999
Forwards Util Suppl to GL 95-07 Response Re pressure-locking & Thermal Binding of safety-related power-operated Gate Valves,In Response to NRC Second RAI Dtd 990225
ML17241A3391999-05-20020 May 1999
Forwards Notification of Change to Small Break LOCA ECCS Evaluation Model Used for St Lucie Unit 1.Anomaly Was Discovered & Corrected That Resulted in Reducing Calculated PCT for Limiting SBLOCA by More than 50 F
ML17241A3301999-05-17017 May 1999
Forwards LER 99-004-00 Re as Found Cycle 10 Psv Setpoints Outside TS Limits,Which Occurred on 990415.Root Cause Determination Not Yet Complete.Suppl to Include Root Cause & Corrective Actions Will Be Submitted
ML17309A9821999-05-10010 May 1999
Forwards Rev 36 to St Lucie Emergency Plan, Per 10CFR50.54(q).Executive Summary & Summary of Changes Incorporated by Rev,Encl
ML17241A3221999-04-29029 April 1999
Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated
IR 05000335/19980141999-04-29029 April 1999
Provides Confirmation of NRC Staff Conclusions Re Cited & non-cited Violations in Insp Rept 50-335/98-14 & 50-389/98-14.Utils Position Re Consideration of Multiple Spurious Actuations in Event of Fire,Reiterated
ML17229B1071999-04-28028 April 1999
Forwards 1998 Annual Environ Operating Rept for St Lucie Unit 2. Rept Includes Discussions of 5-inch Barrier Net Maint & Taprogge Condenser Tube Cleaning Sys Ball Loss,As Agreed at First Biennial Sea Turtle Meeting Held on 980120
ML17229B1051999-04-22022 April 1999
Requests That Listed Individuals Be Placed on Official Serve List for Nuclear Matl Safety & Safeguards Info Notices
ML17229B1061999-04-21021 April 1999
Notifies NRC of Change in Medical Status of Licensed Operator Pf Farnsworth (Docket 55-21285,license SOP-21094). NRC Form 3996, Medical Exam Certification, Encl.Encl Withheld Per 10CFR2.790(a)(6)
ML17309A9851999-04-15015 April 1999
Requests That NRC Review Denial of Appeal from Assessment of Fees Assessed in 981101 Invoice RS0062-99 & Assessment of Fees in Invoice RS0182-99 Which Was Also Denied in 990305 Ltr.Both Invoices Are for Fees Re Inspector GG Warnick
ML17229B0951999-04-0808 April 1999
Requests Approval of Encl Revised Relief Request 6,in Response to 990322 Telcon with NRC & 10CFR55.55a(a)(3). Request States That Visual VT-3 Exams Will Be Conducted IAW IWA-2213 & Repairs Will Be IAW Util ASME Section IX Program
ML17229B0821999-04-0707 April 1999
Requests Approval of Interim Relief Request 26 Re Repair Requirements for Class 2 ECCS Piping,Per 10CFR50.55a(a)(3) & 50.55a(g)(iii).Alternative Actions Apply Guidance of GLs 91-18 & 90-05 & ASME Code Case N-513.Evaluation,encl
ML17229B0851999-04-0505 April 1999
Requests Approval of Encl Relief Request 25 Which Proposes to Use Alternative Requirements of ASME Code Case N-613 in Lieu of Requirements of ASME Section XI Figures IWB-2500-7(a) & IWB-2500-7(b).Action Requested by Aug 1999
ML17309A9791999-03-31031 March 1999
Forwards Revised EPIPs Including Rev 2 to EPIP-00,rev 2 to EPIP-09,rev 2 to EPIP-10 & Rev 10 to HP-207.Summary of Revs Listed
ML17309A9761999-03-23023 March 1999
Forwards Revised Epips,Including Rev 4 to EPIP-03, Er Organization Notification/Staff Augmentation, Rev 3 to EPIP-05, Activation & Operation of OSC & Rev 14 to HP-200, HP Emergency Organization. Changes to Epips,Discussed
ML17229B0691999-03-19019 March 1999
Transmits TS Pages Requested by NRC for Use in Issuance of Proposed License Amend Re SFP Storage Capacity,Per Soluble Boron Credit
ML17229B0721999-03-16016 March 1999
Requests Approval of Enclosed Relief Requests 23 & 24 Re ISI Plan for Second ten-year Interval.Nrc Action Is Requested to Be Complete by Aug 1999 to Support Planning for Spring 2000 Unit 2 Refueling Outage
ML17355A2631999-03-12012 March 1999
Forwards FPL Decommissioning Fund Status Repts for St Lucie, Units 1 & 2 & Turkey Point,Units 3 & 4.Rept for St Lucie, Unit 2 Provides Status of Decommissioning Funds for All Three Owners of That Unit
ML17229B0481999-03-10010 March 1999
Informs That Util Delivered Matls Requested in Encl 1 of NRC Ltr by Hand on 990308,as Requested by NRC Ltr Dtd 990218
1999-09-25
[Table view]
Category:UTILITY TO NRC
MONTHYEARML17223A9401990-09-13013 September 1990
Forwards Evaluation of Potential Safety Impact of Failed Control Element Assemblies on Limiting Transients for Facility
ML17223A9341990-09-10010 September 1990
Forwards Addl Info Re Generic Implications & Resolution of Control Element Assembly (CEA) Failure at Facility,Per NRC Request.Description of Testing Program for Old Style CEAs in Unit 1 Core Encl
L-90-315, Advises That Util Has Completed Evaluation of NUREG-0737, Item II.D.1,SER Item 81990-08-30030 August 1990
Advises That Util Has Completed Evaluation of NUREG-0737, Item II.D.1,SER Item 8
ML17223A9201990-08-28028 August 1990
Forwards Forms NIS-1 & NIS-2, Owners Rept for Inservice Insps as Required by Provisions of ASME Code Rules, Per 900725 Ltr
ML17223A8911990-08-20020 August 1990
Forwards Corrected Monthly Operating Repts for Jul 1990 for St Lucie Units 1 & 2 & Summary of Operating Experience
ML17348A5041990-08-17017 August 1990
Forwards fitness-for-duty Program Performance Data for Jan-June 1990
L-90-301, Discusses Generic Implications & Resolution of Control Element Assemblies Failure at Plant1990-08-16016 August 1990
Discusses Generic Implications & Resolution of Control Element Assemblies Failure at Plant
ML17223A8751990-08-0909 August 1990
Responds to Violations Noted in Insp Rept 50-335/90-14. Corrective Actions:Rcs Flow Determination by Calorimetric Procedure Repeated W/Supervisor of Individual Observing & Individual Counseled by Supervisor
IR 05000335/19900141990-08-0909 August 1990
Responds to Violations Noted in Insp Rept 50-335/90-14. Corrective Actions:Rcs Flow Determination by Calorimetric Procedure Repeated W/Supervisor of Individual Observing & Individual Counseled by Supervisor
ML17348A4701990-07-27027 July 1990
Forwards Rept Detailing Investigative Analysis of Unsatisfactory Blind Specimen Results,Identification of Causes & Corrective Actions Taken by Lab to Prevent Recurrence,Per Unsatisfactory Performance Testing
ML17223A8621990-07-25025 July 1990
Advises That NIS-1 & NIS-2 Forms,As Part of Inservice Insp Rept,Will Be Submitted by 900831
ML17348A4281990-07-25025 July 1990
Forwards Decommissioning Financial Assurance Repts for Plants,Per 10CFR50.33(k) & 50.75(b)
ML17223A8631990-07-25025 July 1990
Submits Addl Info Re Implementation of Programmed Enhancements Per Generic Ltr 88-17, Loss of Dhr. All Mods for Unit 1 Completed & Operational.Mods for Unit 2 Schedule for Upcoming Refueling Outage
L-90-271, Responds to NRC Ltr Re Violations Noted in Insp Repts 50-335/90-09 & 50-389/90-09.Corrective Actions:Procedural Expectation Re Hanging & Removal of Deficiency Tags Will Be Reemphasized to Personnel Generating Work Orders1990-07-20020 July 1990
Responds to NRC Ltr Re Violations Noted in Insp Repts 50-335/90-09 & 50-389/90-09.Corrective Actions:Procedural Expectation Re Hanging & Removal of Deficiency Tags Will Be Reemphasized to Personnel Generating Work Orders
ML17223A8581990-07-19019 July 1990
Forwards Implementation Status of 10CFR50.62 Mod at Facility Re Requirements for Reduction of Risk from ATWS Events for Light Water Cooled Nuclear Power Plants
ML17223A8491990-07-18018 July 1990
Responds to NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. No Rosemount Transmitters Models 1153 Series B,1153 Series D & 1154 Mfg Prior to 890711 Supplied by Different Vendor
ML17223A8521990-07-17017 July 1990
Forwards Addl Info Requested Re Generic Implications & Resolution of Control Element Assembly Failure at Plant.Encl Confirms Util Intent to Follow C-E Regulatory Response Group Action Program
IR 05000335/19900131990-07-0909 July 1990
Responds to Violations Noted in Insp Repts 50-335/90-13 & 50-389/90-13.Corrective Actions:Maint Personnel Counseled & Aware of Importance of Verifying Design Configuration Requirements
ML17223A8421990-07-0909 July 1990
Responds to Violations Noted in Insp Repts 50-335/90-13 & 50-389/90-13.Corrective Actions:Maint Personnel Counseled & Aware of Importance of Verifying Design Configuration Requirements
ML17348A3881990-07-0505 July 1990
Requests Audit of NRC Records to Independently Verify Reasonableness of Charges Assessed Against Util,Per 10CFR170 Svcs
ML17223A8391990-07-0303 July 1990
Forwards Results of Beach Survey Procedure & Reduction of Field Survey Data,Per Tech Spec 4.7.6.1.1.Unit 1 Updated Fsar,Section 2.4.2.2,concluded That Dune Condition Acceptable Per Tech Spec 5.1.3
ML17223A8381990-07-0202 July 1990
Requests Termination of Operator License for s Lavelle.Util Also Requests That Ltr Be Withheld (Ref 10CFR2.790)
L-90-239, Forwards Rev 6 to Guard Training & Qualification Plan.Rev Withheld (Ref 10CFR73.21)1990-07-0202 July 1990
Forwards Rev 6 to Guard Training & Qualification Plan.Rev Withheld (Ref 10CFR73.21)
ML17223A8371990-06-27027 June 1990
Provides Details of Implementation Plan Re Recommendations & Schedular Requirements in Generic Ltr 89-10,per 891228 Ltr.Design Basis Review of safety-related motor-operated Valves & Determination of Switch Settings in Progress
ML17308A4981990-06-27027 June 1990
Responds to Generic Ltr 90-04 Re Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions
ML17223A8341990-06-19019 June 1990
Forwards Corrected Proposed Tech Spec Figure 3.4-2 Per 900207 Application for Amend to License NPF-16,incorporating Revised Pressure/Temp Limits & Results of Revised Low Temp Overpressure Protection Analysis Into Tech Specs
ML17223A8241990-06-18018 June 1990
Forwards Revised Combined Semiannual Radioactive Effluent Release Rept for Jan-June 1988.
ML17223A8271990-06-18018 June 1990
Forwards Ma Smith 900601 Ltr to WR Cunningham of EPA Requesting Mod to Plant NPDES Permit to Permit Cleaning of Facility & to Establish Discharge Limits for Chemical Cleaning Wastes
ML17348A2981990-06-12012 June 1990
Forwards Rev 16 to Topical QA Rept.
ML17223A6761990-05-31031 May 1990
Advises That Air Operated safety-related Components Will Perform All Design Basis Events,Per 881227 Ltr.All Actions Required by Generic Ltr 88-14 Complete for Plant
ML17348A2651990-05-29029 May 1990
Submits Rept Detailing Investigative Analysis of Unsatisfactory Blind Specimen Results,Identification of Causes & Corrective Actions Taken by Lab to Prevent Recurrence,Per 10CFR26,App A.2.8(e)(4)
ML17223A6741990-05-22022 May 1990
Forwards Info Re Status of 10CFR50.62 Mods to Meet ATWS Requirements as of 900515.Plant Change/Mod Package Necessary for Installing ATWS Will Be Issued by 900630.Hardware Procurement for Diverse Scram Sys Approx 90% Complete
ML17223A6361990-05-0808 May 1990
Forwards Final Response to NRC Bulletin 88-010, Nonconforming Molded-Case Circuit Breakers. One Untraceable Circuit Breaker Installed in Unit 2 Qualified SPDS & Replaced W/Traceable Breaker
ML17223A6281990-04-21021 April 1990
Forwards St Lucie Unit 2 Annual Environ Operating Rept, Vol 1 1989.
ML17223A6081990-04-13013 April 1990
Responds to Violations Noted in Insp Repts 50-335/90-02 & 50-389/90-02.Corrective Actions:Nuclear Plant Supervisor Required to Remain in Control Room During Significant Changes in Power Operation & Preventive Maint Upgraded
ML17223A6071990-04-0505 April 1990
Responds to NRC Bulletin 89-001, Failure of Westinghouse Steam Generator Tube Mechanical Plugs. Removal & Replacement of Cold Leg Side Plugs of Heat Number 3513 for Unit 1 Completed During Refueling Outage
ML17308A4911990-04-0202 April 1990
Forwards Description & Summary of Safety Evaluations of Plant Changes/Mods Reportable Per 10CFR50.59.Repair &/Or Replacement of Protective Coatings on Surfaces Inside Bldg Pose No Unreviewed Safety Question
ML17223A5931990-03-30030 March 1990
Forwards Status of 10CFR50.62, Requirements for Reduction of Risk from ATWS Mods at Plant as of 900315.Diverse Scram Sys Module Prototype Fabrication in Progress
ML17223A5921990-03-27027 March 1990
Forwards Addl Info on Proposed License Amend Re Increased Max Allowable Resistance Temp Detector Delay Time,Per 891219 Telcon & Advises That Util Request to Increase Plant Resistance Temp Detector Response Time Remain Unchanged
ML17223A5831990-03-19019 March 1990
Forwards Response to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implications of Control Sys in LWR Nuclear Power Plants,' Per 10CFR50.54(f)
ML17347B6191990-03-13013 March 1990
Provides Listing of Property Insurance Programs
ML17223A5531990-03-0909 March 1990
Submits Results of Investigation of Error Detected in Dose Assessment During 900124 NRC Evaluated Exercise at Plant. Operator Error Caused Keyboard Hangup Requiring Computer Restart
ML17223A5451990-03-0808 March 1990
Forwards Revised Tech Specs Re Steam Generator Tube Repairs, Per 890602 Telcon & Subsequent Discussions W/Nrc
ML17308A4871990-03-0707 March 1990
Forwards Response to Eight Audit Questions & Licensing Bases Criteria to Resolve Station Blackout Issue.Util Currently Has Procedures to Mitigate Effects of Hurricanes & Tornados Which Meet or Exceed NUMARC 87-00 Guidelines
ML17347B5881990-03-0101 March 1990
Responds to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey. Info Covers Time Spent by Key Power Plant Managers in Responding to Operational Insps & Audits
ML17347B6031990-02-27027 February 1990
Requests Approval to Use Code Case N-468 at Plants
ML17223A5321990-02-26026 February 1990
Forwards CEN-396 (L)-NP, Verification of Acceptability of 1-Pin Burnup Limit of 60 Mwd/Kg for St Lucie Unit 2.
ML20012A0011990-02-26026 February 1990
Notifies That Followup Actions Completed on Schedule & Incorporated Into Rev 25 to Plant Physical Security Plan,Per NRC 890605 Request
ML17223A5411990-02-26026 February 1990
Provides Addl Info Re Proposed License Amends Re Moderator Temp Coefficient Surveillance Requirements,Per 891026 & 900109 Telcons
IR 05000335/19890241990-02-22022 February 1990
Responds to Violations Noted in Insp Repts 50-335/89-24 & 50-389/89-24.Corrective Actions:Applicable Procedures Changed to Clarify Which Spaces & Blocks Required to Be Completed on Plant Work Order & QC Supervisor Counseled
1990-09-13
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Text
ACCELERATED DIRIBUTION DEMON~TION SYSIEM j 'e y
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9003260394 DOC.DATE: 90/03/19 NOTARIZED: YES DOCKET FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power 6 Light Co. 05000335 50-389 St. Lucie Plant, Unit 2, Florida Power 6 Light Co. 05000389 AUTH. NAME AUTHOR AFFILIATION GOLDBERG,J.H. Florida Power 6 Light Co.
RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Responds to Generic Ltr 89-19, "Request for Action re Resolution of USI A;47."
CODE: AOOID k'ISTRIBUTION COPIES RECEIVED:LTR Q ENCL ( SIZE:
TITLE: OR Submittal: General Distribution NOTES:
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 NORRIS,J 5 5 INTERNAL: ACRS 6 6 NRR/DET/ECMB 9H 1 1 NRR/DOEA/OTSB11 1 1 NRR/DST 8E2 1 1 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 OC 1 0 OGC/HDS2 1 0 R FILE 1 1 RES/DSIR/EIB 1 1 EXTERNAL: LPDR 1 1 NRC PDR NSIC 1 1 NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP, US TO REDUCE WASTEI CONTACT THE.DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDt TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 25
0 P.O. Box14000, Juno Beach, FL 33408-0420 MARCH 4 9 ~99O L-90-106 10 CFR 50.54(f)
U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Gentlemen:
Re: St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Generic Letter 89-19 Request. for Action Related to Resolution of Unresolved Safety Issue A-47 "Safety Implications of Control Systems in LWR Nuclear Power Plants" Pursuant to 10 CFR 50.54 f As a result of the technical resolution of USI A-47, "Safety Implications of Control Systems in LWR Nuclear Power Plants," the
'NRC concluded that protection should be provided for certain control system failures, and, for certain plants, that selected emergency procedures should be modified to assure that plant transients resulting from control system failures do not compromise public safety.
Generic Letter 89-19, issued September 20, 1989, provided recommendations for control system design and procedural modifications for resolution of USI A-47. Specifically, Generic Letter 89-19 recommended that all Combustion Engineering plant designs provide automatic steam generator overfill protection to mitigate main feedwater (MFW) overfeed events, and that plant procedures and technical specifications include provisions to periodically verify the operability of the MFW overfill protection and ensure that the automatic overfill protection is operable during reactor power operation. Additionally, it that all utilities that have plants designed with high pressure was recommended injection pump discharge pressures less than or equal to 1275 psi reassess their emergency procedures and operator training programs and modify them, as needed, to ensure that the operators can handle the full spectrum of possible small-break loss of coolant accident scenarios.
Florida Power 6-Light Company's response to the recommendations in Generic Letter 89-19 for St. Lucie Units 1 and 2 is attached.
9003260394 9003i9 PDR ADOCK 05000335 P PD.C ,oI an FPL Group company
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U. S. Nuclear Regulatory Commission L-90-106 Page two Should there be any questions regarding the attached information, please contact us.
V ry truly yours, J. H. Goldberg Executive Vice President JHG/MSD/gp Attachment cc: Stewart D. Ebneter, Regional Administrator, Region Senior Resident Inspector, USNRC, St. Lucie Plant II, USNRC
STATE OF FLORIDA )
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COUNTY OF PALM BEACH )
J. H. Goldber being first duly sworn, deposes and says:
That he is Executive Vice President, of Florida Power & Light Company, the Licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information and belief, and that he is authorized to execute the document on behalf of said Licensee.
J. H. Goldbe Subscribed and sworn to before me this
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'NOTARY PUBL C, in and for th ounty of
<,. Palm Peach, State of Florida Not%, pobrir, State of HorMa on Exp res lone 1 1993 Ny t:omm~ s Iasaraaca Inc.
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ATTACHMENT Response to Generic Letter 89-19, Unresolved Safety Issue Im lementation of Control S stems in LWR Power Plants~~ A-47'~Safet Back round The purpose of Generic Letter 89-19 is to ensure that all power plants have reactor vessel or steam generator overfill protection.
Additionally, it was recommended that all utilities that have plants designed with high pressure injection pump discharge pressures less than or equal to 1275 psi reassess their emergency procedures and operator training programs and modify them, as needed, to ensure that the operators can handle the full spectrum of possible small-break loss of coolant accident scenarios. to Generic Letter 89-19 discusses the recommendations for Combustion Engineering (CE) Nuclear Steam Supply System pressurized water reactors like St. Lucie. The recommendations are:
(1) All CE plants provide automatic steam generator overfill protection.
(2) All CE plants provide plant procedures and technical specifications for periodic surveillance of the overfill protection.
(3) CE plants with high pressure injection pump discharge pressures less than 1275 psi reassess their emergency procedures and operator training to ensure safe shutdown during any postulated small break loss of coolant accident.
In Enclosure 2 the NRC concludes, "CE-designed plants do not provide automatic steam generator overfill protection that terminates MFW flow." FPL is of the position that, with respect to St. Lucie Units 1 and 2, this is incorrect and both units have steam generator overfill protection systems to satisfactorily resolve USI A-47. The recommendations outlined in Section 4 of are discussed below for St. Lucie Units 1 and 2.
Recommendation 4a It is recommended automatic, steam that all Combustion Engineering plants provide generator overfill protection to mitigate main feedwater (MFW) overfeed events. The design for the overfill-protection system should be sufficiently separate from the MFW control system to ensure that the MFW pump will trip on a steam generator high-water-level signal when required, even if a loss of power, a loss of ventilation, or a fire in the control portion of the MFW control system should occur. Common failure modes that could disable overfill protection and the feedwater control system, but, would still result in a feedwater pump trip, are considered acceptable failure modes.
FPL Res onse:
Steam Generator Overfill S stem Descri tion St. Lucie Units 1 and 2 use a 2-of-4 coincidence logic for steam generator overfill protection from four Reactor Protective System (RPS) level loops per steam generator. The loops provide high and high-,high level isolation signals. These signals result in feedwater regulating valve closu're on high steam generator level and feedwater pump trip and main turbine trip on high-high steam generator level. The initiating signals are used in non-safety related overfill protection circuits. The level loops which initiate steam generator isolation (feedwater regulating valve closure, feedwater pump trip and main turbine trip) are different from the level channels used for normal feedwater regulating valve control.
The high level isolation signal from the RPS level indicating controllers provides a feedwater control override signal, which results in closure of the feedwater regulating valve for the respective train. Although the initiating signal is generated from the RPS level loops, the main feedwater regulating valve isolation function is integral to the feedwater control system. The high-high level isolation function is accomplished independently of the feedwater system via a separate circuit. Upon receipt of a high-high level isolation signal, with a 2-of-4 coincidence from either steam generator, a trip relay is energized. This relay trips both feedwater pumps and energizes a second trip relay and an .auto stop solenoid. The second trip relay provides annunciation. The auto stop solenoid provides for a main turbine trip by closing the turbine stop/control valves.
The initiating logic for steam generator overfill protection is from either of two sets of four safety-grade, independent RPS transmitter loops. The signals initiate from level indicators which are mounted on the Reactor Turbine Generator Board. The St.
Lucie Unit 2 signals pass through isolation cabinets; the Unit 1 signals do not. The signals then pass through common (separate circuits) "normal/bypass" control switches. The high level signal then ties into the feedwater regulating system modules. The high-high level signal ties into the two trip relays and turbine stop solenoid.
Review of the power supplies for the feedwater control system and the steam generator overfill protection circuit shows they are from different sources. The normal power supplies for a single feedwater control train and the overfill protection system are ultimately fed from a common 480 volt switchgear (Unit 1) or 4160 volt bus (Unit 2); however, the feedwater control system is supplied 120 volt AC power and the steam generator overfill protection system is provided 125 volt DC. The steam generator overfill protection circuit is provided back-up emergency power from the 2B(1A) station battery in the event normal power is lost.
Each train of the feedwater control system is also provided with an automatic transfer to a back-up power source from the 120 volt vital AC cabinets, to ensure further reliability. This source is
V
~ t ultimately powered from the 2AB(lAB) swing bus, which can be manually aligned to either train "A" or "B" 480 volt AC switchgear.
Emergency back-up power is also provided via the 2D(1D) station battery and associated inverter.
Common Failure Modes Discussion Various potential common mode failures between the feedwater control system and the steam generator overfill protection circuits were evaluated. The limiting common mode failures evaluated were for a loss of power, a common fire, and failure of the shared override/manual control switch.
Power Loss The normal power supplies for the feedwater control and steam generator overfill protection circuits are not totally independent (as discussed above); however, the back-up supplies are sufficiently independent to ensure operability of one or both systems in the event a common normal bus is lost. The steam generator overfill protection circuitry is provided back-up 125 volt DC power from safety-related station battery 2B(1A), such that the feedwater pump trip should still occur if a high-high steam generator level condition occurs coincident with a loss of normal power. Similarly, the feedwater control system is provided with an automatic transfer to the 120 volt AC vital bus (fed off the "AB" 480 volt Motor Control Center) in the event normal power is lost. This feedwater control power source is further backed-up by the 2D(1D) battery. Furthermore, the design of the feedwater control system requires the feedwater regulating valve to fail closed on a loss of power, such that even in the remote event (multiple failure) the common bus and the back-up power supplies are lost, feedwater flow will still be isolated for the affected train.
Fire In the event of a fire in the control room, the feedwater control and overfill protection systems are not sufficiently separate to ensure the operability of at least one system. The two systems have circuitry, which share common "Norm/bypass" handswitches in the control room, common cable tray routing and are located within adjacent control room panels. Given a fire of a large magnitude, which renders the control room uninhabitable, a manual reactor/turbine trip would be initiated prior to leaving the control room and safe shutdown would be accomplished from the remote shutdown panel. Main feedwater flow would be isolated per operating procedures and steam generator level is controlled via auxiliary feedwater.
Shared Control Switch The final common mode failure postulated was the failure of a common "normal/bypass" handswitch due to inadvertent mispositioning, contact failure "open," or short to ground.
Inadvertent mispositioning of the handswitches is not credible, as annunciation is provided to warn mispositioned. Furthermore, if if any of the four switches is a single handswitch is mispositioned or its contacts fail open, the high and high-high overfill protection are still operable. A 2-of-3 coincidence will now be required to initiate a high/high-high steam generator isolation signal. In the event a single switch shorts to ground, both the feedwater control and steam generator overfill protection circuits lose power and the feedwater regulating valves fail close.
Based on the above, FPL concludes that the steam generator overfill protection system currently installed in St. Lucie Units 1 and 2 provides adequate assurance that a steam generator overfill event will be prevented.
Recommendation 4b It is recommended that plant specifications for all Combustion procedures and technical Engineering plants include provisions to verify periodically the operability of overfill protection and ensure that automatic main feedwater overfill protection is operable during reactor power operation. The instrumentation should be demonstrated to be operable by the performance of a channel check, channel functional testing, and channel calibration, including setpoint verification and by identifying the LCOs. These technical specifications should be commensurate with existing plant technical specification requirements for channels that initiate protection actions.
FPL Res onse:
St. Lucie Plant procedures include provisions to periodically verify. steam generator overfill protection and ensure that automatic main feedwater overfill protection is operable during reactor power operation. Testing of the steam generator high and high-high level circuitry is completed on an 18 month refueling basis by functionally testing the circuitry using the 2-of-4 logic to close the feedwater regulating valve, trip the feedwater pump and trip the turbine.
The NRC staff is currently reviewing the Nuclear Steam Supply System (NSSS) vendor specific revised standard technical specifications developed using the criteria of the NRC interim policy statement on technical specification improvements (Federal Register p. 3788/Vol. 52 No. 25/Friday, February 6, 1987).
Recommendations for specific Limiting Conditions for Operation (LCO) pertaining to steam generator overfill protection should be addressed as part of the review of the improved NSSS vendor technical specification submittals. In addition, FPL reviewed the criteria of the Commission's interim policy statement on technical specification improvements pertaining to the identification of LCOs to be included in the technical specifications. It was concluded through use of the criteria in the interim policy statement that additional LCOs for the steam generator overfill protection system
are not appropriate for inclusion in the St. Lucie Plant Technical Specifications. FPL believes that, the requirements for testing the steam generator overfill protection system can be controlled through the plant procedures.
Recommendation 4c Reassess emergency procedures and operator training programs and modify them, as needed, to ensure that the operators can handle the full spectrum of possible small-break loss-of-coolant-accident (SBLOCA) scenarios. This may include the need to depressurize the primary system via the atmospheric dump valves or the turbine bypass valves and cool down the plant during some SBLOCA. The reassessment should ensure that a single failure would not negate the operability of the valves needed to achieve safe shutdown. The procedures should clearly describe any actions the operator is required to perform in the event a loss of instrument air or electric power prevents remote operation of the valves. The use of the pressurizer PORVs to depressurize the plant during a SBLOCA, if needed, and the means to ensure that the RT>>, (reference temperature, nil ductility transition) limits are not compromised, should also be clearly described.
FPL Res onse:
St. Lucie Plant Emergency Operating Procedure, EOP-3, Loss of Coolant Accident (LOCA), provides operator instructions and contingency actions for the full spectrum of LOCA scenarios.
Guidance is provided within EOP-3 for cooldown and depressurization of the plant. This procedure addresses cooldown and depressurization by means of the Steam Bypass Control System (SBCS). When the SBCS is unavailable, guidance contained in the contingency actions of EOP-3 is given to use the Atmospheric Dump Valves (ADV) to cooldown and depressurize the plant. EOP-3 is written to allow cooldown and depressurization of the plant with either a Loss of Offsite Power (LOOP) or the loss of a single emergency electrical train. Loss of instrument air or an electrical casualty may require manual operation of the ADVs.
Manual operation of the ADVs is possible and is included in operator training classes. 'he inability to cooldown and depressurize the plant using the SBCS or the ADVs constitutes the loss of a Safety Function. Upon the loss of Safety Function, procedural guidance and training on the use of Emergency Operating Procedures directs the operator to EOP-15, Functional Recovery.
Cooldown and depressurization of the plant would then be affected through use of EOP-15, RCS and Core Heat Removal Success Path 4, once through cooling using the Pilot Operated Relief Valves (PORV).
The RT>>I criteria are met through compliance with Figure One (Pressure/Temperature Curves), which is provided in all Emergency Operating Procedures, and Pressurized Thermal Shock (PTS) Criteria which are provided in the appropriate Emergency Operating Procedures involving complicated reactor trips.
Additionally, FPL is participating in a Combustion Engineering Owner's Group task of conducting an assessment of the potential for inadequate core cooling during a small break LOCA.
