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| {{#Wiki_filter:I/ix/'Qllj~gslÃl/I~Jlili$1illliiI!1<rlivi<piivlROCHESTERGASANDELECTRICCORPORATIONrl~~Oos~Oo89EASTAVENUE,ROCHESTER,N.Y.L4649LEONO.iVHITE,JR.VICCPRCSIDCNTTCLCPHOVCiPCiCOOCria546-2700April30,1980Mr.BoyceH.Grier,DirectorU.S.NuclearRegulatoryCommissionOfficeofInspectionandEnforcementRegionI651ParkAvenueKingofPrussia,Pennsylvania19406 | | {{#Wiki_filter:I/ix/'Qllj~gslÃl/I |
| | ~Jlili$1illliiI!1<rlivi<piivl ROCHESTER GASANDELECTRICCORPORATION rl~~Oos~Oo89EASTAVENUE,ROCHESTER, N.Y.L4649LEONO.iVHITE,JR.VICCPRCSIDCNT TCLCPHOVC iPCiCOOCria546-2700April30,1980Mr.BoyceH.Grier,DirectorU.S.NuclearRegulatory Commission OfficeofInspection andEnforcement RegionI651ParkAvenueKingofPrussia,Pennsylvania 19406 |
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| ==DearMr.Grier.:== | | ==DearMr.Grier.:== |
| EnclosedisourresponsetoIEBulletinNo.80-04.ThisbulletinwasreceivedonFebruary8,1980.Sincerely,XC:OfficeofInspectionandEnforcementDivisionofReactorOperationsInspectionU.S.NuclearRegulatoryCommissionWashington,D.C.20555 | | EnclosedisourresponsetoIEBulletinNo.80-04.ThisbulletinwasreceivedonFebruary8,1980.Sincerely, XC:OfficeofInspection andEnforcement DivisionofReactorOperations Inspection U.S.NuclearRegulatory Commission Washington, D.C.20555 |
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| ResonsetoIEBulletin80-04Request1.IResponse1.Reviewthecontainmentpressureresponseanalysistodetermineifthepotentialforcontainmentoverpressureforamainsteamlinebreakinsidecontainmentincludedtheimpactofrunoutflowfromtheauxiliaryfeedwatersystemandtheimpactofotherenergysources,suchascontinuationoffeedwaterorcondensateflow.Inyourreview,consider.yourabilitytodetectandisolatethedamagedsteamgeneratorfromthesesourcesandtheabilityofthepumpstoremainoperableafterextendedoperationat,runoutflow.AlthoughtheGinnapost-steamlinebreakcontainmentpressureanalysisintheFSARdidnotincludetheeffectsofauxiliaryfeedwaterflowtotheaffectedsteamgenerator,itisimportanttorecognizethattheevaluationalsodidnotincludethebenefitsofpassiveandactiveheatsinksinsidecontainment.Continuedfeedwater/condensateadditiontothesteamgeneratorwillnotoccur,since,thesafetyinjectionsignal(generatedbyavarietyofprocessparameters,includinghighsteamlineflow,highcontainmentpressure,andlowpressurizerpressure)willclosethefeedwatercontrolvalvesandstopthefeedwaterpumps.Theadditionofmaximumauxiliaryfeedwaterflowtothebrokensteamgeneratorwilleventuallyrequireoperatoractionto1)realignflowtotheintactgenerator,2)terminateauxiliaryfeedwaterflowtothebrokengenerators.Positiveinformationisavailabletotheoperatortodeterminewhichistheaffectedsteamgenerator.Steamgeneratorlevelinstrumentationislocatedinsidecontainmentandsteamgeneratorpressureislocatedoutsidecontain-mentwhereit.wouldnotbeaffectedbytheaccidentenvironment,insidecontainment.Itisexpectedthat,throughpropertrainingandbyuseoftheemergencyprocedures,theoperatorswillbecapableofquicklyrecognizingthesteamlinebreak,andwillperformtheproperoperations.Thereissubstantialtimeavailablefortheoperatortoperformthetwosafetyfunctionsnotedabove.TheSEPSafeShutdownreviewconcludedfollowingtheirsitevisitinJune1978thatonesteamgeneratorwouldnotboildryforoverthirtyminutes.Thusthereissubstantialtimetoalignflowtotheintactsteamgenerator.Theterminationofauxiliaryfeedwaterflowtotheaffectedsteamgenerator,underthepessimisticcircumstances,wouldrequiremorerapidaction(butstilleasilywithinthecapabilityoftheoperators)tomaintaincontainment | | ResonsetoIEBulletin80-04Request1.IResponse1.Reviewthecontainment pressureresponseanalysistodetermine ifthepotential forcontainment overpressure foramainsteamlinebreakinsidecontainment includedtheimpactofrunoutflowfromtheauxiliary feedwater systemandtheimpactofotherenergysources,suchascontinuation offeedwater orcondensate flow.Inyourreview,consider. |
| | yourabilitytodetectandisolatethedamagedsteamgenerator fromthesesourcesandtheabilityofthepumpstoremainoperableafterextendedoperation at,runoutflow.AlthoughtheGinnapost-steamline breakcontainment pressureanalysisintheFSARdidnotincludetheeffectsofauxiliary feedwater flowtotheaffectedsteamgenerator, itisimportant torecognize thattheevaluation alsodidnotincludethebenefitsofpassiveandactiveheatsinksinsidecontainment. |
| | Continued feedwater/condensate additiontothesteamgenerator willnotoccur,since,thesafetyinjection signal(generated byavarietyofprocessparameters, including highsteamlineflow,highcontainment |
| | : pressure, andlowpressurizer pressure) willclosethefeedwater controlvalvesandstopthefeedwater pumps.Theadditionofmaximumauxiliary feedwater flowtothebrokensteamgenerator willeventually requireoperatoractionto1)realignflowtotheintactgenerator, 2)terminate auxiliary feedwater flowtothebrokengenerators. |
| | Positiveinformation isavailable totheoperatortodetermine whichistheaffectedsteamgenerator. |
| | Steamgenerator levelinstrumentation islocatedinsidecontainment andsteamgenerator pressureislocatedoutsidecontain-mentwhereit.wouldnotbeaffectedbytheaccidentenvironment, insidecontainment. |
| | Itisexpectedthat,throughpropertrainingandbyuseoftheemergency procedures, theoperators willbecapableofquicklyrecognizing thesteamlinebreak,andwillperformtheproperoperations. |
| | Thereissubstantial timeavailable fortheoperatortoperformthetwosafetyfunctions notedabove.TheSEPSafeShutdownreviewconcluded following theirsitevisitinJune1978thatonesteamgenerator wouldnotboildryforoverthirtyminutes.Thusthereissubstantial timetoalignflowtotheintactsteamgenerator. |
| | Thetermination ofauxiliary feedwater flowtotheaffectedsteamgenerator, underthepessimistic circumstances, wouldrequiremorerapidaction(butstilleasilywithinthecapability oftheoperators) tomaintaincontainment |
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| pressurebelowdesignpressure.TheanalysispresentedinAttachment1concludesthat,assumingminimumsafeguardsfor.containmentcooling,auxiliaryfeedwaterflowwouldhavetobeterminatedinabout26minutes.Withmaximumsafeguards,thistimewouldbeextendedtoabout44minutes.Thereisnoneedtoconsidertheoperationoftheauxiliaryfeedwaterpumpsatrunoutflow.Theturbine-drivenpumpsarecontrolledbya'overnor,andwillnot,exceedabout400gpm.ThemotordrivenpumpflowiscontrolledbytheAFWcontrolvalves,whichreceiveanautomaticthrottlesignalto200gpmfromtheirflowcontrollers.Apotentialsinglefailureoftheflowcontrollertocontrol.flowto200gpmisnotconsideredaworst-casesinglefailureintermsofnetenergyadditiontothecontainment,sincetheoperationofallcontainmentcoolingsafeguards(vs.theminimumsafeguardsassumedinthisevaluation)wouldresultinasubstantialincreaseinenergyremovalfromcontainment.Request2.Reviewyouranalysisofthereactivityincreasewhichresultsfromamainsteamlinebreakinsideor"outsidecontainment.Thisreviewshouldconsiderthereactorcooldownrateandthepotentialforthereactortoreturntopowerwiththemostreactivecontrolrodinthefullywithdrawnposition.Ifyourpreviousanalysisdidnotconsiderallpotentialwatersources(suchasthoselistedin1above)andifthereactivityincreaseisgreaterthanpreviousanalysisindicatedthereportofth'isreviewshouldinclude:aITheboundaryconditionsfortheanalysis,e.g.,theendoflifeshutdownmargin,themoderatortemperaturecoefficient,powerlevelandtheneteffectoftheassociatedsteamgeneratorwaterinventoryonthereactorsystemcooling,etc.,b.Themostrestrictivesingleactivefailureinthesafetyinjectionsystemandtheeffectofthatfailureondelayingthedeliveryofhighconcentrationboricacidsolutiontothereactorcoolantsystem,cTheeffectofextendedwatersupplytotheaffectedsteamgeneratoronthecorecriticalityandreturntopower, | | pressurebelowdesignpressure. |
| | Theanalysispresented inAttachment 1concludes that,assumingminimumsafeguards for.containment cooling,auxiliary feedwater flowwouldhavetobeterminated inabout26minutes.Withmaximumsafeguards, thistimewouldbeextendedtoabout44minutes.Thereisnoneedtoconsidertheoperation oftheauxiliary feedwater pumpsatrunoutflow.Theturbine-driven pumpsarecontrolled bya'overnor, andwillnot,exceedabout400gpm.Themotordrivenpumpflowiscontrolled bytheAFWcontrolvalves,whichreceiveanautomatic throttlesignalto200gpmfromtheirflowcontrollers. |
| | Apotential singlefailureoftheflowcontroller tocontrol.flowto200gpmisnotconsidered aworst-case singlefailureintermsofnetenergyadditiontothecontainment, sincetheoperation ofallcontainment coolingsafeguards (vs.theminimumsafeguards assumedinthisevaluation) wouldresultinasubstantial increaseinenergyremovalfromcontainment. |
| | Request2.Reviewyouranalysisofthereactivity increasewhichresultsfromamainsteamlinebreakinsideor"outsidecontainment. |
| | Thisreviewshouldconsiderthereactorcooldownrateandthepotential forthereactortoreturntopowerwiththemostreactivecontrolrodinthefullywithdrawn position. |
| | Ifyourpreviousanalysisdidnotconsiderallpotential watersources(suchasthoselistedin1above)andifthereactivity increaseisgreaterthanpreviousanalysisindicated thereportofth'isreviewshouldinclude:aITheboundaryconditions fortheanalysis, e.g.,theendoflifeshutdownmargin,themoderator temperature coefficient, powerlevelandtheneteffectoftheassociated steamgenerator waterinventory onthereactorsystemcooling,etc.,b.Themostrestrictive singleactivefailureinthesafetyinjection systemandtheeffectofthatfailureondelayingthedeliveryofhighconcentration boricacidsolutiontothereactorcoolantsystem,cTheeffectofextendedwatersupplytotheaffectedsteamgenerator onthecorecriticality andreturntopower, |
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| d.Thehotchannelfactorscorrespondingtothemostreactiverodinthefullywithdrawn'ositionattheendoflife,andtheMinimumDeparturefromNucleateBoilingRatio(MDNBR)valuesfortheanalyzedtransient.Response2.WestinghouseElectricCorporationperformedtheoriginalsteambreakanalysisforGinnaasreportedintheFSARandareanalysissubmittedtotheNRCinSeptember1975.Westinghousehasreviewedtheassumptionsmadeformainandauxiliaryfeedwaterflowastheyapplytolicensingbasissteamlinebreaktransients.Severaloftherelevantassumptionsusedinallcoretransientanalysesfollow,andarefurtherexplainedintheGinnaFSAR.1.Thereactorisassumedinitiallytobeathotshutdownconditions,attheminimumallowableshutdownmargin.2.FortheConditionIVbreaks,i.e.,double-endedruptureofamainsteampipe,fullmainfeedwaterisassumedfromthebeginningofthetransientataveryconservativecoldtemperature.3.Allauxiliaryfeedwaterpumpsareinitiallyassumedtobeoperating,inadditiontothemainfeedwater.Theflowisequivalenttotheratedflowofallpumpsatthesteamgeneratordesignpressure.4Feedwaterisassumedtocontinueatitsinitialflowrateuntilfeedwaterisolationiscomplete,approximately10secondsafterthebreakoccurs,whileauxiliaryfeedwaterisassumedtocontinueatitsinitialflow'rate.5.Mainfeedwaterflowiscompletelyterminatedfollowingfeedwaterisolation.Basedonthemannerinwhichtheanalysisisperformed~forWestinghouseplants,thecoretransientresultsareveryinsensitivetoauxiliaryfeedwaterflow.Thefirstminuteofthetransientisdominatedentirelybythesteamflowcontributiontoprimary-secondaryheattransfer,whichistheforcingfunctionforboththereactivityandthermal-hydraulictransientsinthecore.Theeffectofauxiliaryfeedwaterrunout(orfailureofrunoutprotectionwhereapplicable)isminimal.Greaterfeedwaterflowduringthelargesteamlinebreaksservestoreducesecondarypressures,acceleratingtheautomaticsafeguardsactions,i.e.steamlineisolation,feedwaterisolationandsafetyinjection.Theassumptionsdescribedaboveare | | d.Thehotchannelfactorscorresponding tothemostreactiverodinthefullywithdrawn'osition attheendoflife,andtheMinimumDeparture fromNucleateBoilingRatio(MDNBR)valuesfortheanalyzedtransient. |
| | Response2.Westinghouse ElectricCorporation performed theoriginalsteambreakanalysisforGinnaasreportedintheFSARandareanalysis submitted totheNRCinSeptember 1975.Westinghouse hasreviewedtheassumptions madeformainandauxiliary feedwater flowastheyapplytolicensing basissteamlinebreaktransients. |
| | Severaloftherelevantassumptions usedinallcoretransient analysesfollow,andarefurtherexplained intheGinnaFSAR.1.Thereactorisassumedinitially tobeathotshutdownconditions, attheminimumallowable shutdownmargin.2.FortheCondition IVbreaks,i.e.,double-ended ruptureofamainsteampipe,fullmainfeedwater isassumedfromthebeginning ofthetransient ataveryconservative coldtemperature. |
| | 3.Allauxiliary feedwater pumpsareinitially assumedtobeoperating, inadditiontothemainfeedwater. |
| | Theflowisequivalent totheratedflowofallpumpsatthesteamgenerator designpressure. |
| | 4Feedwater isassumedtocontinueatitsinitialflowrateuntilfeedwater isolation iscomplete, approximately 10secondsafterthebreakoccurs,whileauxiliary feedwater isassumedtocontinueatitsinitialflow'rate.5.Mainfeedwater flowiscompletely terminated following feedwater isolation. |
| | Basedonthemannerinwhichtheanalysisisperformed |
| | ~forWestinghouse plants,thecoretransient resultsareveryinsensitive toauxiliary feedwater flow.Thefirstminuteofthetransient isdominated entirelybythesteamflowcontribution toprimary-secondary heattransfer, whichistheforcingfunctionforboththereactivity andthermal-hydraulic transients inthecore.Theeffectofauxiliary feedwater runout(orfailureofrunoutprotection whereapplicable) isminimal.Greaterfeedwater flowduringthelargesteamline breaksservestoreducesecondary pressures, accelerating theautomatic safeguards actions,i.e.steamline isolation, feedwater isolation andsafetyinjection. |
| | Theassumptions described aboveare |
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| thereforeappropriateandconservativefortheshort-termaspectofthesteamlinebreaktransient.Theauxiliaryfeedwaterflowbecomesa-dominantfactorindeterminingthedurationand,magnitudeofthesteamflowtransient'uringlaterstagesinthetransient.However,thelimitingportionofthetransientoccursduringthefirstminute,bothduetohighersteamflowsinherentlypresentearlyinthetransientandduetotheintroductionofborontothecoreviathesafetyinjectionsystem.Inconclusion,Westinghousehasevaluatedtheeffectofrunoutauxiliaryfeedwaterflowsinthecoretransientforsteamlinebreak,andbasedonthisevaluation,hasdeterminedthattheassumptionspresentlymadeareappropriateforuseasalicensingbasis.TheconcernsoutlinedintheintroductiontoIEBulletin80-04relativeto,1)limitingcoreconditionsoccurringduringportionsofthetransientwhereauxiliaryfeedwaterflowisarelevantcontributortoplantcooldown;and2)incompleteisolationofmainfeedwaterflow,arenotrepresentativeoftheWestinghouseNSSSdesignsandassociatedBalanceofPlantrequirements.ThemostlimitingsteamlinebreakdeterminedbyWestinghousewasanalyzedbyExxonNuclearCo.,Inc.andpresentedinXN-NF-77-40Supplement1,"PlantTransientAnalysisfortheR.E.GinnaUnit1NuclearPowerPlant,"March1980.Thistransientoccursathotzeropowerwithoutsidepoweravailableandthebreakoccurringattheexitofthesteamgenerator.TheExxonanalysisdoesnotspecificallyaccountforauxiliaryfeedwater.However,theSteamGeneratorheat.transfermodel,usingconstantheattransfercoefficients,co'ntinuestocalculateheattransferfromtheprimarytothesecondarysideafterthebrokensteamgeneratorhasbeenestimatedtobeempty.Ifauxiliaryflowwasspecificallyaccountedfor,itseffectwouldbenegligibleduringtheinitialportionofthetransientandwouldhaveminimaleffectduringlaterportionsofthetransientsincebythetimethebrokensteamgeneratorempties,thetotalsystemreactivityisnegativeandcorepowerisdecreasing.Theadditionalreactivityadditionassociatedwiththeslightcooldownduetorunoutflowismorethannegatedbytheboronreactivityinsertedbysafetyinjection.Therefore,theseverityofthetransientisnotincreased.
| | therefore appropriate andconservative fortheshort-term aspectofthesteamline breaktransient. |
| | Theauxiliary feedwater flowbecomesa-dominant factorindetermining thedurationand,magnitude ofthesteamflowtransient'uring laterstagesinthetransient. |
| | However,thelimitingportionofthetransient occursduringthefirstminute,bothduetohighersteamflowsinherently presentearlyinthetransient andduetotheintroduction ofborontothecoreviathesafetyinjection system.Inconclusion, Westinghouse hasevaluated theeffectofrunoutauxiliary feedwater flowsinthecoretransient forsteamline break,andbasedonthisevaluation, hasdetermined thattheassumptions presently madeareappropriate foruseasalicensing basis.Theconcernsoutlinedintheintroduction toIEBulletin80-04relativeto,1)limitingcoreconditions occurring duringportionsofthetransient whereauxiliary feedwater flowisarelevantcontributor toplantcooldown; and2)incomplete isolation ofmainfeedwater flow,arenotrepresentative oftheWestinghouse NSSSdesignsandassociated BalanceofPlantrequirements. |
| | Themostlimitingsteamlinebreakdetermined byWestinghouse wasanalyzedbyExxonNuclearCo.,Inc.andpresented inXN-NF-77-40 Supplement 1,"PlantTransient AnalysisfortheR.E.GinnaUnit1NuclearPowerPlant,"March1980.Thistransient occursathotzeropowerwithoutsidepoweravailable andthebreakoccurring attheexitofthesteamgenerator. |
| | TheExxonanalysisdoesnotspecifically accountforauxiliary feedwater. |
| | However,theSteamGenerator heat.transfermodel,usingconstantheattransfercoefficients, co'ntinues tocalculate heattransferfromtheprimarytothesecondary sideafterthebrokensteamgenerator hasbeenestimated tobeempty.Ifauxiliary flowwasspecifically accounted for,itseffectwouldbenegligible duringtheinitialportionofthetransient andwouldhaveminimaleffectduringlaterportionsofthetransient sincebythetimethebrokensteamgenerator empties,thetotalsystemreactivity isnegativeandcorepowerisdecreasing. |
| | Theadditional reactivity additionassociated withtheslightcooldownduetorunoutflowismorethannegatedbytheboronreactivity insertedbysafetyinjection. |
| | Therefore, theseverityofthetransient isnotincreased. |
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| Request3.Ifthepotentialforcontainmentoverpressureexists:orthereactor-return-to-powerresponseworsens,provideaproposedcorrectiveaction.Iftheunitisoperating,provideadescriptionofanyinterimactionthatwillbetakenuntiltheproposedcorrectiveactioniscompleted.Response3.Sinceneitherthepotentialforcontainmentover-pressurizationnorthereactor-return-to-powerresponseworsensnocorrectiveactionisrequired.Request4.Within90daysofthedateofthisBulletin,completethereviewandevaluationrequiredbythisBulletinandprovideawrittenresponsedescribingyourreviewsandactionstakeninresponsetoeachitem.Response4.Thisattachmentprovidestherequired90dayresponsetoIEBulletinNo.80-04. | | Request3.Ifthepotential forcontainment overpressure exists:orthereactor-return-to-power responseworsens,provideaproposedcorrective action.Iftheunitisoperating, provideadescription ofanyinterimactionthatwillbetakenuntiltheproposedcorrective actioniscompleted. |
| | Response3.Sinceneitherthepotential forcontainment over-pressurization northereactor-return-to-power responseworsensnocorrective actionisrequired. |
| | Request4.Within90daysofthedateofthisBulletin, completethereviewandevaluation requiredbythisBulletinandprovideawrittenresponsedescribing yourreviewsandactionstakeninresponsetoeachitem.Response4.Thisattachment providestherequired90dayresponsetoIEBulletinNo.80-04. |
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| Attachment1'-ContainmentEnerBalanceThepurposeofthisevaluationistodeterminethelengthoftimeavailabletotheoperatortoterminate'AFWflowtothebrokensteamgeneratorfollowingasteamlinebreakinsidecontainment,.priortocontainmentpressureexceeding60psig.InitialConditionsandAssumtionsa.,ZOCAenergyreleasetocontainment(includescreditforcontainmentheatsinks):191.3x10BTU(takenfromTable'14.3.4-2ofFSAR).Resultsinpeakcontainmentpressureof53psig.b.Additionalenergytoreach60psig:16x10BTU(takenfromFig.14.3.4-3ofFSAR).C.Minimumsafeguardsheatremovalcapability(1spraypumpand2fancoolers)=55x10BTU/sec=3.3x106BTU/min.(takenfromFig.14.3.4-9of'theFSAR).d.Energyinputfrom600gpmtothebrokensteamIgenerator=600galx1180BTUmanibmx62.6~lbx1ft=5.91x10B'fU6ft7.68galmz.ne.Energyreleasedtocontainmentfromtheinitialsteamlinebreakaccidentblowdown=140x10BTU(takenfromFig.14.2.5-10oftheFSAR).
| | Attachment 1'-Containment EnerBalanceThepurposeofthisevaluation istodetermine thelengthoftimeavailable totheoperatortoterminate'AFW flowtothebrokensteamgenerator following asteamlinebreakinsidecontainment,.prior tocontainment pressureexceeding 60psig.InitialConditions andAssumtionsa.,ZOCAenergyreleasetocontainment (includes creditforcontainment heatsinks):191.3x10BTU(takenfromTable'14.3.4-2 ofFSAR).Resultsinpeakcontainment pressureof53psig.b.Additional energytoreach60psig:16x10BTU(takenfromFig.14.3.4-3ofFSAR).C.Minimumsafeguards heatremovalcapability (1spraypumpand2fancoolers)=55x10BTU/sec=3.3x106BTU/min.(takenfromFig.14.3.4-9of'theFSAR).d.Energyinputfrom600gpmtothebrokensteamIgenerator |
| | =600galx1180BTUmanibmx62.6~lbx1ft=5.91x10B'fU6ft7.68galmz.ne.Energyreleasedtocontainment fromtheinitialsteamlinebreakaccidentblowdown=140x10BTU(takenfromFig.14.2.5-10 oftheFSAR). |
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| Calculationsl.Additionalenergywhichcouldbeabsorbedbycontainment,takingcreditforpassiveandminimumactivecontainment,heatsinksfollowingasteamlinebreak.(froma+b-2.eabove).(191.3+16)x10BTU-140x10BTU=67.3NBTU66Netenergyadditiontocontainmentfollowinginitialsteamlinebreakblowdown(fromd-cabove):(5.91x10BTU-3.3x10BTU)=2.61x10BTU666manmz.nmz.n3.OperatorActionTime=67.3NBTU2.61NBTU/man f"lg}}
| | Calculations l.Additional energywhichcouldbeabsorbedbycontainment, takingcreditforpassiveandminimumactivecontainment, heatsinksfollowing asteamlinebreak.(froma+b-2.eabove).(191.3+16)x10BTU-140x10BTU=67.3NBTU66Netenergyadditiontocontainment following initialsteamlinebreakblowdown(fromd-cabove):(5.91x10BTU-3.3x10BTU)=2.61x10BTU666manmz.nmz.n3.OperatorActionTime=67.3NBTU2.61NBTU/man f"lg}} |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999
Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated
IR 05000244/19990081999-10-14014 October 1999
Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy
ML17265A7651999-10-0808 October 1999
Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request
ML17265A7641999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel
ML17265A7631999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000
ML20212J3801999-09-30030 September 1999
Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21
ML20212J3561999-09-30030 September 1999
Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21
IR 05000244/19992011999-09-24024 September 1999
Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld
IR 05000244/19990051999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7571999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7461999-08-31031 August 1999
Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726
ML17265A7411999-08-26026 August 1999
Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded
ML17265A7401999-08-26026 August 1999
Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components
ML17265A7451999-08-23023 August 1999
Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d)
ML17250B3021999-08-23023 August 1999
Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806
ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
1999-09-30
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999
Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated
ML17265A7651999-10-0808 October 1999
Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request
ML17265A7641999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel
ML17265A7631999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000
ML20212J3801999-09-30030 September 1999
Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21
ML20212J3561999-09-30030 September 1999
Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21
IR 05000244/19990051999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7571999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7461999-08-31031 August 1999
Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726
ML17265A7411999-08-26026 August 1999
Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded
ML17265A7401999-08-26026 August 1999
Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components
ML17265A7451999-08-23023 August 1999
Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d)
ML17250B3021999-08-23023 August 1999
Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806
ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
ML17265A5091999-01-11011 January 1999
Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 41 Encl
ML17265A5101999-01-11011 January 1999
Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl
1999-09-30
[Table view]
Category:UTILITY TO NRC
MONTHYEARML20059L4481990-09-18018 September 1990
Forwards Responses to Rer 900709-13 Team Visit Findings, Per .Responses Withheld (Ref 10CFR73.21)
ML17309A4491990-09-13013 September 1990
Forwards Slides Presented by Facility & Westinghouse to NRC in 900724 Meeting.Encl Withheld
ML17262A1331990-09-11011 September 1990
Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion
ML17261B1511990-08-29029 August 1990
Forwards Semiannual Radioactive Effluent Release Rept Jan- June 1990 & Rev 3 to Process Control Program for Ginna Station, Per Tech Specs 6.9.1.4 & 6.16,respectively
ML17261B1481990-08-28028 August 1990
Lists Understanding of Issues Util Planning to Address Re Containment Integrity,Per 900718 Telcon.Any Concerns or Action Items Different from Listed Submittal Should Be Provided to Util Prior to NRC 900905 & 06 Visit to Plant
ML17261B1501990-08-20020 August 1990
Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented
IR 05000244/19880261990-08-20020 August 1990
Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented
ML17261B1371990-08-17017 August 1990
Forwards Rev 0 to, Inservice Insp Rept for Third Interval (1990-1999),First Period,First Outage (1990) at Re Ginna Nuclear Power Plant.
ML17309A4481990-08-16016 August 1990
Responds to NRC 900717 Ltr Re Violations Noted in Insp Rept 50-244/90-80.Corrective Actions:Westinghouse Drawing E-2508 Approved & Issued to Central Records
ML17261B1351990-08-15015 August 1990
Provides Info & Assessment on Integrity of Connection of Containment Bldg to Foundation,Per 900808 Telcon.Util Believes That Existing Condition Not Safety Concern & No Reason Exists to Suspect Joint Will Not Perform Function
ML17261B1361990-08-14014 August 1990
Responds to Commitment Tracking Concerns Noted in Insp Rept 50-244/90-09 & Planned Corrective Actions.Util Confirms Commitments Dates for Implementation of Effective Shelf Life Program & Comprehensive Preventive Maint Program for Parts
ML17261B1331990-08-13013 August 1990
Recommits to Performing Enhanced primary-to-secondary Leak Rate Monitoring,Per NRC Bulletin 88-002
ML17261B1261990-07-30030 July 1990
Clarifies Commitment Made in 900316 RO Re Restoring Inoperable Fire Damper I-411-21-P.Util Plans to Design Removable Track Which Will Allow Charcoal Drawer to Be Manipulated.Definitive Schedule Will Be Provided in 60 Days
ML20055H9291990-07-23023 July 1990
Forwards Revised Page 6:4 of Plant Security Plan.Page Withheld Per 10CFR73.21
ML17261B1091990-07-20020 July 1990
Advises That Structural Evaluations of Containment Sys Being Performed in Response to Containment Integrity Insp
ML17309A4471990-07-17017 July 1990
Forwards Decommissioning Rept, for Plant,Per 10CFR50.33(k) & 50.75(b)
ML17261B0881990-07-13013 July 1990
Provides Update to Util 860616 Ltr Re Implementation of NUREG-0737,Item 6.2,Suppl 1, Emergency Response Capability.
ML17261B0921990-07-11011 July 1990
Responds to 900709 Request for Addl Info Re Containment Integrity Insp.Util Will Provide Preliminary Results by 900716 Re Where Groundwater Entering Annular Access Area. Meeting Proposed for Wk of 900723
ML20044B0481990-07-10010 July 1990
Discusses Review of Station Blackout Documentation,Per 10CFR50.63.Licensee Will Complete Enhancements to Station Blackout Documentation Identified in Attachment 1 as Indicated & Other Items Will Be Completed within 2 Yrs
ML17309A4461990-07-0909 July 1990
Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions
ML17261B0901990-07-0909 July 1990
Advises That Licensee Will Install Containment Isolation Signal Going to Valve AOV 745 by End of 1992 Refueling Outage,Per Util 900608 Ltr Notifying of Condition Outside Design Basis of Plant Under 10CFR50.72 Criterion
ML17261B0931990-07-0909 July 1990
Responds to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Util Completed All Actions Requested in Generic Ltr & Will Retain Documentation Verification for Min of 2 Yrs
ML17250B2151990-06-29029 June 1990
Forwards Application for Amend to License DPR-18, Reformatting Auxiliary Electrical Sys Tech Specs & Action Statements for Offsite & Onsite Power Sources Available for Plant Auxiliaries
ML17261B0851990-06-28028 June 1990
Forwards LER 89-016-02 Re 891117 Failure of Safety Injection Block/Unblock Switch Which Could Render Both Trains of Safety Injection Sys Inoperable.Also Reported Per Part 21
ML17250B1991990-06-0808 June 1990
Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983
ML17250B1951990-06-0505 June 1990
Responds to Generic Ltr 89-08, Erosion/Corrosion Induced Pipe Wall Thinning. Util Developed Erosion/Corrosion Program for Single & Two Phase Sys Consistent W/Requirements of NUREG-1344 & NUMARC 870611 Rept
ML17261B0691990-06-0101 June 1990
Discusses Testing Frequency for Insp of Incore Neutron Monitoring Sys Thimble Tubes,Per NRC Bulletin 88-009.Thimble Tube Indicating Greatest Wear Recently Repositioned in Effort to Minimize Future Wear
ML17250B1851990-06-0101 June 1990
Forwards Application for Amend to License DPR-18,providing Guidance for Action Statements Associated W/Power Distribution Limit Specs
ML17250B1811990-05-31031 May 1990
Forwards Addl Info Re Response to Notice of Violation from Insp Rept 50-244/90-04.Info Withheld (Ref 10CFR73.21)
ML17261B0961990-05-29029 May 1990
Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-244/90-07.Corrective Actions:Worker Records Immediately Corrected Indicating That Worker Received No Significant Exposure for Time While Error Occurred
ML17250B1801990-05-29029 May 1990
Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Util Does Not Need to Develop Enhanced Surveillance Program to Monitor Currently Installed Transmitter Based Upon Limited Installed Quantity
ML17250B1841990-05-29029 May 1990
Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (SS) Internal Pre-Load Bolting.... Valves Disassembled & 410 SS Studs Removed & Visually & Liquid Penetrant Examined
ML17250B1721990-05-18018 May 1990
Forwards Rev 1 to Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. Rept Summarizes Observations & Corrective Actions Resulting from Insps Performed During 1990 Outage
ML20042G8931990-05-0808 May 1990
Forwards Revised Emergency Operating Procedures,Including Rev 3 to AP-RHR.2,Rev 1 to ES-0.3 & Rev 11 to ES-1.3
ML17261B0611990-04-26026 April 1990
Advises of Changes to 890201 Commitments Made Under Programmed Enhancement Response to Generic Ltr 88-17.Changes Will Not Reduce Capability to Operate Safely in Reduced Inventory Condition or Scope of Programmed Enhancements
ML17261B0591990-04-17017 April 1990
Advises That No Interlocks Required for RHR motor-operated Valves 701 & 720 Based on Present Arrangement.Listed Failures Would Have to Occur in Order for Potential Overpressurization of RHR Sys to Occur
ML17261B0461990-04-12012 April 1990
Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys
ML17262A1431990-04-12012 April 1990
Responds to NRC 900202 Ltr Re Weaknesses Noted in Insp Rept 50-244/89-80.Corrective Actions:Procedure ES-0.3 Modified to Provide Guidance for Rapid Cooldown & Depressurization W/ & W/O Reactor & Vessel Instrumentation Sys
ML17250B1451990-04-0606 April 1990
Discusses Impact of SER Issuance for Inservice Insp & Inservice Testing Programs & Advises That Timing Will Not Affect Util Implementation Plans for Programs,Except as Listed
ML17250B1441990-04-0505 April 1990
Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-244/90-02.Corrective Actions:Positive Indicator Will Be Installed for All Check Valves to Enhance Visible Positive Position Verification Ability & to Avoid Confusion
ML17261B0401990-03-30030 March 1990
Advises That Final Results of Station Blackout Documentation Review Will Be Submitted to NRC on or About 900501
ML17261B0321990-03-28028 March 1990
Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp
ML17261B0331990-03-28028 March 1990
Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date
ML17261B0231990-03-26026 March 1990
Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated
ML17261B0281990-03-23023 March 1990
Forwards Addl Info Re Proposed Tech Spec Amend Concerning Use of Reconstituted Fuel,Per 900315 Telcon.During Fuel Assembly Reconstitution,Failed Fuel Rods Will Be Placed W/ Filler Rods
ML17261B0221990-03-22022 March 1990
Provides Revised Test Schedule for motor-operated Valve Diagnostic Test Program,Per IE Bulletin 85-003.NRC Notification of Changes to Valve Operability Program Required by Generic Ltr 89-10,dtd 890828
ML17261B0411990-03-20020 March 1990
Forwards Summary of Onsite Property Damage Coverage Currently in Force at Plant,Per 10CFR50.54(w)(4)
ML17261B0151990-03-19019 March 1990
Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Plants (USI A-47). Overfill Protection Provided Through Trip Bistables in Reactor Protection Racks Powered from 120-volt Instrument Buses
ML17261B0101990-03-15015 March 1990
Forwards Application for Amend to License DPR-18,allowing Use of Reconstituted Fuel Assemblies
ML20012D2831990-03-14014 March 1990
Responds to Issues Discussed During 900307 Telcon W/Util & Eg&G Re Inservice Testing Program Status & Relief Request. Valves 5960A & 5960B Will Be Disassembled to Verify Forward Flow.Relief Requests PR-8 & VR-25 Encl
1990-09-18
[Table view] |
Text
I/ix/'Qllj~gslÃl/I
~Jlili$1illliiI!1<rlivi<piivl ROCHESTER GASANDELECTRICCORPORATION rl~~Oos~Oo89EASTAVENUE,ROCHESTER, N.Y.L4649LEONO.iVHITE,JR.VICCPRCSIDCNT TCLCPHOVC iPCiCOOCria546-2700April30,1980Mr.BoyceH.Grier,DirectorU.S.NuclearRegulatory Commission OfficeofInspection andEnforcement RegionI651ParkAvenueKingofPrussia,Pennsylvania 19406
Subject:
ResponsetoIEBulletinNo.'80-04o
DearMr.Grier.:
EnclosedisourresponsetoIEBulletinNo.80-04.ThisbulletinwasreceivedonFebruary8,1980.Sincerely, XC:OfficeofInspection andEnforcement DivisionofReactorOperations Inspection U.S.NuclearRegulatory Commission Washington, D.C.20555
ResonsetoIEBulletin80-04Request1.IResponse1.Reviewthecontainment pressureresponseanalysistodetermine ifthepotential forcontainment overpressure foramainsteamlinebreakinsidecontainment includedtheimpactofrunoutflowfromtheauxiliary feedwater systemandtheimpactofotherenergysources,suchascontinuation offeedwater orcondensate flow.Inyourreview,consider.
yourabilitytodetectandisolatethedamagedsteamgenerator fromthesesourcesandtheabilityofthepumpstoremainoperableafterextendedoperation at,runoutflow.AlthoughtheGinnapost-steamline breakcontainment pressureanalysisintheFSARdidnotincludetheeffectsofauxiliary feedwater flowtotheaffectedsteamgenerator, itisimportant torecognize thattheevaluation alsodidnotincludethebenefitsofpassiveandactiveheatsinksinsidecontainment.
Continued feedwater/condensate additiontothesteamgenerator willnotoccur,since,thesafetyinjection signal(generated byavarietyofprocessparameters, including highsteamlineflow,highcontainment
- pressure, andlowpressurizer pressure) willclosethefeedwater controlvalvesandstopthefeedwater pumps.Theadditionofmaximumauxiliary feedwater flowtothebrokensteamgenerator willeventually requireoperatoractionto1)realignflowtotheintactgenerator, 2)terminate auxiliary feedwater flowtothebrokengenerators.
Positiveinformation isavailable totheoperatortodetermine whichistheaffectedsteamgenerator.
Steamgenerator levelinstrumentation islocatedinsidecontainment andsteamgenerator pressureislocatedoutsidecontain-mentwhereit.wouldnotbeaffectedbytheaccidentenvironment, insidecontainment.
Itisexpectedthat,throughpropertrainingandbyuseoftheemergency procedures, theoperators willbecapableofquicklyrecognizing thesteamlinebreak,andwillperformtheproperoperations.
Thereissubstantial timeavailable fortheoperatortoperformthetwosafetyfunctions notedabove.TheSEPSafeShutdownreviewconcluded following theirsitevisitinJune1978thatonesteamgenerator wouldnotboildryforoverthirtyminutes.Thusthereissubstantial timetoalignflowtotheintactsteamgenerator.
Thetermination ofauxiliary feedwater flowtotheaffectedsteamgenerator, underthepessimistic circumstances, wouldrequiremorerapidaction(butstilleasilywithinthecapability oftheoperators) tomaintaincontainment
pressurebelowdesignpressure.
Theanalysispresented inAttachment 1concludes that,assumingminimumsafeguards for.containment cooling,auxiliary feedwater flowwouldhavetobeterminated inabout26minutes.Withmaximumsafeguards, thistimewouldbeextendedtoabout44minutes.Thereisnoneedtoconsidertheoperation oftheauxiliary feedwater pumpsatrunoutflow.Theturbine-driven pumpsarecontrolled bya'overnor, andwillnot,exceedabout400gpm.Themotordrivenpumpflowiscontrolled bytheAFWcontrolvalves,whichreceiveanautomatic throttlesignalto200gpmfromtheirflowcontrollers.
Apotential singlefailureoftheflowcontroller tocontrol.flowto200gpmisnotconsidered aworst-case singlefailureintermsofnetenergyadditiontothecontainment, sincetheoperation ofallcontainment coolingsafeguards (vs.theminimumsafeguards assumedinthisevaluation) wouldresultinasubstantial increaseinenergyremovalfromcontainment.
Request2.Reviewyouranalysisofthereactivity increasewhichresultsfromamainsteamlinebreakinsideor"outsidecontainment.
Thisreviewshouldconsiderthereactorcooldownrateandthepotential forthereactortoreturntopowerwiththemostreactivecontrolrodinthefullywithdrawn position.
Ifyourpreviousanalysisdidnotconsiderallpotential watersources(suchasthoselistedin1above)andifthereactivity increaseisgreaterthanpreviousanalysisindicated thereportofth'isreviewshouldinclude:aITheboundaryconditions fortheanalysis, e.g.,theendoflifeshutdownmargin,themoderator temperature coefficient, powerlevelandtheneteffectoftheassociated steamgenerator waterinventory onthereactorsystemcooling,etc.,b.Themostrestrictive singleactivefailureinthesafetyinjection systemandtheeffectofthatfailureondelayingthedeliveryofhighconcentration boricacidsolutiontothereactorcoolantsystem,cTheeffectofextendedwatersupplytotheaffectedsteamgenerator onthecorecriticality andreturntopower,
d.Thehotchannelfactorscorresponding tothemostreactiverodinthefullywithdrawn'osition attheendoflife,andtheMinimumDeparture fromNucleateBoilingRatio(MDNBR)valuesfortheanalyzedtransient.
Response2.Westinghouse ElectricCorporation performed theoriginalsteambreakanalysisforGinnaasreportedintheFSARandareanalysis submitted totheNRCinSeptember 1975.Westinghouse hasreviewedtheassumptions madeformainandauxiliary feedwater flowastheyapplytolicensing basissteamlinebreaktransients.
Severaloftherelevantassumptions usedinallcoretransient analysesfollow,andarefurtherexplained intheGinnaFSAR.1.Thereactorisassumedinitially tobeathotshutdownconditions, attheminimumallowable shutdownmargin.2.FortheCondition IVbreaks,i.e.,double-ended ruptureofamainsteampipe,fullmainfeedwater isassumedfromthebeginning ofthetransient ataveryconservative coldtemperature.
3.Allauxiliary feedwater pumpsareinitially assumedtobeoperating, inadditiontothemainfeedwater.
Theflowisequivalent totheratedflowofallpumpsatthesteamgenerator designpressure.
4Feedwater isassumedtocontinueatitsinitialflowrateuntilfeedwater isolation iscomplete, approximately 10secondsafterthebreakoccurs,whileauxiliary feedwater isassumedtocontinueatitsinitialflow'rate.5.Mainfeedwater flowiscompletely terminated following feedwater isolation.
Basedonthemannerinwhichtheanalysisisperformed
~forWestinghouse plants,thecoretransient resultsareveryinsensitive toauxiliary feedwater flow.Thefirstminuteofthetransient isdominated entirelybythesteamflowcontribution toprimary-secondary heattransfer, whichistheforcingfunctionforboththereactivity andthermal-hydraulic transients inthecore.Theeffectofauxiliary feedwater runout(orfailureofrunoutprotection whereapplicable) isminimal.Greaterfeedwater flowduringthelargesteamline breaksservestoreducesecondary pressures, accelerating theautomatic safeguards actions,i.e.steamline isolation, feedwater isolation andsafetyinjection.
Theassumptions described aboveare
therefore appropriate andconservative fortheshort-term aspectofthesteamline breaktransient.
Theauxiliary feedwater flowbecomesa-dominant factorindetermining thedurationand,magnitude ofthesteamflowtransient'uring laterstagesinthetransient.
However,thelimitingportionofthetransient occursduringthefirstminute,bothduetohighersteamflowsinherently presentearlyinthetransient andduetotheintroduction ofborontothecoreviathesafetyinjection system.Inconclusion, Westinghouse hasevaluated theeffectofrunoutauxiliary feedwater flowsinthecoretransient forsteamline break,andbasedonthisevaluation, hasdetermined thattheassumptions presently madeareappropriate foruseasalicensing basis.Theconcernsoutlinedintheintroduction toIEBulletin80-04relativeto,1)limitingcoreconditions occurring duringportionsofthetransient whereauxiliary feedwater flowisarelevantcontributor toplantcooldown; and2)incomplete isolation ofmainfeedwater flow,arenotrepresentative oftheWestinghouse NSSSdesignsandassociated BalanceofPlantrequirements.
Themostlimitingsteamlinebreakdetermined byWestinghouse wasanalyzedbyExxonNuclearCo.,Inc.andpresented inXN-NF-77-40 Supplement 1,"PlantTransient AnalysisfortheR.E.GinnaUnit1NuclearPowerPlant,"March1980.Thistransient occursathotzeropowerwithoutsidepoweravailable andthebreakoccurring attheexitofthesteamgenerator.
TheExxonanalysisdoesnotspecifically accountforauxiliary feedwater.
However,theSteamGenerator heat.transfermodel,usingconstantheattransfercoefficients, co'ntinues tocalculate heattransferfromtheprimarytothesecondary sideafterthebrokensteamgenerator hasbeenestimated tobeempty.Ifauxiliary flowwasspecifically accounted for,itseffectwouldbenegligible duringtheinitialportionofthetransient andwouldhaveminimaleffectduringlaterportionsofthetransient sincebythetimethebrokensteamgenerator empties,thetotalsystemreactivity isnegativeandcorepowerisdecreasing.
Theadditional reactivity additionassociated withtheslightcooldownduetorunoutflowismorethannegatedbytheboronreactivity insertedbysafetyinjection.
Therefore, theseverityofthetransient isnotincreased.
Request3.Ifthepotential forcontainment overpressure exists:orthereactor-return-to-power responseworsens,provideaproposedcorrective action.Iftheunitisoperating, provideadescription ofanyinterimactionthatwillbetakenuntiltheproposedcorrective actioniscompleted.
Response3.Sinceneitherthepotential forcontainment over-pressurization northereactor-return-to-power responseworsensnocorrective actionisrequired.
Request4.Within90daysofthedateofthisBulletin, completethereviewandevaluation requiredbythisBulletinandprovideawrittenresponsedescribing yourreviewsandactionstakeninresponsetoeachitem.Response4.Thisattachment providestherequired90dayresponsetoIEBulletinNo.80-04.
Attachment 1'-Containment EnerBalanceThepurposeofthisevaluation istodetermine thelengthoftimeavailable totheoperatortoterminate'AFW flowtothebrokensteamgenerator following asteamlinebreakinsidecontainment,.prior tocontainment pressureexceeding 60psig.InitialConditions andAssumtionsa.,ZOCAenergyreleasetocontainment (includes creditforcontainment heatsinks):191.3x10BTU(takenfromTable'14.3.4-2 ofFSAR).Resultsinpeakcontainment pressureof53psig.b.Additional energytoreach60psig:16x10BTU(takenfromFig.14.3.4-3ofFSAR).C.Minimumsafeguards heatremovalcapability (1spraypumpand2fancoolers)=55x10BTU/sec=3.3x106BTU/min.(takenfromFig.14.3.4-9of'theFSAR).d.Energyinputfrom600gpmtothebrokensteamIgenerator
=600galx1180BTUmanibmx62.6~lbx1ft=5.91x10B'fU6ft7.68galmz.ne.Energyreleasedtocontainment fromtheinitialsteamlinebreakaccidentblowdown=140x10BTU(takenfromFig.14.2.5-10 oftheFSAR).
Calculations l.Additional energywhichcouldbeabsorbedbycontainment, takingcreditforpassiveandminimumactivecontainment, heatsinksfollowing asteamlinebreak.(froma+b-2.eabove).(191.3+16)x10BTU-140x10BTU=67.3NBTU66Netenergyadditiontocontainment following initialsteamlinebreakblowdown(fromd-cabove):(5.91x10BTU-3.3x10BTU)=2.61x10BTU666manmz.nmz.n3.OperatorActionTime=67.3NBTU2.61NBTU/man f"lg
