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| {{#Wiki_filter:pCATEGORY1gjREGULATORY INFORMATION DISTRIBUTION SYSTEM(RIDS)ACCESSION NBR:9906290303 DOC.DATE: | | {{#Wiki_filter:p CATEGORY 1gj REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9906290303 DOC.DATE: 99/06/21 NOTARIZED: |
| 99/06/21NOTARIZED: | | NO FACIL:50-244,Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION ST MARTIN,J.T. |
| NOFACIL:50-244,Robert EmmetGinnaNuclearPlant,Unit1,Rochester GAUTH.NAMEAUTHORAFFILIATION STMARTIN,J.T.
| | Rochester Gas&: Electric Corp.MECREDY,R.C. |
| Rochester Gas&:ElectricCorp.MECREDY,R.C. | | Rochester Gas&, Electric Corp.RECIP.NAME RECIPIENT AFFILIATION DOCKET 05000244 VISSING,G.S. |
| Rochester Gas&,ElectricCorp.RECIP.NAME RECIPIENT AFFILIATION DOCKET05000244VISSING,G.S. | |
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| ==SUBJECT:== | | ==SUBJECT:== |
| LER99-001-01:on 990222,deficiencies inNSSSvendorsteam-linebrakemass&energyreleaseanalysisresultsinplantbeingoutsidedesignbasesoccurred.
| | LER 99-001-01:on 990222,deficiencies in NSSS vendor steam-line brake mass&energy release analysis results in plant being outside design bases occurred.Caused by deficiencies in W.Temporary administrative replaced.With 990621 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR I ENCL 1 SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). |
| Causedbydeficiencies inW.Temporary administrative replaced.
| | Q 05000244 0 RECIPIENT ID CODE/NAME LPD1-1 PD INTERNAL: AEOD/SPD/RRAB NRR/DIPM/IOLB NRR/DRIP/REXB RES/DET/EIB EXTERNAL: L ST LOBBY WARD NOAC POORE,W.NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME VISSING,G. |
| With990621ltr.DISTRIBUTION CODE:IE22TCOPIESRECEIVED:LTR IENCL1SIZE:TITLE:50.73/50.9 LicenseeEventReport(LER),IncidentRpt,etc.NOTES:License Expdateinaccordance with10CFR2,2.109(9/19/72).
| | FILE CEDER/DUPE IQMB NRR/DSSA/SPLB RGN1 FILE 01 LMITCO MARSHALL NOAC QUEENER,DS NUDOCS FULL TXT COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 R D'E N NOTE TO ALL"RIDS" RECIPIENTS: |
| Q050002440RECIPIENT IDCODE/NAME LPD1-1PDINTERNAL:
| | PLEASE HELP US TO REDUCE WASTE TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMEN DESK (DCD)ON EXTENSION 415-2083 HE DOCUMENT CONTROL FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 16 ENCL 16 0 |
| AEOD/SPD/RRAB NRR/DIPM/IOLB NRR/DRIP/REXB RES/DET/EIB EXTERNAL: | | ROCHESTER GAS AND ELECTRIC CORR7RATION |
| LSTLOBBYWARDNOACPOORE,W.NRCPDRCOPIESLTTRENCL1111111111111111RECIPIENT IDCODE/NAME VISSING,G.
| | +89 EAST AVENLJE, ROCHESTER, N.Y.14'-0001 AREA CODE 7I6 SrI6-2700 ROBERT C.MECREDY Vice President gvc!ear Operations June 21, 1999 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Guy S.Vissing Project Directorate I-1 Washington, D.C.20555 Subj ect: LER 1999-001, Revision 1, Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside its Design Basis R.E.Ginna Nuclear Power Plant Docket No.50-244 |
| FILECEDER/DUPEIQMBNRR/DSSA/SPLB RGN1FILE01LMITCOMARSHALLNOACQUEENER,DS NUDOCSFULLTXTCOPIESLTTRENCL1111111111111111RD'ENNOTETOALL"RIDS"RECIPIENTS:
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| PLEASEHELPUSTOREDUCEWASTETOHAVEYOURNAMEORORGANIZATION REMOVEDFROMDISTRIBUTION LISTSORREDUCETHENUMBEROFCOPIESRECEIVEDBYYOUORYOURORGANIZATION, CONTACTTHEDOCUMENDESK(DCD)ONEXTENSION 415-2083HEDOCUMENTCONTROLFULLTEXTCONVERSION REQUIREDTOTALNUMBEROFCOPIESREQUIRED:
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| LTTR16ENCL16 0
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| ROCHESTER GASANDELECTRICCORR7RATION | |
| +89EASTAVENLJE,ROCHESTER, N.Y.14'-0001AREACODE7I6SrI6-2700 ROBERTC.MECREDYVicePresident gvc!earOperations June21,1999U.S.NuclearRegulatory Commission DocumentControlDeskAttn:GuyS.VissingProjectDirectorate I-1Washington, D.C.20555 | |
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| ==Subject:== | | ==Dear Mr.Vissing:== |
| LER1999-001, Revision1,Deficiencies inNSSSVendorSteamline BreakMassandEnergyReleaseAnalysisResultsinPlantBeingOutsideitsDesignBasisR.E.GinnaNuclearPowerPlantDocketNo.50-244
| | The attached Licensee Event Report LER 1999-001, Revision 1, is submitted in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(ii)(B), which requires a report of,"Any event or condition...that resulted the nuclear power plant being...In a condition that was outside the design basis of the plant", and in accordance with 10 CFR 21,"Reporting of Defects and Noncompliances". |
| | | Very truly yours, Robert C.Mec edy xc: Mr.Guy S.Vissing (Mail Stop SC2)Project Directorate I-1 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation U.S.Nuclear Regulatory Commission Washington, D.C.20555 Regional Administrator, Region I U.S.Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S.NRC Ginna Senior Resident Inspector 0 9906290308 990b21 PDR ADOCK 05000244 S PDR |
| ==DearMr.Vissing:==
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| TheattachedLicenseeEventReportLER1999-001, Revision1,issubmitted inaccordance with10CFR50.73,LicenseeEventReportSystem,item(a)(2)(ii)(B),whichrequiresareportof,"Anyeventorcondition
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| ...thatresultedthenuclearpowerplantbeing...Inacondition thatwasoutsidethedesignbasisoftheplant",andinaccordance with10CFR21,"Reporting ofDefectsandNoncompliances". | |
| Verytrulyyours,RobertC.Mecedyxc:Mr.GuyS.Vissing(MailStopSC2)ProjectDirectorate I-1DivisionofReactorProjects-I/IIOfficeofNuclearReactorRegulation U.S.NuclearRegulatory Commission Washington, D.C.20555RegionalAdministrator, RegionIU.S.NuclearRegulatory Commission 475Allendale RoadKingofPrussia,PA19406U.S.NRCGinnaSeniorResidentInspector 09906290308 990b21PDRADOCK05000244SPDR
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| NRCFORM366IILIBBS)U.S.NUCLEARREGULATORY COMMISSION LICENSEEEVENTREPORT(LER)(Seereverseforrequirednumberofdigits/characters foreachblock)OPAL)WZ<~Is~>R;&0 AL:%N(P"informaaon cosectlon request:50hrs.Reportedlessonslearnedareincorporated htothelicensing processandfedbacktoindustry.
| | NRC FORM 366 IILI BBS)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)OPAL)WZ<~Is~>R;&0 A L:%N(P" informaaon cosectlon request: 50 hrs.Reported lessons learned are incorporated hto the licensing process and fed back to industry.Forward comments regarding burden estimate to the Records Management Branch (TA F33).U.S.Nudear Regulatory Commissbn. |
| Forwardcommentsregarding burdenestimatetotheRecordsManagement Branch(TAF33).U.S.NudearRegulatory Commissbn.
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| Washington. | | Washington. |
| DC205554001, andtothePaperwork Reduction Project(31~104).
| | DC 205554001, and to the Paperwork Reduction Project (31~104).Ofrce of Management and Budget.Washington, DC 20503.If an information colection does not display a currently valid OMB control number.Lhe NRC may not conduct or sponsor.and a FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DocKET NUMBER I2)05000244 PAGE I3)1 OF 8 TITlE tel Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside its Design Basis EVENT DATE (5)MONTH OAY 02 22 LER NUMBER (6)SEQUENTIAL REYSION NUMBER NUMBER 1999 1999-001" 01 MONTH OAY 06 21 1999 REPORT DATE (7)FACIUTY NAME FACILITY NAME DOCKET NUMBER 05000 DOCKET NVMBER 05000 OTHER FACILITIES INVOI.VED (6)OPERATING MODE (B)POWER LEVEL (10)75 20.2201(b), 20.2203(a)(1) 20.2203(a)(2)(i) 20.2203(a) |
| OfrceofManagement andBudget.Washington, DC20503.Ifaninformation colection doesnotdisplayacurrently validOMBcontrolnumber.LheNRCmaynotconductorsponsor.andaFACILITYNAME(1)R.E.GinnaNuclearPowerPlantDocKETNUMBERI2)05000244PAGEI3)1OF8TITlEtelDeficiencies inNSSSVendorSteamline BreakMassandEnergyReleaseAnalysisResultsinPlantBeingOutsideitsDesignBasisEVENTDATE(5)MONTHOAY0222LERNUMBER(6)SEQUENTIAL REYSIONNUMBERNUMBER19991999-001"01MONTHOAY06211999REPORTDATE(7)FACIUTYNAMEFACILITYNAMEDOCKETNUMBER05000DOCKETNVMBER05000OTHERFACILITIES INVOI.VED (6)OPERATING MODE(B)POWERLEVEL(10)7520.2201(b),
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| 20.2203(a)(1) 20.2203(a)(2)(i) 20.2203(a) | |
| (2)(ii)20.2203(a) | | (2)(ii)20.2203(a) |
| (2)(iii)20.2203(a) | | (2)(iii)20.2203(a) |
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| (2)(v)20.2203(a)(3)(i) | | (2)(v)20.2203(a)(3)(i) |
| '0.2203(a)(3)(ii) 20.2203(a) | | '0.2203(a)(3)(ii) 20.2203(a) |
| (4)50.36(c)(1)50.36(c)(2)LICENSEECONTACTFORTHISLER(12)5073(a)(2)(i)(B) 50.73(a)(2)(ii)50.73(a)(2)(iii) 50.73(a)(2) | | (4)50.36(c)(1)50.36(c)(2)LICENSEE CONTACT FOR THIS LER (12)50 73(a)(2)(i)(B) 50.73(a)(2)(ii)50.73(a)(2)(iii) 50.73(a)(2)(iv)~50.73(al(2)(v)50.73(a)(2)(vii) |
| (iv)~50.73(al(2) | | TS OF 10 CFR 5: (Check THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMEN one or more)(11)50.73(a)(2)(viii]50.73(a)(2)(x) 73.71 X OTHER Part 21 Specify In Abstract below or in NRC Form 366A TELEPHONE NVMSER aootvde Aree Code)John T.St.Martin-Technical Assistant (716)771-3641 CAUSE COMPONENT MANUPACTURER REPORTABLE TO EPOS SYSTEM COMPONENT MANUPACTURER REPORTABLE TO EPIX SUPPLEMENTAL REPORT EXPECTED (I4)YES (If yes, complete EXPECTED SUBMISSION DATE).NO X EXPECTED SUBMISSION DATE (15)MONTH OAY TEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On February 22, 1999, Ginna Station was notified by its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a main steam)inc break inside containment with an assumed single failure of a main feedwater regulating valve.These errors had a nonconservative impact on calculated peak containment pressure.The first error involved the omission of a previously unaccounted for volume of high temperature feedwater. |
| (v)50.73(a)(2)(vii) | | The second error resulted in an incorrect feedwater isolation time.The consequences of these two modeling errors previously created the potential for containment to exceed its design basis value of 60 psig for a main steamline break with a concurrent failure of a main feedwater regulating valve for R.E.Ginna Cycles 12 through 25.The cause of the plant being in a condition that was outside the design basis of the plant from Cycle 12 to Cycle 25 was deficiencies in the Westinghouse main steam)ine break mass and energy release analysis.Immediate corrective action for Cycle 27 was to place a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature, a measure of Service Water/Containment'fan cooler heat removal and pressure reduction capability, for Modes 1 through 4 (when containment is required to be operable)to ensure adequate post-accident environmental conditions margin existed.Corrective action to justify return to power for Cyc)e 28 included a cycle-specific analysis showing peak containment pressure less than 60 psig.Corrective action to prevent recurrence is outlined in Section V.B. |
| TSOF10CFR5:(CheckTHISREPORTISSUBMITTED PURSUANTTOTHEREQUIREMEN oneormore)(11)50.73(a)(2)(viii]50.73(a)(2)(x) 73.71XOTHERPart21SpecifyInAbstractbeloworinNRCForm366ATELEPHONE NVMSERaootvdeAreeCode)JohnT.St.Martin-Technical Assistant (716)771-3641CAUSECOMPONENT MANUPACTURER REPORTABLE TOEPOSSYSTEMCOMPONENT MANUPACTURER REPORTABLE TOEPIXSUPPLEMENTAL REPORTEXPECTED(I4)YES(Ifyes,completeEXPECTEDSUBMISSION DATE).NOXEXPECTEDSUBMISSION DATE(15)MONTHOAYTEARABSTRACT(Limitto1400spaces,i.e.,approximately 15single-spaced typewritten lines)(16)OnFebruary22,1999,GinnaStationwasnotifiedbyitsNSSSvendor(Westinghouse) oftwomodelingerrorsintheanalysisforamainsteam)inc breakinsidecontainment withanassumedsinglefailureofamainfeedwater regulating valve.Theseerrorshadanonconservative impactoncalculated peakcontainment pressure.
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| Thefirsterrorinvolvedtheomissionofapreviously unaccounted forvolumeofhightemperature feedwater.
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| Theseconderrorresultedinanincorrect feedwater isolation time.Theconsequences ofthesetwomodelingerrorspreviously createdthepotential forcontainment toexceeditsdesignbasisvalueof60psigforamainsteamline breakwithaconcurrent failureofamainfeedwater regulating valveforR.E.GinnaCycles12through25.Thecauseoftheplantbeinginacondition thatwasoutsidethedesignbasisoftheplantfromCycle12toCycle25wasdeficiencies intheWestinghouse mainsteam)ine breakmassandenergyreleaseanalysis.
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| Immediate corrective actionforCycle27wastoplaceatemporary administrative restriction of40degreesFahrenheit maximumonscreenhouse baytemperature, ameasureofServiceWater/Containment'fan coolerheatremovalandpressurereduction capability, forModes1through4(whencontainment isrequiredtobeoperable) toensureadequatepost-accident environmental conditions marginexisted.Corrective actiontojustifyreturntopowerforCyc)e28includedacycle-specific analysisshowingpeakcontainment pressurelessthan60psig.Corrective actiontopreventrecurrence isoutlinedinSectionV.B. | |
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| NRCFORM366AIBIBBB)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAMEI1)R.E.GinnaNuclearPowerPlantDOCKETIz)05000244LERNUMBER(6)BEOUENTNL REVISIONNUMBERNUMBER1999-001-01PAGEI3)2OFI)TEXTflfmorespeceisrequired, useeddirionel copiesofNRCForm366A)I17)PRE-EVENT PLANTCONDITIONS:
| | NRC FORM 366A IB IBBB)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME I1)R.E.Ginna Nuclear Power Plant DOCKET Iz)05000244 LER NUMBER (6)BEOUENTNL REVISION NUMBER NUMBER 1999-001-01 PAGE I3)2 OF I)TEXT flf more speceis required, use eddirionel copies of NRC Form 366A)I17)PRE-EVENT PLANT CONDITIONS: |
| In1992,theNSSSvendorforGinnaStation(Westinghouse) analyzedafullspectrumofcasesforthemainsteamlinebreak(MSLB)andcontainment responsefortheR.E.Ginnaplant.Thesetofanalyseswasperformed tosupporttheprogramtoreducetheboricacidstoragetank(BAST)boronconcentration to2000ppm.Thisanalysisentailedover120cases,whichvariedbreaksizeandlocation, initialpowerlevel,singlefailure,offsitepoweravailability, anddifferent reactivity coefficients (roddedversusunrodded)
| | In 1992, the NSSS vendor for Ginna Station (Westinghouse) analyzed a full spectrum of cases for the main steam line break (MSLB)and containment response for the R.E.Ginna plant.The set of analyses was performed to support the program to reduce the boric acid storage tank (BAST)boron concentration to 2000 ppm.This analysis entailed over 120 cases, which varied break size and location, initial power level, single failure, offsite power availability, and different reactivity coefficients (rodded versus unrodded)~In 1995, Westinghouse re-visited the steamline break mass and energy release event due to replacement steam generators (RSGs)and an increase to 18 month fuel cycles.Many of the previous cases could be eliminated due to: An integral flow restrictor on the RSGs, eliminating any breaks larger than 1.4 square feet (the RSGs have flow restricting outlet nozzles which make any break size larger than the throat area of the restrictor not credible and thus, break flow areas greater than 1.4 square feet no longer need to be considered), and 2.Credit being taken for the main steam non-return check valve, which eliminated several types of breaks and the need to consider a main steam isolation valve (MSIV)single failure for containment overpressurization cases.After eliminating cases that were no longer credible, cases for the 1995 analyses were selected from the 1992 BAST reduction program.Six of the most limiting cases were selected, including different break sizes, initial power levels, single failures, and offsite power assumptions. |
| ~In1995,Westinghouse re-visited thesteamline breakmassandenergyreleaseeventduetoreplacement steamgenerators (RSGs)andanincreaseto18monthfuelcycles.Manyofthepreviouscasescouldbeeliminated dueto:Anintegralflowrestrictor ontheRSGs,eliminating anybreakslargerthan1.4squarefeet(theRSGshaveflowrestricting outletnozzleswhichmakeanybreaksizelargerthanthethroatareaoftherestrictor notcredibleandthus,breakflowareasgreaterthan1.4squarefeetnolongerneedtobeconsidered), | | Another difference from the previous analyses is explained in the Ginna Station Updated Final Safety Analysis Report (UFSAR)Section 6.2.1.2.3: "In the Reference 3 analyses[1992 BAST reduction program), the most limiting feedwater control system failure was assumed to be a failure in the auxiliary feedwater system control.This failure resulted in increased auxiliary feedwater flow rates for the entire duration of the transient. |
| and2.Creditbeingtakenforthemainsteamnon-return checkvalve,whicheliminated severaltypesofbreaksandtheneedtoconsideramainsteamisolation valve(MSIV)singlefailureforcontainment overpressurization cases.Aftereliminating casesthatwerenolongercredible, casesforthe1995analyseswereselectedfromthe1992BASTreduction program.Sixofthemostlimitingcaseswereselected, including different breaksizes,initialpowerlevels,singlefailures, andoffsitepowerassumptions.
| | A failure of the isolation function of the main feedwater regulating valves (MFRVs)was not considered because a feedwater isolation signal also results in a main feedwater pump trip and in closure of the pump discharge valve (MFPDV).Thus, no single failure would have resulted in a failure to isolate main feedwater or a significant delay in the isolation of main feedwater to the affected steam generator. |
| Anotherdifference fromthepreviousanalysesisexplained intheGinnaStationUpdatedFinalSafetyAnalysisReport(UFSAR)Section6.2.1.2.3:
| | It has subsequently been realized that the condensate pumps would continue to pump through the main feedwater pumps at low steam generator pressures and that the pump discharge valve (MFPDV)closure time is approximately 80 seconds.Therefore, a failure of the isolation function of the main feedwater regulating valve (MFRV)could result in main feedwater flow via the condensate pumps for approximately 80 seconds to the affected steam generator. |
| "IntheReference 3analyses[1992BASTreduction program), | | This new failure scenario is a more limiting failure than was assumed in the previous analyses due to the timing and magnitude of the increased flow." The requirement for the 80 second stroke time requirement of the MFPDV is addressed in the Ginna Station Improved Technical Specifications (ITS)Section 3.7.3. |
| themostlimitingfeedwater controlsystemfailurewasassumedtobeafailureintheauxiliary feedwater systemcontrol.Thisfailureresultedinincreased auxiliary feedwater flowratesfortheentiredurationofthetransient.
| | NRC FORM 366A (B IBBS)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)DOCKET (2)LER NUMBER (6)SEmINTIAE AEVISIBH NUMB(A IIUMBEA PAGE (3)R.E.Ginna Nuclear Power Plant 05000244 1 ass ()()1 3 OF 8 TEXT fff more speceis required, use eddidonel copies of NRC Form 366A/(17)DESCRIPTION OF EVENT: DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES: |
| Afailureoftheisolation functionofthemainfeedwater regulating valves(MFRVs)wasnotconsidered becauseafeedwater isolation signalalsoresultsinamainfeedwater pumptripandinclosureofthepumpdischarge valve(MFPDV).Thus,nosinglefailurewouldhaveresultedinafailuretoisolatemainfeedwater orasignificant delayintheisolation ofmainfeedwater totheaffectedsteamgenerator.
| | February 22, 1999: Event date.February 22, 1999, 1540 EST: Discovery date and time.February 22, 1999, 1601 EST: NRC is notified of this condition per 10CFR50.72 (b)(1)(ii)(8).March 1, 1999, 2018 EST: Plant is shutdown in Mode 4 for the planned 1999 refueling outage.April 16, 1999: NSSS analysis shows MSLB containment pressure is less than 60 psig.B.EVENT: On February 22, 1999, in activities unrelated to plant conditions, Ginna Station was notified by its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a MSLB inside containment with an assumed single failure of a MFRV.(Refer to the letter RGE-99-207 from Mr.Steve M.Ira of the Westinghouse Electric Corporation to Mr.Peter Bamford, Rochester Gas and Electric Corporation (RGSE), dated February 22, 1999, |
| Ithassubsequently beenrealizedthatthecondensate pumpswouldcontinuetopumpthroughthemainfeedwater pumpsatlowsteamgenerator pressures andthatthepumpdischarge valve(MFPDV)closuretimeisapproximately 80seconds.Therefore, afailureoftheisolation functionofthemainfeedwater regulating valve(MFRV)couldresultinmainfeedwater flowviathecondensate pumpsforapproximately 80secondstotheaffectedsteamgenerator.
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| Thisnewfailurescenarioisamorelimitingfailurethanwasassumedinthepreviousanalysesduetothetimingandmagnitude oftheincreased flow."Therequirement forthe80secondstroketimerequirement oftheMFPDVisaddressed intheGinnaStationImprovedTechnical Specifications (ITS)Section3.7.3.
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| NRCFORM366A(BIBBS)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)DOCKET(2)LERNUMBER(6)SEmINTIAE AEVISIBHNUMB(AIIUMBEAPAGE(3)R.E.GinnaNuclearPowerPlant050002441ass()()13OF8TEXTfffmorespeceisrequired, useeddidonel copiesofNRCForm366A/(17)DESCRIPTION OFEVENT:DATESANDAPPROXIMATE TIMESOFMAJOROCCURRENCES:
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| February22,1999:Eventdate.February22,1999,1540EST:Discovery dateandtime.February22,1999,1601EST:NRCisnotifiedofthiscondition per10CFR50.72 (b)(1)(ii)(8).March1,1999,2018EST:PlantisshutdowninMode4fortheplanned1999refueling outage.April16,1999:NSSSanalysisshowsMSLBcontainment pressureislessthan60psig.B.EVENT:OnFebruary22,1999,inactivities unrelated toplantconditions, GinnaStationwasnotifiedbyitsNSSSvendor(Westinghouse) oftwomodelingerrorsintheanalysisforaMSLBinsidecontainment withanassumedsinglefailureofaMFRV.(RefertotheletterRGE-99-207 fromMr.SteveM.IraoftheWestinghouse ElectricCorporation toMr.PeterBamford,Rochester GasandElectricCorporation (RGSE),datedFebruary22,1999,
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| ==Subject:== | | ==Subject:== |
| Steamline BreakMassandEnergyReleaseAnalysisNonconformance.) | | Steamline Break Mass and Energy Release Analysis Nonconformance.) |
| Theseerrorshadanonconservative impactoncalculated peakcontainment pressure.
| | These errors had a nonconservative impact on calculated peak containment pressure.The first error involves the volume of feedwater that exists between the specified isolation valve and the faulted steam generator. |
| Thefirsterrorinvolvesthevolumeoffeedwater thatexistsbetweenthespecified isolation valveandthefaultedsteamgenerator.
| | When the MFRV is assumed to fail to perform its isolation function, then the isolation function is performed by the MFPDV.Because of the relative locations of these valves to the steam generator, the volume of water is greater between the MFPDV and the steam generator than between the MFRV and the steam generator. |
| WhentheMFRVisassumedtofailtoperformitsisolation
| | This increased volume of high temperature feedwater was not accounted for in the analysis.The second error was inadvertently isolating feedwater flow at 15 seconds instead of at 85 seconds (80 seconds valve stroke time and 5 seconds for signal delay)as was intended.This meant that feedwater flow should have continued for 70 seconds longer than was assumed in the analysis.The consequences of these two modeling errors has the potential for containment to exceed its design basis value of 60 psig for a MSLB with an assumed concurrent failure of a MFRV.This is due to adding higher energy fluid to containment for 70 seconds longer than was assumed in the 1995 MSLB analysis of record.There was also the potential for the containment to exceed its design basis value as documented in the 1992 analyses, due to discovery of a more limiting single failure. |
| : function, thentheisolation functionisperformed bytheMFPDV.Becauseoftherelativelocations ofthesevalvestothesteamgenerator, thevolumeofwaterisgreaterbetweentheMFPDVandthesteamgenerator thanbetweentheMFRVandthesteamgenerator.
| | NRC FORM 366A (9 19991 LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET (2)05000244 LER NUMBER (6)BEUUENTIAL AENSION NUMBEA NUMBER 1999-001-01 PAGE (3)4 OF 8 TEXT llf more spaceis required, use additional copies of NRC Form 366AI (17)At the time of discovery and until the planned shutdown on March 1, 1999, the screenhouse bay temperature was maintained at approximately 39 degrees Fahrenheit, while lake temperature remained at approximately 35 degrees Fahrenheit. |
| Thisincreased volumeofhightemperature feedwater wasnotaccounted forintheanalysis.
| | Ginna Station placed a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature, a measure of Service Water System temperature, for Modes 1 through 4 (when containment is required to be operable)to ensure adequate margin existed.By maintaining this temperature less than 40 degrees Fahrenheit, increased containment fan cooler heat removal ability was ensured and this, combined with a high cycle-specific excess shutdown margin, was sufficient to ensure containment pressure would remain less than 60 psig for this postulated scenario.To support operation in Cycle 28 and beyond, the MSLB analysis (with the assumed MFRV failure)was reanalyzed. |
| Theseconderrorwasinadvertently isolating feedwater flowat15secondsinsteadofat85seconds(80secondsvalvestroketimeand5secondsforsignaldelay)aswasintended.
| | Significant assumption changes to support operation were as follows: Assumed shutdown margin of 2.40%versus the base case reference cycle value of 1.80%.The Ginna Station Core Operating Limits Report (COLR)was subsequently revised to incorporate this value.P Assumed operation at the normal Ginna Station average operating temperature (Tavg)of 561 degrees F versus the upper end of the analyzed Tavg window of 573.5 degrees F.This operating restriction has been placed in Ginna Station operating procedures and corresponds to the full power rod control program setpoint.These changes, when combined with several minor modeling refinements, led to peak containment pressure under this scenario to be less than 60 psig.Accordingly, operations for all of Cycle 28 was justified. |
| Thismeantthatfeedwater flowshouldhavecontinued for70secondslongerthanwasassumedintheanalysis.
| | This includes the removal of the lake temperature restriction previously imposed.C.INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: None D.OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None E.METHOD OF DISCOVERY: |
| Theconsequences ofthesetwomodelingerrorshasthepotential forcontainment toexceeditsdesignbasisvalueof60psigforaMSLBwithanassumedconcurrent failureofaMFRV.Thisisduetoaddinghigherenergyfluidtocontainment for70secondslongerthanwasassumedinthe1995MSLBanalysisofrecord.Therewasalsothepotential forthecontainment toexceeditsdesignbasisvalueasdocumented inthe1992analyses, duetodiscovery ofamorelimitingsinglefailure.
| | This event was disclosed to RGSE by the notification from its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a MSLB inside containment with an assumed single failure of a MFRV.(Refer to letter RGE-99-207 from Mr.Steve M.Ira of the Westinghouse Electric Corporation to Mr.Peter Bamford, Rochester Gas and Electric Corporation (RG(t(E), dated February 22, 1999, |
| NRCFORM366A(919991LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKET(2)05000244LERNUMBER(6)BEUUENTIAL AENSIONNUMBEANUMBER1999-001-01PAGE(3)4OF8TEXTllfmorespaceisrequired, useadditional copiesofNRCForm366AI(17)Atthetimeofdiscovery anduntiltheplannedshutdownonMarch1,1999,thescreenhouse baytemperature wasmaintained atapproximately 39degreesFahrenheit, whilelaketemperature remainedatapproximately 35degreesFahrenheit.
| |
| GinnaStationplacedatemporary administrative restriction of40degreesFahrenheit maximumonscreenhouse baytemperature, ameasureofServiceWaterSystemtemperature, forModes1through4(whencontainment isrequiredtobeoperable) toensureadequatemarginexisted.Bymaintaining thistemperature lessthan40degreesFahrenheit, increased containment fancoolerheatremovalabilitywasensuredandthis,combinedwithahighcycle-specific excessshutdownmargin,wassufficient toensurecontainment pressurewouldremainlessthan60psigforthispostulated scenario.
| |
| Tosupportoperation inCycle28andbeyond,theMSLBanalysis(withtheassumedMFRVfailure)wasreanalyzed.
| |
| Significant assumption changestosupportoperation wereasfollows:Assumedshutdownmarginof2.40%versusthebasecasereference cyclevalueof1.80%.TheGinnaStationCoreOperating LimitsReport(COLR)wassubsequently revisedtoincorporate thisvalue.PAssumedoperation atthenormalGinnaStationaverageoperating temperature (Tavg)of561degreesFversustheupperendoftheanalyzedTavgwindowof573.5degreesF.Thisoperating restriction hasbeenplacedinGinnaStationoperating procedures andcorresponds tothefullpowerrodcontrolprogramsetpoint. | |
| Thesechanges,whencombinedwithseveralminormodelingrefinements, ledtopeakcontainment pressureunderthisscenariotobelessthan60psig.Accordingly, operations forallofCycle28wasjustified.
| |
| Thisincludestheremovalofthelaketemperature restriction previously imposed.C.INOPERABLE STRUCTURES, COMPONENTS, ORSYSTEMSTHATCONTRIBUTED TOTHEEVENT:NoneD.OTHERSYSTEMSORSECONDARY FUNCTIONS AFFECTED:
| |
| NoneE.METHODOFDISCOVERY:
| |
| Thiseventwasdisclosed toRGSEbythenotification fromitsNSSSvendor(Westinghouse) oftwomodelingerrorsintheanalysisforaMSLBinsidecontainment withanassumedsinglefailureofaMFRV.(RefertoletterRGE-99-207 fromMr.SteveM.IraoftheWestinghouse ElectricCorporation toMr.PeterBamford,Rochester GasandElectricCorporation (RG(t(E),
| |
| datedFebruary22,1999,
| |
|
| |
|
| ==Subject:== | | ==Subject:== |
| Steamline BreakMassandEnergyReleaseAnalysisNonconformance.) | | Steamline Break Mass and Energy Release Analysis Nonconformance.) |
| NRCFORM366AISISSSILICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME11IDOCKETl2lLERNUMBERI6)PAGEI3IR.E.GinnaNuclearPowerPlant05000244SEGUENEME REVISIUNNUMBERNUMBER1999-001-015OF8TEXTllfmorespaceisrequired, useaddidonal copiesofNRCForm366Al117)OPERATORACTION:AfterreviewofthisWestinghouse correspondence, ReactorEngineering andAnalysispersonnel notifiedOperations supervision, whonotifiedtheControlRoomoperators.
| | NRC FORM 366A IS ISSSI LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME 11I DOCKET l2l LER NUMBER I6)PAGE I 3 I R.E.Ginna Nuclear Power Plant 05000244 SEGUENEME REVISIUN NUMBER NUMBER 1999-001-01 5 OF 8 TEXT llf more spaceis required, use addidonal copies of NRC Form 366Al 117)OPERATOR ACTION: After review of this Westinghouse correspondence, Reactor Engineering and Analysis personnel notified Operations supervision, who notified the Control Room operators. |
| TheNRCResidentwasnotifiedatthistime.Atapproximately 1540ESTonFebruary22,1999,plantstaffdetermined thatanon-emergency onehournotification, per10CFR50.72 (b)(1)(ii)(B),shouldbemadetotheNRCOperations Center.TheShiftSupervisor madethisnotification atapproximately 1601ESTonFebruary22,1999.Operators maintained arestriction of40degreesFahrenheit maximumonscreenhouse baytemperature fromthetimeofdiscovery untilplantshutdownonMarch1toensureadequatemarginexisted.AnalysesforCycle28havebeencompleted.
| | The NRC Resident was notified at this time.At approximately 1540 EST on February 22, 1999, plant staff determined that a non-emergency one hour notification, per 10CFR50.72 (b)(1)(ii)(B), should be made to the NRC Operations Center.The Shift Supervisor made this notification at approximately 1601 EST on February 22, 1999.Operators maintained a restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature from the time of discovery until plant shutdown on March 1 to ensure adequate margin existed.Analyses for Cycle 28 have been completed. |
| Theseanalysesdemonstrate thatthereisadequatecontainment (CNMT)pressuremarginforaMSLBforCycle28.Therestriction previously imposedonlaketemperature waslifted.G.SAFETYSYSTEMRESPONSES:
| | These analyses demonstrate that there is adequate containment (CNMT)pressure margin for a MSLB for Cycle 28.The restriction previously imposed on lake temperature was lifted.G.SAFETY SYSTEM RESPONSES: |
| NoneIII.CAUSEOFEVENT:A.IMMEDIATE CAUSE:Theimmediate causeottheplantbeinginacondition thatwasoutsidethedesignbasisoftheplantwasdeficiencies intheWestinghouse MSLBmassandenergyreleaseanalysis.
| | None III.CAUSE OF EVENT: A.IMMEDIATE CAUSE: The immediate cause ot the plant being in a condition that was outside the design basis of the plant was deficiencies in the Westinghouse MSLB mass and energy release analysis.B.INTERMEDIATE CAUSE: The intermediate cause of the deficiencies in the Westinghouse analysis was the analysis did not account for a greater volume of high temperature feedwater and used an incorrect feedwater isolation time.The additional mass and energy release from the MSLB are penalties in the containment pressure calculation. |
| B.INTERMEDIATE CAUSE:Theintermediate causeofthedeficiencies intheWestinghouse analysiswastheanalysisdidnotaccountforagreatervolumeofhightemperature feedwater andusedanincorrect feedwater isolation time.Theadditional massandenergyreleasefromtheMSLBarepenalties inthecontainment pressurecalculation. | | C.ROOT CAUSE: The underlying cause of the error in the volume of feedwater assumption was insufficient knowledge by Westinghouse personnel of the specific configuration of the Ginna Station feedwater piping.The Underlying cause of the error in the assumed feedwater isolation time was human error in the use of the LOFTRAN code by Westinghouse personnel. |
| C.ROOTCAUSE:Theunderlying causeoftheerrorinthevolumeoffeedwater assumption wasinsufficient knowledge byWestinghouse personnel ofthespecificconfiguration oftheGinnaStationfeedwater piping.TheUnderlying causeoftheerrorintheassumedfeedwater isolation timewashumanerrorintheuseoftheLOFTRANcodebyWestinghouse personnel. | | NRC FORM 366A IB 1999)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME{1)R.E.Ginna Nuclear Power Plant DOCKET (2)05000244 LER NUMBER I6)SEQUENTUL REVIION NUMBER NUMBER 1999-001-01 PAGE I3)6 OF 8 TEXT llf more space is required, use eddidonel copies of NRC Form 366A/I17)IV.ANALYSIS OF EVENT: This event is reportable in accordance with 10 CFR 21 and in accordance with 10 CFR 50.73, Licensee Event Report System, item (6)(2)(ii)(B), which requires a report of,"Any event or condition...that resulted in the nuclear power plant being...In a condition that was outside the design basis of the plant".The modeling errors in the Westinghouse MSLB mass and energy release analysis had the potential to pressurize the Ginna Station containment beyond its design pressure due to this postulated post-accident scenario.An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions: |
| NRCFORM366AIB1999)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME{1)R.E.GinnaNuclearPowerPlantDOCKET(2)05000244LERNUMBERI6)SEQUENTUL REVIIONNUMBERNUMBER1999-001-01PAGEI3)6OF8TEXTllfmorespaceisrequired, useeddidonel copiesofNRCForm366A/I17)IV.ANALYSISOFEVENT:Thiseventisreportable inaccordance with10CFR21andinaccordance with10CFR50.73,LicenseeEventReportSystem,item(6)(2)(ii)(B),whichrequiresareportof,"Anyeventorcondition
| | There were no operational or safety consequences or implications attributed to the deficiencies in the Westinghouse MSLB mass and energy release analysis because: The MSLB containment analysis is presented in Ginna Station UFSAR Section 6.2.1.2.3. |
| ...thatresultedinthenuclearpowerplantbeing...Inacondition thatwasoutsidethedesignbasisoftheplant".ThemodelingerrorsintheWestinghouse MSLBmassandenergyreleaseanalysishadthepotential topressurize theGinnaStationcontainment beyonditsdesignpressureduetothispostulated post-accident scenario. | | A revised analysis was performed (in March/April 1999)by offsetting the adverse effects of the additional mass and energy release through refinements in the existing analysis assumptions. |
| Anassessment wasperformed considering boththesafetyconsequences andimplications ofthiseventwiththefollowing resultsandconclusions:
| | These improvements do not alter the total quantity of steam released to containment, but provide benefits in lowering the rate of the release and increasing the rate of contairiment heat removal.Applying these analysis improvements to the MFRV failure cases yields peak containment pressures that are less than the containment design pressure.The Westinghouse analysis, as of March 24, 1999, showed that the peak containment pressure design basis value of 60 psig would not be exceeded provided that service water (SW)temperature (SW supplies the containment recirculation fans)is limited to less than 45 degrees Fahrenheit. |
| Therewerenooperational orsafetyconsequences orimplications attributed tothedeficiencies intheWestinghouse MSLBmassandenergyreleaseanalysisbecause:TheMSLBcontainment analysisispresented inGinnaStationUFSARSection6.2.1.2.3.
| | Prior to entering Mode 4 for the planned 1999 refueling outage, the screenhouse bay temperature was being maintained at approximately 39 degrees Fahrenheit, while lake temperature was approximately 35 degrees Fahrenheit. |
| Arevisedanalysiswasperformed (inMarch/April 1999)byoffsetting theadverseeffectsoftheadditional massandenergyreleasethroughrefinements intheexistinganalysisassumptions.
| | Thus, plant operation remained within its design basis since the time of discovery. |
| Theseimprovements donotalterthetotalquantityofsteamreleasedtocontainment, butprovidebenefitsinloweringtherateofthereleaseandincreasing therateofcontairiment heatremoval.Applyingtheseanalysisimprovements totheMFRVfailurecasesyieldspeakcontainment pressures thatarelessthanthecontainment designpressure.
| | The current Westinghouse analysis, completed during the 1999 refueling outage, shows that the peak containment pressure design basis value of 60 psig was not exceeded for Cycle 27, nor will it be exceeded for Cycle 28 (the current operating cycle).Operation during future cycles are also justified, with restrictions similar to those imposed for Cycle 28.Any small increase in containment pressure beyond the 60 psig design basis value would not be expected to fail the containment. |
| TheWestinghouse
| | In 1996, the containment was successfully tested to 72 psig during the Structural Integrity Test for the steam.generator replacement project.A steam line break does not result in severe radiological consequences at Ginna Station;thus full containment integrity is not considered required for this event (even though containment isolation is assumed to occur). |
| : analysis, asofMarch24,1999,showedthatthepeakcontainment pressuredesignbasisvalueof60psigwouldnotbeexceededprovidedthatservicewater(SW)temperature (SWsuppliesthecontainment recirculation fans)islimitedtolessthan45degreesFahrenheit.
| | NRC FORM 366A (6 1998)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)DOCKET (2)LER NUMBER (6)PAGE (3)R.E.Ginna Nuclear Power Plant 05000244 SEI)ENTIAE REVSIOM HUMBER)(UMBER 1999-.001-01 7 OF B TEXT ill more speceis required, use addi tionel copies ol NRC Form 366Al (17)o For Ginna Station Cycles 1 (the original cycle in 1970)through 11, the MSLB containment pressure was not part of the Ginna Station licensing basis.For Cycles 12 through 25, containment pressure on a MSLB was part of the Ginna Station licensing basis and the potential existed to exceed 60 psig under the postulated accident scenario described above.Cycles 26 and 27 operated at the current Tavg of 561 degrees F, and would have had a peak containment pressure less than 60 psig under the postulated conditions. |
| PriortoenteringMode4fortheplanned1999refueling outage,thescreenhouse baytemperature wasbeingmaintained atapproximately 39degreesFahrenheit, whilelaketemperature wasapproximately 35degreesFahrenheit.
| | Cycle 28 has also been shown be acceptable. |
| Thus,plantoperation remainedwithinitsdesignbasissincethetimeofdiscovery. | | o At no time in the operation of Ginna Station has a main steam line break occurred.Based on the above, it can be concluded that there were no unreviewed safety questions, and that the public's health and safety was assured at all times.V.CORRECTIVE ACTION: A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: For the operation of Ginna Station between the time of discovery and plant shutdown on March 1, 1999, the screenhouse bay temperature was maintained at approximately 39 degrees Fahrenheit, while lake temperature remained at approximately 35 degrees Fahrenheit. |
| ThecurrentWestinghouse
| | Thus, plant operation remained within its design basis from the discovery date until the 1999 refueling outage.Ginna Station placed a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature for Modes 1 through 4 (when containment is required to be operable)to ensure adequate margin existed.Completion of analyses demonstrated that operation for Cycle 28 would be within design basis.The restriction previously imposed on lake temperature was lifted.'B.ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE: |
| : analysis, completed duringthe1999refueling outage,showsthatthepeakcontainment pressuredesignbasisvalueof60psigwasnotexceededforCycle27,norwillitbeexceededforCycle28(thecurrentoperating cycle).Operation duringfuturecyclesarealsojustified, withrestrictions similartothoseimposedforCycle28.Anysmallincreaseincontainment pressurebeyondthe60psigdesignbasisvaluewouldnotbeexpectedtofailthecontainment.
| | o Additional analyses were performed to support current and future cycle operation. |
| In1996,thecontainment wassuccessfully testedto72psigduringtheStructural Integrity Testforthesteam.generator replacement project.Asteamlinebreakdoesnotresultinsevereradiological consequences atGinnaStation;thusfullcontainment integrity isnotconsidered requiredforthisevent(eventhoughcontainment isolation isassumedtooccur).
| | The MSLB analysis (with the assumed MFRV failure)was reanalyzed. |
| NRCFORM366A(61998)LICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)DOCKET(2)LERNUMBER(6)PAGE(3)R.E.GinnaNuclearPowerPlant05000244SEI)ENTIAE REVSIOMHUMBER)(UMBER1999-.001-017OFBTEXTillmorespeceisrequired, useadditionelcopiesolNRCForm366Al(17)oForGinnaStationCycles1(theoriginalcyclein1970)through11,theMSLBcontainment pressurewasnotpartoftheGinnaStationlicensing basis.ForCycles12through25,containment pressureonaMSLBwaspartoftheGinnaStationlicensing basisandthepotential existedtoexceed60psigunderthepostulated accidentscenariodescribed above.Cycles26and27operatedatthecurrentTavgof561degreesF,andwouldhavehadapeakcontainment pressurelessthan60psigunderthepostulated conditions.
| | Significant assumption changes to support operation were as follows: Assumed shutdown margin of 2.40%versus the base case reference cycle value of 1.80%.The Ginna Station Core Operating Limits Report (COLR)was subsequently revised to incorporate this value. |
| Cycle28hasalsobeenshownbeacceptable.
| | NRC FORM 366h IB IBBBI LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET Iz)05000244 LER NUMSER (6)BrppaRML RNBION IIUMBTR IIUMBER 1999-001-01 PAGE I3)8 OF 8 TEXT (If more spaceis required, use additional copies of NRC Form 866AI I17)Assumed operation at the normal Ginna Station average operating temperature (Tavg)of 561 degrees F versus the upper end of the analyzed Tavg window of 573.5 degrees F.This operating restriction has been placed in Ginna Station operating procedures and corresponds to the full power rod control program setpoint.Plant modifications to recover the full range of the previously analyzed Tavg window and to restore the previous shutdown margin limits will be considered in the future.Nuclear Engineering Services (NES)has implemented a process for independent review of vendor inputs supplied by RG&E and is in the process of performing an independent review of all accident analysis assumptions. |
| oAtnotimeintheoperation ofGinnaStationhasamainsteamlinebreakoccurred.
| | Westinghouse is evaluating the need for internal corrective actions as part of their review of this condition. |
| Basedontheabove,itcanbeconcluded thattherewerenounreviewed safetyquestions, andthatthepublic'shealthandsafetywasassuredatalltimes.V.CORRECTIVE ACTION:A.ACTIONTAKENTORETURNAFFECTEDSYSTEMSTOPRE-EVENT NORMALSTATUS:Fortheoperation ofGinnaStationbetweenthetimeofdiscovery andplantshutdownonMarch1,1999,thescreenhouse baytemperature wasmaintained atapproximately 39degreesFahrenheit, whilelaketemperature remainedatapproximately 35degreesFahrenheit.
| |
| Thus,plantoperation remainedwithinitsdesignbasisfromthediscovery dateuntilthe1999refueling outage.GinnaStationplacedatemporary administrative restriction of40degreesFahrenheit maximumonscreenhouse baytemperature forModes1through4(whencontainment isrequiredtobeoperable) toensureadequatemarginexisted.Completion ofanalysesdemonstrated thatoperation forCycle28wouldbewithindesignbasis.Therestriction previously imposedonlaketemperature waslifted.'B.ACTIONTAKENORPLANNEDTOPREVENTRECURRENCE: | |
| oAdditional analyseswereperformed tosupportcurrentandfuturecycleoperation.
| |
| TheMSLBanalysis(withtheassumedMFRVfailure)wasreanalyzed.
| |
| Significant assumption changestosupportoperation wereasfollows:Assumedshutdownmarginof2.40%versusthebasecasereference cyclevalueof1.80%.TheGinnaStationCoreOperating LimitsReport(COLR)wassubsequently revisedtoincorporate thisvalue. | |
| NRCFORM366hIBIBBBILICENSEEEVENTREPORT(LER)TEXTCONTINUATION U.S.NUCLEARREGULATORY COMMISSION FACILITYNAME(1)R.E.GinnaNuclearPowerPlantDOCKETIz)05000244LERNUMSER(6)BrppaRMLRNBIONIIUMBTRIIUMBER1999-001-01PAGEI3)8OF8TEXT(Ifmorespaceisrequired, useadditional copiesofNRCForm866AII17)Assumedoperation atthenormalGinnaStationaverageoperating temperature (Tavg)of561degreesFversustheupperendoftheanalyzedTavgwindowof573.5degreesF.Thisoperating restriction hasbeenplacedinGinnaStationoperating procedures andcorresponds tothefullpowerrodcontrolprogramsetpoint.
| |
| Plantmodifications torecoverthefullrangeofthepreviously analyzedTavgwindowandtorestorethepreviousshutdownmarginlimitswillbeconsidered inthefuture.NuclearEngineering Services(NES)hasimplemented aprocessforindependent reviewofvendorinputssuppliedbyRG&Eandisintheprocessofperforming anindependent reviewofallaccidentanalysisassumptions.
| |
| Westinghouse isevaluating theneedforinternalcorrective actionsaspartoftheirreviewofthiscondition. | |
| VI~ADDITIONAL INFORMATION: | | VI~ADDITIONAL INFORMATION: |
| A.FAILEDCOMPONENTS: | | A.FAILED COMPONENTS: |
| NoneB.PREVIOUSLERsONSIMILAREVENTS:AsimilarLERhistorical searchwasconducted withthefollowing results:Nodocumentation ofsimilarLEReventswiththesamerootcauseatGinnaStationcouldbeidentified.
| | None B.PREVIOUS LERs ON SIMILAR EVENTS: A similar LER historical search was conducted with the following results: No documentation of similar LER events with the same root cause at Ginna Station could be identified. |
| C.SPECIALCOMMENTS: | | C.SPECIAL COMMENTS: None}} |
| None}} | |
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Subject Vanes Were Dismantled & Correctly re-installed ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A3671998-07-14014 July 1998 LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr ML17265A1921998-03-11011 March 1998 LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells ML17265A1641998-02-0606 February 1998 LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified ML17265A1601998-02-0606 February 1998 LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status ML17264B1441997-12-17017 December 1997 LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr ML17264B1291997-12-0303 December 1997 LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr ML17264B1271997-12-0101 December 1997 LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset ML17264B1211997-11-24024 November 1997 LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage ML17264B0461997-09-29029 September 1997 LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr ML17264B0111997-08-27027 August 1997 LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr ML17264A9941997-08-19019 August 1997 LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr ML17264A9911997-08-11011 August 1997 LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe ML17264A8271997-03-0303 March 1997 LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr ML17264A8071997-01-22022 January 1997 LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr ML17264A7471996-11-27027 November 1996 LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr ML17264A6051996-09-19019 September 1996 LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr ML17264A6061996-09-19019 September 1996 LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr ML17264A5911996-09-0505 September 1996 LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr ML17264A5921996-09-0505 September 1996 LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr ML17264A5891996-08-22022 August 1996 LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr ML17264A5781996-08-0606 August 1996 LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr ML17264A5561996-07-12012 July 1996 LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr ML17264A5421996-06-20020 June 1996 LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr ML17264A5411996-06-17017 June 1996 LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr ML17264A5051996-05-17017 May 1996 LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed ML17264A4481996-04-0808 April 1996 LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr ML17264A4471996-04-0808 April 1996 LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr ML17264A4101996-03-18018 March 1996 LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr ML17264A2971995-12-14014 December 1995 LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr ML17264A1711995-09-25025 September 1995 LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr 1999-09-22
[Table view] Category:RO)
MONTHYEARML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A3671998-07-14014 July 1998 LER 98-002-00:on 971019,CR Emergency Air Treatment Sys Actuating Function Was Not Operable.Caused by Mispositioned Switch.Revised Procedure CPI-MON-R37.W/980714 Ltr ML17265A1921998-03-11011 March 1998 LER 98-001-00:on 980209,discovered That Boraflex Degradation in SPF Was Greater than Was Assumed.Caused by Dissolution of Boron on Boraflex Matrix,Per 10CFR50.21.Removed Spent Fuel Assemblies from Selected Degraded Storage Rack Cells ML17265A1641998-02-0606 February 1998 LER 97-007-01:on 971117,reactor Engineer Recognized That Neutron Flux Low Range Trip Circuitry for Channel Was Not in Tripped Condition as Required.Caused by Technical Inadequacies.Channel Defeat Will Be Identified ML17265A1601998-02-0606 February 1998 LER 97-006-01:on 971103,verification of B Concentration Was Not Performed Due to Misinterpretation of Event Sequence. Audible Count Rate Function Was Restored to Operable Status ML17264B1441997-12-17017 December 1997 LER 97-007-00:on 971117,NF Low Range Trip Circuitry for Channel N-44 Was Not Placed in Tripped Condition.Caused by Technical Inadequacies in Procedures.Implemented EWR 4862 to Resolve Design deficiency.W/971217 Ltr ML17264B1291997-12-0303 December 1997 LER 97-006-00:on 971103,NIS Audible Count Rate Function Was Inoperable.Caused by Misinterpretation of Event Sequence Due to Not Verifying Boron Concentration.B Verification Occurred Every 12 H Per ITS LCO Action 3.9.2.C.3.W/971203 Ltr ML17264B1271997-12-0101 December 1997 LER 97-005-00:on 971031,undetected Unblocking of SI Actuation Signal Occurred at Low Pressure Condition,Due to Faulty Bistable Which Resulted in Inadvertent SI Actuation Signal.Sias,Ci & CVI Signals Were Reset ML17264B1211997-11-24024 November 1997 LER 97-004-00:on 971024,radiation Monitor Alarm Were Noted Due to Higher than Normal Radioactive Gas Concentration Resulted in Cvi.New R-12 Alarm Setpoint Was Maintained for Duration of Refueling Outage ML17264B0461997-09-29029 September 1997 LER 97-003-01:on 970730,bistable Instrument Trip Setpoint Could Have Exceeded Allowable Value.Caused by Insufficient Existing Margin Between Trip Setpoint & Allowable Value. Held Switches in Tripped configuration.W/970929 Ltr ML17264B0111997-08-27027 August 1997 LER 97-003-00:on 970730,high Steam Flow Bistable Instrument Setpoint Plus Instrument Uncertainty Could Exceed Allowable Value in ITS Was Identified.Caused by Entry Into ITS LCO 3.0.3.Switches Placed in Tripped configuration.W/970827 Ltr ML17264A9941997-08-19019 August 1997 LER 97-002-00:on 970720,34.5 Kv Offsite Power Circuit 751 Was Lost.Caused by Automatic Actuation of B Emergency DG Due to Undervoltage on Safeguards Buses 16 & 17.Offsite Power Restored to Safeguards Buses 16 & 17.W/970819 Ltr ML17264A9911997-08-11011 August 1997 LER 96-009-02:on 960723,determined That Leak Rate Outside Containment Was Greater than Program Limit.Caused by Weld Defect.Isolated Leak & Cut Out & Replaced Leaking Pipe ML17264A8271997-03-0303 March 1997 LER 97-001-00:on 970131,discovered Service Water Temp Was Less than Specified Value.Caused by non-representative Method of Monitoring.Increased Water Temp in Screenhouse Bay to Greater than 35 Degrees F.W/970303 Ltr ML17264A8071997-01-22022 January 1997 LER 96-015-00:on 961223,discovered Thermally Induced Overpressure Transient Could Occur.Caused by Thermal Expansion of Fluid During Design Basis Accident Condition. Installed Relief Valve on Affected line.W/970122 Ltr ML17264A7471996-11-27027 November 1996 LER 96-013-00:on 961029,circuit Breakers Closed While in Mode 3 & Resulted in Condition Prohibited by TS Due to Personnel Error.Circuit Breakers for MOV-878B & MOV-878D Were re-opened.W/961127 Ltr ML17264A6051996-09-19019 September 1996 LER 96-012-00:on 960820,feedwater Transient Occurred,Due to Closure of Feedwater Regulating Valve,Causing Lo Lo Steam Generator Level Reactor Trip.Sgs Were Restored & Missing Screw in 1/P-476 Was replaced.W/960919 Ltr ML17264A6061996-09-19019 September 1996 LER 96-009-01:on 960723,leakage Outside Containment Occurred,Due to Weld Defect,Resulting in Leak Rate Greater than Program Limits.Source of Leakage Isolated from RWST by Freeze Seal,Allowing Exit from ITS LCO 3.0.3.W/960919 Ltr ML17264A5911996-09-0505 September 1996 LER 96-011-00:on 960807,improper Configuration of Circuit Breaker Occurred,Due to Undetected Internal Interference, Resulting in Automatic Start of Both Auxiliary Feedwater Pumps.Running AFW Pumps Were secured.W/960905 Ltr ML17264A5921996-09-0505 September 1996 LER 96-010-00:on 960806,latching of Main Turbine While in Mode 4 Occurred,Due to Defective Procedure,Resulting in Automatic Start of Auxiliary Feedwater Pump.Caused by Defective Maint Procedure.Procedure revised.W/960905 Ltr ML17264A5891996-08-22022 August 1996 LER 96-009-00:on 960723,determined Leak on Piping Sys Outside Containment Greater than Program Limit.Caused by Weld Defect.Pipe & Socket Welds Were Cut Out & Replaced. W/960822 Ltr ML17264A5781996-08-0606 August 1996 LER 96-008-00:on 960707,main Feedwater Pump Breakers Opened. Caused by Change in Seal Water Differential Pressure Occurred During Sys Realignment.Afw Flow Controlled as Desired to Maintain S/G level.W/960806 Ltr ML17264A5561996-07-12012 July 1996 LER 96-007-00:on 960612,CR Operators Identified Control Rods Misaligned & Not Moving in Proper Sequence.Caused by Faulty Firing Circuit Card in Rod Control Sys.Faulty Firing Circuit Card in 1BD Power Cabinet replaced.W/960712 Ltr ML17264A5421996-06-20020 June 1996 LER 96-006-00:on 960521,discovered Containment Penetration Not in Required Status.Caused by Personnel Error.Installed Flange Inside Containment Penetration 2.W/960620 Ltr ML17264A5411996-06-17017 June 1996 LER 96-005-00:on 960516,PORC Determined Deficient Procedures Do Not Meet SRs for Testing safety-related Logic Circuits. Caused by Inadequancies in Individual Testing Procedures. Procedures Re Improved TSs revised.W/960617 Ltr ML17264A5051996-05-17017 May 1996 LER 96-003-01:on 960308,identified That Both Pressurizer PORVs Inoperable Concurrently Due to Disconnection of Flex Hose to Both PORV Actuators to Install air-sets for Benchset & Limit Switch Activities.Hpes Completed ML17264A4481996-04-0808 April 1996 LER 96-003-00:on 960308,both Pressurizer Relief Valves Inoperable.Hpes Evaluation Is Being Conducted to Determined Cause of Event.C/As:Both PORVs restored.W/960408 Ltr ML17264A4471996-04-0808 April 1996 LER 96-002-00:on 960307,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump.C/As: Thermography performed.W/960408 Ltr ML17264A4101996-03-18018 March 1996 LER 96-001-00:on 950504,inservice Test Not Performed During Refueling Outage.Caused by Inadequate Tracking of Surveillance Frequency.Valve Test Performed & Disassembled. W/960318 Ltr ML17264A2971995-12-14014 December 1995 LER 95-009-00:on 950817,surveillance Was Not Performed Due to Improper Application of TS Requirements Resulting in TS Violation.Testing of MOV-515 Was Performed on 951115.W/ 951214 Ltr ML17264A1711995-09-25025 September 1995 LER 95-008-00:on 950825,secondary Transient Occurred.Caused by Loss of B Condenser Circulating Water Pump That Resulted in Manual Rt.Returned S/G Levels to Normal Operating levels.W/950925 Ltr 1999-09-22
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML17265A7601999-10-0505 October 1999 Part 21 Rept Re W2 Switch Supplied by W Drawn from Stock, Did Not Operate Properly After Being Installed on 990409. Switch Returned to W on 990514 for Evaluation & Root Cause Analysis ML17265A7621999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Re Ginna Npp.With 991008 Ltr ML17265A7531999-09-23023 September 1999 Part 21 Rept Re Corrective Action & Closeout of 10CFR21 Rept of Noncompliance Re Unacceptable Part for 30-4 Connector. Unacceptable Parts Removed from Stock & Scrapped ML17265A7541999-09-22022 September 1999 LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr ML17265A7471999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Re Ginna Npp.With 990909 Ltr ML17265A7431999-08-24024 August 1999 LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed ML17265A7341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Re Ginna Npp.With 990806 Ltr ML17265A7291999-07-29029 July 1999 Interim Part 21 Rept Re safety-related DB-25 Breaker Mechanism Procured from W Did Not Pas Degradatin Checks When Drawn from Stock to Be Installed Into BUS15/03A.Holes Did Not line-up & Tripper Pan Bent ML17265A7181999-07-23023 July 1999 LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians ML17265A7081999-07-22022 July 1999 LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored ML17265A7131999-07-22022 July 1999 Special Rept:On 990407,radiation Monitor RM-14A Was Declared Inoperable.Caused by Failed Communication Link from TSC to Plant Process Computer Sys.Communication Link Was re-established & RM-14A Was Declaed Operable on 990521 ML17265A7031999-07-19019 July 1999 LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr ML17265A7211999-07-19019 July 1999 ISI Rept for Third Interval (1990-1999) Third Period, Second Outage (1999) at Re Ginna Npp. ML17265A7021999-07-15015 July 1999 LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr ML17265A7661999-06-30030 June 1999 1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr ML17265A7011999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Re Ginna Npp.With 990712 Ltr ML17265A6851999-06-21021 June 1999 LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr ML17265A6761999-06-16016 June 1999 Part 21 Rept Re Defects & noncompliances,10CFR21(d)(3)(ii), Which Requires Written Notification to NRC on Identification of Defect or Failure to Comply. Relays Were Returned to Eaton for Evaluation & Root Cause Analysis ML17265A6661999-06-0202 June 1999 LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr ML17265A6681999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Re Ginna Nuclear Power Plant.With 990608 Ltr ML17265A6651999-05-27027 May 1999 Interim Rept Re W2 Control Switch,Procured from W,Did Not Operate Satisfactorily When Drawn from Stock to Be Installed in Main Control Board for 1C2 Safety Injection Pump. Estimated That Evaluation Will Be Completed by 991001 ML17309A6541999-05-27027 May 1999 LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr ML17265A6631999-05-24024 May 1999 LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved ML17265A6601999-05-21021 May 1999 LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr ML17265A6591999-05-17017 May 1999 Part 21 Rept Re Relay Deficiency Detected During pre-installation Testing.Caused by Incorrectly Wired Relay Coil.Relays Were Returned to Eaton Corp for Investigation. Relays Were Repaired & Retested ML17265A6441999-05-13013 May 1999 LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr ML17265A6431999-05-12012 May 1999 LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed ML17265A6381999-05-0707 May 1999 Part 21 Rept Re Replacement Turbocharger Exhaust Turbine Side Drain Port Not Functioning as Design Intended.Caused by Manufacturing Deficiency.Turbocharger Was Reaasembled & Reinstalled on B EDG ML17265A6391999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Re Ginna Nuclear Power Plant.With 990510 Ltr ML17265A6361999-04-23023 April 1999 Part 21 Rept Re Power Supply That Did Not Work Properly When Drawn from Stock & Installed in -25 Vdc Slot.Power Supply Will Be Sent to Vendor to Perform Failure Mode Assessment.Evaluation Will Be Completed by 991001 ML17265A6301999-04-18018 April 1999 Rev 1 to Cycle 28 COLR for Re Ginna Npp. ML17265A6251999-04-15015 April 1999 Special Rept:On 990309,halon Systems Were Removed from Svc & Fire Door F502 Was Blocked Open.Caused by Mods Being Made to CR Emergency Air Treatment Sys.Continuous Fire Watch Was Established with Backup Fire Suppression Equipment ML17265A6551999-04-0909 April 1999 Initial Part 21 Rept Re Mfg Deficiency in Replacement Turbocharger for B EDG Supplied by Coltec Industries. Deficiency Consisted of Missing Drain Port in Intermediate Casing.Required Oil Drain Port Machined Open ML17265A6291999-03-31031 March 1999 Rev 0 to Cycle 28 COLR for Re Ginna Npp. ML17265A6241999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Ginna Station.With 990409 Ltr ML17265A6141999-03-31031 March 1999 LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr ML17265A6131999-03-29029 March 1999 LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr ML17265A6061999-03-24024 March 1999 LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp ML17265A5661999-03-0101 March 1999 Rev 26 to QA Program for Station Operation. ML17265A5961999-02-28028 February 1999 Monthly Operating Rept for Feb 1999 for Ginna Nuclear Power Plant.With 990310 Ltr ML17265A5371999-01-31031 January 1999 Monthly Operating Rept for Jan 1999 for Re Ginna Nuclear Power Plant.With 990205 Ltr ML17265A5951998-12-31031 December 1998 Rg&E 1998 Annual Rept. ML17265A5001998-12-21021 December 1998 Rev 26 to QA Program for Station Operation. ML17265A4951998-12-21021 December 1998 LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr ML17265A4931998-12-17017 December 1998 LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery ML17265A4761998-11-30030 November 1998 Monthly Operating Rept for Nov 1998 for Re Ginna Nuclear Power Plant.With 981210 Ltr ML17265A4691998-11-25025 November 1998 LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR ML17265A4531998-10-31031 October 1998 Monthly Operating Rept for Oct 1998 for Re Ginna Nuclear Power Plant.With 981110 Ltr ML17265A4271998-10-0505 October 1998 LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr ML17265A4291998-09-30030 September 1998 Monthly Operating Rept for Sept 1998 for Re Ginna Nuclear Power Plant.With 981009 Ltr 1999-09-30
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p CATEGORY 1gj REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9906290303 DOC.DATE: 99/06/21 NOTARIZED:
NO FACIL:50-244,Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION ST MARTIN,J.T.
Rochester Gas&: Electric Corp.MECREDY,R.C.
Rochester Gas&, Electric Corp.RECIP.NAME RECIPIENT AFFILIATION DOCKET 05000244 VISSING,G.S.
SUBJECT:
LER 99-001-01:on 990222,deficiencies in NSSS vendor steam-line brake mass&energy release analysis results in plant being outside design bases occurred.Caused by deficiencies in W.Temporary administrative replaced.With 990621 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR I ENCL 1 SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
Q 05000244 0 RECIPIENT ID CODE/NAME LPD1-1 PD INTERNAL: AEOD/SPD/RRAB NRR/DIPM/IOLB NRR/DRIP/REXB RES/DET/EIB EXTERNAL: L ST LOBBY WARD NOAC POORE,W.NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME VISSING,G.
FILE CEDER/DUPE IQMB NRR/DSSA/SPLB RGN1 FILE 01 LMITCO MARSHALL NOAC QUEENER,DS NUDOCS FULL TXT COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 R D'E N NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMEN DESK (DCD)ON EXTENSION 415-2083 HE DOCUMENT CONTROL FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 16 ENCL 16 0
ROCHESTER GAS AND ELECTRIC CORR7RATION
+89 EAST AVENLJE, ROCHESTER, N.Y.14'-0001 AREA CODE 7I6 SrI6-2700 ROBERT C.MECREDY Vice President gvc!ear Operations June 21, 1999 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Guy S.Vissing Project Directorate I-1 Washington, D.C.20555 Subj ect: LER 1999-001, Revision 1, Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside its Design Basis R.E.Ginna Nuclear Power Plant Docket No.50-244
Dear Mr.Vissing:
The attached Licensee Event Report LER 1999-001, Revision 1, is submitted in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(ii)(B), which requires a report of,"Any event or condition...that resulted the nuclear power plant being...In a condition that was outside the design basis of the plant", and in accordance with 10 CFR 21,"Reporting of Defects and Noncompliances".
Very truly yours, Robert C.Mec edy xc: Mr.Guy S.Vissing (Mail Stop SC2)Project Directorate I-1 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation U.S.Nuclear Regulatory Commission Washington, D.C.20555 Regional Administrator, Region I U.S.Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U.S.NRC Ginna Senior Resident Inspector 0 9906290308 990b21 PDR ADOCK 05000244 S PDR
NRC FORM 366 IILI BBS)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)OPAL)WZ<~Is~>R;&0 A L:%N(P" informaaon cosectlon request: 50 hrs.Reported lessons learned are incorporated hto the licensing process and fed back to industry.Forward comments regarding burden estimate to the Records Management Branch (TA F33).U.S.Nudear Regulatory Commissbn.
Washington.
DC 205554001, and to the Paperwork Reduction Project (31~104).Ofrce of Management and Budget.Washington, DC 20503.If an information colection does not display a currently valid OMB control number.Lhe NRC may not conduct or sponsor.and a FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DocKET NUMBER I2)05000244 PAGE I3)1 OF 8 TITlE tel Deficiencies in NSSS Vendor Steamline Break Mass and Energy Release Analysis Results in Plant Being Outside its Design Basis EVENT DATE (5)MONTH OAY 02 22 LER NUMBER (6)SEQUENTIAL REYSION NUMBER NUMBER 1999 1999-001" 01 MONTH OAY 06 21 1999 REPORT DATE (7)FACIUTY NAME FACILITY NAME DOCKET NUMBER 05000 DOCKET NVMBER 05000 OTHER FACILITIES INVOI.VED (6)OPERATING MODE (B)POWER LEVEL (10)75 20.2201(b), 20.2203(a)(1) 20.2203(a)(2)(i) 20.2203(a)
(2)(ii)20.2203(a)
(2)(iii)20.2203(a)
(2)(iv)20.2203(a)
(2)(v)20.2203(a)(3)(i)
'0.2203(a)(3)(ii) 20.2203(a)
(4)50.36(c)(1)50.36(c)(2)LICENSEE CONTACT FOR THIS LER (12)50 73(a)(2)(i)(B) 50.73(a)(2)(ii)50.73(a)(2)(iii) 50.73(a)(2)(iv)~50.73(al(2)(v)50.73(a)(2)(vii)
TS OF 10 CFR 5: (Check THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMEN one or more)(11)50.73(a)(2)(viii]50.73(a)(2)(x) 73.71 X OTHER Part 21 Specify In Abstract below or in NRC Form 366A TELEPHONE NVMSER aootvde Aree Code)John T.St.Martin-Technical Assistant (716)771-3641 CAUSE COMPONENT MANUPACTURER REPORTABLE TO EPOS SYSTEM COMPONENT MANUPACTURER REPORTABLE TO EPIX SUPPLEMENTAL REPORT EXPECTED (I4)YES (If yes, complete EXPECTED SUBMISSION DATE).NO X EXPECTED SUBMISSION DATE (15)MONTH OAY TEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On February 22, 1999, Ginna Station was notified by its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a main steam)inc break inside containment with an assumed single failure of a main feedwater regulating valve.These errors had a nonconservative impact on calculated peak containment pressure.The first error involved the omission of a previously unaccounted for volume of high temperature feedwater.
The second error resulted in an incorrect feedwater isolation time.The consequences of these two modeling errors previously created the potential for containment to exceed its design basis value of 60 psig for a main steamline break with a concurrent failure of a main feedwater regulating valve for R.E.Ginna Cycles 12 through 25.The cause of the plant being in a condition that was outside the design basis of the plant from Cycle 12 to Cycle 25 was deficiencies in the Westinghouse main steam)ine break mass and energy release analysis.Immediate corrective action for Cycle 27 was to place a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature, a measure of Service Water/Containment'fan cooler heat removal and pressure reduction capability, for Modes 1 through 4 (when containment is required to be operable)to ensure adequate post-accident environmental conditions margin existed.Corrective action to justify return to power for Cyc)e 28 included a cycle-specific analysis showing peak containment pressure less than 60 psig.Corrective action to prevent recurrence is outlined in Section V.B.
NRC FORM 366A IB IBBB)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME I1)R.E.Ginna Nuclear Power Plant DOCKET Iz)05000244 LER NUMBER (6)BEOUENTNL REVISION NUMBER NUMBER 1999-001-01 PAGE I3)2 OF I)TEXT flf more speceis required, use eddirionel copies of NRC Form 366A)I17)PRE-EVENT PLANT CONDITIONS:
In 1992, the NSSS vendor for Ginna Station (Westinghouse) analyzed a full spectrum of cases for the main steam line break (MSLB)and containment response for the R.E.Ginna plant.The set of analyses was performed to support the program to reduce the boric acid storage tank (BAST)boron concentration to 2000 ppm.This analysis entailed over 120 cases, which varied break size and location, initial power level, single failure, offsite power availability, and different reactivity coefficients (rodded versus unrodded)~In 1995, Westinghouse re-visited the steamline break mass and energy release event due to replacement steam generators (RSGs)and an increase to 18 month fuel cycles.Many of the previous cases could be eliminated due to: An integral flow restrictor on the RSGs, eliminating any breaks larger than 1.4 square feet (the RSGs have flow restricting outlet nozzles which make any break size larger than the throat area of the restrictor not credible and thus, break flow areas greater than 1.4 square feet no longer need to be considered), and 2.Credit being taken for the main steam non-return check valve, which eliminated several types of breaks and the need to consider a main steam isolation valve (MSIV)single failure for containment overpressurization cases.After eliminating cases that were no longer credible, cases for the 1995 analyses were selected from the 1992 BAST reduction program.Six of the most limiting cases were selected, including different break sizes, initial power levels, single failures, and offsite power assumptions.
Another difference from the previous analyses is explained in the Ginna Station Updated Final Safety Analysis Report (UFSAR)Section 6.2.1.2.3: "In the Reference 3 analyses[1992 BAST reduction program), the most limiting feedwater control system failure was assumed to be a failure in the auxiliary feedwater system control.This failure resulted in increased auxiliary feedwater flow rates for the entire duration of the transient.
A failure of the isolation function of the main feedwater regulating valves (MFRVs)was not considered because a feedwater isolation signal also results in a main feedwater pump trip and in closure of the pump discharge valve (MFPDV).Thus, no single failure would have resulted in a failure to isolate main feedwater or a significant delay in the isolation of main feedwater to the affected steam generator.
It has subsequently been realized that the condensate pumps would continue to pump through the main feedwater pumps at low steam generator pressures and that the pump discharge valve (MFPDV)closure time is approximately 80 seconds.Therefore, a failure of the isolation function of the main feedwater regulating valve (MFRV)could result in main feedwater flow via the condensate pumps for approximately 80 seconds to the affected steam generator.
This new failure scenario is a more limiting failure than was assumed in the previous analyses due to the timing and magnitude of the increased flow." The requirement for the 80 second stroke time requirement of the MFPDV is addressed in the Ginna Station Improved Technical Specifications (ITS)Section 3.7.3.
NRC FORM 366A (B IBBS)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)DOCKET (2)LER NUMBER (6)SEmINTIAE AEVISIBH NUMB(A IIUMBEA PAGE (3)R.E.Ginna Nuclear Power Plant 05000244 1 ass ()()1 3 OF 8 TEXT fff more speceis required, use eddidonel copies of NRC Form 366A/(17)DESCRIPTION OF EVENT: DATES AND APPROXIMATE TIMES OF MAJOR OCCURRENCES:
February 22, 1999: Event date.February 22, 1999, 1540 EST: Discovery date and time.February 22, 1999, 1601 EST: NRC is notified of this condition per 10CFR50.72 (b)(1)(ii)(8).March 1, 1999, 2018 EST: Plant is shutdown in Mode 4 for the planned 1999 refueling outage.April 16, 1999: NSSS analysis shows MSLB containment pressure is less than 60 psig.B.EVENT: On February 22, 1999, in activities unrelated to plant conditions, Ginna Station was notified by its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a MSLB inside containment with an assumed single failure of a MFRV.(Refer to the letter RGE-99-207 from Mr.Steve M.Ira of the Westinghouse Electric Corporation to Mr.Peter Bamford, Rochester Gas and Electric Corporation (RGSE), dated February 22, 1999,
Subject:
Steamline Break Mass and Energy Release Analysis Nonconformance.)
These errors had a nonconservative impact on calculated peak containment pressure.The first error involves the volume of feedwater that exists between the specified isolation valve and the faulted steam generator.
When the MFRV is assumed to fail to perform its isolation function, then the isolation function is performed by the MFPDV.Because of the relative locations of these valves to the steam generator, the volume of water is greater between the MFPDV and the steam generator than between the MFRV and the steam generator.
This increased volume of high temperature feedwater was not accounted for in the analysis.The second error was inadvertently isolating feedwater flow at 15 seconds instead of at 85 seconds (80 seconds valve stroke time and 5 seconds for signal delay)as was intended.This meant that feedwater flow should have continued for 70 seconds longer than was assumed in the analysis.The consequences of these two modeling errors has the potential for containment to exceed its design basis value of 60 psig for a MSLB with an assumed concurrent failure of a MFRV.This is due to adding higher energy fluid to containment for 70 seconds longer than was assumed in the 1995 MSLB analysis of record.There was also the potential for the containment to exceed its design basis value as documented in the 1992 analyses, due to discovery of a more limiting single failure.
NRC FORM 366A (9 19991 LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET (2)05000244 LER NUMBER (6)BEUUENTIAL AENSION NUMBEA NUMBER 1999-001-01 PAGE (3)4 OF 8 TEXT llf more spaceis required, use additional copies of NRC Form 366AI (17)At the time of discovery and until the planned shutdown on March 1, 1999, the screenhouse bay temperature was maintained at approximately 39 degrees Fahrenheit, while lake temperature remained at approximately 35 degrees Fahrenheit.
Ginna Station placed a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature, a measure of Service Water System temperature, for Modes 1 through 4 (when containment is required to be operable)to ensure adequate margin existed.By maintaining this temperature less than 40 degrees Fahrenheit, increased containment fan cooler heat removal ability was ensured and this, combined with a high cycle-specific excess shutdown margin, was sufficient to ensure containment pressure would remain less than 60 psig for this postulated scenario.To support operation in Cycle 28 and beyond, the MSLB analysis (with the assumed MFRV failure)was reanalyzed.
Significant assumption changes to support operation were as follows: Assumed shutdown margin of 2.40%versus the base case reference cycle value of 1.80%.The Ginna Station Core Operating Limits Report (COLR)was subsequently revised to incorporate this value.P Assumed operation at the normal Ginna Station average operating temperature (Tavg)of 561 degrees F versus the upper end of the analyzed Tavg window of 573.5 degrees F.This operating restriction has been placed in Ginna Station operating procedures and corresponds to the full power rod control program setpoint.These changes, when combined with several minor modeling refinements, led to peak containment pressure under this scenario to be less than 60 psig.Accordingly, operations for all of Cycle 28 was justified.
This includes the removal of the lake temperature restriction previously imposed.C.INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT: None D.OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED: None E.METHOD OF DISCOVERY:
This event was disclosed to RGSE by the notification from its NSSS vendor (Westinghouse) of two modeling errors in the analysis for a MSLB inside containment with an assumed single failure of a MFRV.(Refer to letter RGE-99-207 from Mr.Steve M.Ira of the Westinghouse Electric Corporation to Mr.Peter Bamford, Rochester Gas and Electric Corporation (RG(t(E), dated February 22, 1999,
Subject:
Steamline Break Mass and Energy Release Analysis Nonconformance.)
NRC FORM 366A IS ISSSI LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME 11I DOCKET l2l LER NUMBER I6)PAGE I 3 I R.E.Ginna Nuclear Power Plant 05000244 SEGUENEME REVISIUN NUMBER NUMBER 1999-001-01 5 OF 8 TEXT llf more spaceis required, use addidonal copies of NRC Form 366Al 117)OPERATOR ACTION: After review of this Westinghouse correspondence, Reactor Engineering and Analysis personnel notified Operations supervision, who notified the Control Room operators.
The NRC Resident was notified at this time.At approximately 1540 EST on February 22, 1999, plant staff determined that a non-emergency one hour notification, per 10CFR50.72 (b)(1)(ii)(B), should be made to the NRC Operations Center.The Shift Supervisor made this notification at approximately 1601 EST on February 22, 1999.Operators maintained a restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature from the time of discovery until plant shutdown on March 1 to ensure adequate margin existed.Analyses for Cycle 28 have been completed.
These analyses demonstrate that there is adequate containment (CNMT)pressure margin for a MSLB for Cycle 28.The restriction previously imposed on lake temperature was lifted.G.SAFETY SYSTEM RESPONSES:
None III.CAUSE OF EVENT: A.IMMEDIATE CAUSE: The immediate cause ot the plant being in a condition that was outside the design basis of the plant was deficiencies in the Westinghouse MSLB mass and energy release analysis.B.INTERMEDIATE CAUSE: The intermediate cause of the deficiencies in the Westinghouse analysis was the analysis did not account for a greater volume of high temperature feedwater and used an incorrect feedwater isolation time.The additional mass and energy release from the MSLB are penalties in the containment pressure calculation.
C.ROOT CAUSE: The underlying cause of the error in the volume of feedwater assumption was insufficient knowledge by Westinghouse personnel of the specific configuration of the Ginna Station feedwater piping.The Underlying cause of the error in the assumed feedwater isolation time was human error in the use of the LOFTRAN code by Westinghouse personnel.
NRC FORM 366A IB 1999)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME{1)R.E.Ginna Nuclear Power Plant DOCKET (2)05000244 LER NUMBER I6)SEQUENTUL REVIION NUMBER NUMBER 1999-001-01 PAGE I3)6 OF 8 TEXT llf more space is required, use eddidonel copies of NRC Form 366A/I17)IV.ANALYSIS OF EVENT: This event is reportable in accordance with 10 CFR 21 and in accordance with 10 CFR 50.73, Licensee Event Report System, item (6)(2)(ii)(B), which requires a report of,"Any event or condition...that resulted in the nuclear power plant being...In a condition that was outside the design basis of the plant".The modeling errors in the Westinghouse MSLB mass and energy release analysis had the potential to pressurize the Ginna Station containment beyond its design pressure due to this postulated post-accident scenario.An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:
There were no operational or safety consequences or implications attributed to the deficiencies in the Westinghouse MSLB mass and energy release analysis because: The MSLB containment analysis is presented in Ginna Station UFSAR Section 6.2.1.2.3.
A revised analysis was performed (in March/April 1999)by offsetting the adverse effects of the additional mass and energy release through refinements in the existing analysis assumptions.
These improvements do not alter the total quantity of steam released to containment, but provide benefits in lowering the rate of the release and increasing the rate of contairiment heat removal.Applying these analysis improvements to the MFRV failure cases yields peak containment pressures that are less than the containment design pressure.The Westinghouse analysis, as of March 24, 1999, showed that the peak containment pressure design basis value of 60 psig would not be exceeded provided that service water (SW)temperature (SW supplies the containment recirculation fans)is limited to less than 45 degrees Fahrenheit.
Prior to entering Mode 4 for the planned 1999 refueling outage, the screenhouse bay temperature was being maintained at approximately 39 degrees Fahrenheit, while lake temperature was approximately 35 degrees Fahrenheit.
Thus, plant operation remained within its design basis since the time of discovery.
The current Westinghouse analysis, completed during the 1999 refueling outage, shows that the peak containment pressure design basis value of 60 psig was not exceeded for Cycle 27, nor will it be exceeded for Cycle 28 (the current operating cycle).Operation during future cycles are also justified, with restrictions similar to those imposed for Cycle 28.Any small increase in containment pressure beyond the 60 psig design basis value would not be expected to fail the containment.
In 1996, the containment was successfully tested to 72 psig during the Structural Integrity Test for the steam.generator replacement project.A steam line break does not result in severe radiological consequences at Ginna Station;thus full containment integrity is not considered required for this event (even though containment isolation is assumed to occur).
NRC FORM 366A (6 1998)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)DOCKET (2)LER NUMBER (6)PAGE (3)R.E.Ginna Nuclear Power Plant 05000244 SEI)ENTIAE REVSIOM HUMBER)(UMBER 1999-.001-01 7 OF B TEXT ill more speceis required, use addi tionel copies ol NRC Form 366Al (17)o For Ginna Station Cycles 1 (the original cycle in 1970)through 11, the MSLB containment pressure was not part of the Ginna Station licensing basis.For Cycles 12 through 25, containment pressure on a MSLB was part of the Ginna Station licensing basis and the potential existed to exceed 60 psig under the postulated accident scenario described above.Cycles 26 and 27 operated at the current Tavg of 561 degrees F, and would have had a peak containment pressure less than 60 psig under the postulated conditions.
Cycle 28 has also been shown be acceptable.
o At no time in the operation of Ginna Station has a main steam line break occurred.Based on the above, it can be concluded that there were no unreviewed safety questions, and that the public's health and safety was assured at all times.V.CORRECTIVE ACTION: A.ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS: For the operation of Ginna Station between the time of discovery and plant shutdown on March 1, 1999, the screenhouse bay temperature was maintained at approximately 39 degrees Fahrenheit, while lake temperature remained at approximately 35 degrees Fahrenheit.
Thus, plant operation remained within its design basis from the discovery date until the 1999 refueling outage.Ginna Station placed a temporary administrative restriction of 40 degrees Fahrenheit maximum on screenhouse bay temperature for Modes 1 through 4 (when containment is required to be operable)to ensure adequate margin existed.Completion of analyses demonstrated that operation for Cycle 28 would be within design basis.The restriction previously imposed on lake temperature was lifted.'B.ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:
o Additional analyses were performed to support current and future cycle operation.
The MSLB analysis (with the assumed MFRV failure)was reanalyzed.
Significant assumption changes to support operation were as follows: Assumed shutdown margin of 2.40%versus the base case reference cycle value of 1.80%.The Ginna Station Core Operating Limits Report (COLR)was subsequently revised to incorporate this value.
NRC FORM 366h IB IBBBI LICENSEE EVENT REPORT (LER)TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME (1)R.E.Ginna Nuclear Power Plant DOCKET Iz)05000244 LER NUMSER (6)BrppaRML RNBION IIUMBTR IIUMBER 1999-001-01 PAGE I3)8 OF 8 TEXT (If more spaceis required, use additional copies of NRC Form 866AI I17)Assumed operation at the normal Ginna Station average operating temperature (Tavg)of 561 degrees F versus the upper end of the analyzed Tavg window of 573.5 degrees F.This operating restriction has been placed in Ginna Station operating procedures and corresponds to the full power rod control program setpoint.Plant modifications to recover the full range of the previously analyzed Tavg window and to restore the previous shutdown margin limits will be considered in the future.Nuclear Engineering Services (NES)has implemented a process for independent review of vendor inputs supplied by RG&E and is in the process of performing an independent review of all accident analysis assumptions.
Westinghouse is evaluating the need for internal corrective actions as part of their review of this condition.
VI~ADDITIONAL INFORMATION:
A.FAILED COMPONENTS:
None B.PREVIOUS LERs ON SIMILAR EVENTS: A similar LER historical search was conducted with the following results: No documentation of similar LER events with the same root cause at Ginna Station could be identified.
C.SPECIAL COMMENTS: None