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| {{#Wiki_filter:ACCELERATED DISTjUBUTION DEMONSHRATION SYSTEMREGULATORY INFORMATION DISTRIBUTION SYSTEM(RIDS)ACCESSION NBR:9004130169 DOC.DATE: | | {{#Wiki_filter:ACCELERATED DISTjUBUTION DEMONSHRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9004130169 DOC.DATE: 90/03/28 NOTARIZED: |
| 90/03/28NOTARIZED: | | NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C. |
| NOFACIL:50-244 RobertEmmetGinnaNuclearPlant,Unit1,Rochester GAUTH.NAMEAUTHORAFFILIATION MECREDY,R.C.
| | Rochester Gas 6 Electric Corp.RECIP.NAME RECIPIENT AFFILIATXON JOHNSON,A.R. |
| Rochester Gas6ElectricCorp.RECIP.NAME RECIPIENT AFFILIATXON JOHNSON,A.R. | | .Project Directorate I-3 |
| .ProjectDirectorate I-3 | |
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| ==SUBJECT:== | | ==SUBJECT:== |
| ForwardsannualreptofECCSmodelrevsasapplicable tofacility,per 10CFR50.46.
| | Forwards annual rept of ECCS model revs as applicable to facility,per 10CFR50.46. |
| DISTRXBUTION CODE:A001DCOPIESRECEIVED:LTR ENCLSIZE:TITLE:ORSubmittal: | | DISTRXBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE: TITLE: OR Submittal: |
| GeneralDistribution NOTES:License Expdateinaccordance with10CFR2,2.109(9/19/72).
| | General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). |
| DOCKET0500024405000244.SRECIPIENT IDCODE/NAME PDl-3LAJOHNSON,A INTERNAL:
| | DOCKET 05000244 05000244.S RECIPIENT ID CODE/NAME PDl-3 LA JOHNSON,A INTERNAL: NRR/DET/ECMB 9H NRR/DST 8E2 NRR/DST/SXCB 7E NUDOCS-ABSTRACT OGC/HDS2 RES/DSIR/EIB EXTERNAL: LPDR NSIC COPIES LTTR ENCL 1 1 5 5 1 1 1 1 1 1 1 1 1 0 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PD1-3 PD NRR/DOEA/OTS B11 NRR/DST/SELB 8 D NRR/DST/SRXB 8E EMB REG LE 01 NRC PDR COPIES LTTR ENCL 1"1 1 1 1 1 1 1 1 0 1 1 1 1 A N D D Ps9~<rmo7 R A NOTE TO ALL"RIDS" RECIPIENTS: |
| NRR/DET/ECMB 9HNRR/DST8E2NRR/DST/SXCB 7ENUDOCS-ABSTRACT OGC/HDS2RES/DSIR/EIB EXTERNAL: | | PLEASE HELP US TO REDUCE WASTElCONTACT THE.DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM'DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDI TOTAL NUMBER OF COPIES REQUIRED: LTTR 21, ENCL 19 S A~fg 4 ,h g'l>>~)VI Al e~I PJ k C c f 1~)~<, iJ i/(*-P'(I Jl fl 0 i)l/IIIIIIIIIIIIII lillllllE/I i/ANO/j jIIIII/jjjjjIIII SZZZuI l(liltiiZikll IIIIIII/ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N Y.14649.0001 March 28, 1990 TELEP NON E AREA COOK 7ld 546-2700 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Mr.Allen R.Johnson PWR Project Directorate I-3 Washington, D.C.20555 |
| LPDRNSICCOPIESLTTRENCL11551111111110111111RECIPIENT IDCODE/NAME PD1-3PDNRR/DOEA/OTS B11NRR/DST/SELB 8DNRR/DST/SRXB 8EEMBREGLE01NRCPDRCOPIESLTTRENCL1"1111111101111ANDDPs9~<rmo7 RANOTETOALL"RIDS"RECIPIENTS:
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| PLEASEHELPUSTOREDUCEWASTElCONTACTTHE.DOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATE YOURNAMEFROM'DISTRIBUTION LISISFORDOCUMENTS YOUDON'TNEEDITOTALNUMBEROFCOPIESREQUIRED:
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| LTTR21,ENCL19S A~fg4,hg'l>>~)VIAle~IPJkCcf1~)~<,iJi/(*-P'(IJlfl 0i)l/IIIIIIIIIIIIII lillllllE/I i/ANO/jjIIIII/jjjjjIIII SZZZuIl(liltiiZikll IIIIIII/ROCHESTER GASANDELECTRICCORPORATION o89EASTAVENUE,ROCHESTER, NY.14649.0001 March28,1990TELEPNONEAREACOOK7ld546-2700U.S.NuclearRegulatory Commission DocumentControlDeskAttn:Mr.AllenR.JohnsonPWRProjectDirectorate I-3Washington, D.C.20555
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| ==Subject:== | | ==Subject:== |
| 10CFR50.46 AnnualReportECCSEvaluation ModelRevisions R.E.GinnaNuclearPowerPlantDocketNo.50-244 | | 10CFR50.46 Annual Report ECCS Evaluation Model Revisions R.E.Ginna Nuclear Power Plant Docket No.50-244 |
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| ==Reference:== | | ==Reference:== |
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| 1)NS-NRC-89-3463, "10CFR50.46 AnnualNotification for1989ofModifications intheWestinghouse ECCSEvaluation Models",LetterfromW.J.Johnson(Westinghouse) toT.E.Murley(NRC),Dated.October5,1989.2)NS-NRC-89-3464, "Correction ofErrorsandModifications totheNOTRUMPCodeintheWestinghouse SmallBreakLOCAECCSEvaluation ModelWhichArePotentially Significant", | | 1)NS-NRC-89-3463,"10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models", Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), Dated.October 5, 1989.2)NS-NRC-89-3464,"Correction of Errors and Modifications to the NOTRUMP Code in the Westinghouse Small Break LOCA ECCS Evaluation Model Which Are Potentially Significant", Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), Dated October 5, 1989. |
| LetterfromW.J.Johnson(Westinghouse) toT.E.Murley(NRC),DatedOctober5,1989.
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| ==DearMr.Johnson:== | | ==Dear Mr.Johnson:== |
| ThisletterprovidestheannualreportofEmergency CoreCoolingSystem(ECCS)modelrevisions astheyapplytoR.E.Ginna.ZnReferences 1)and2),Westinghouse ElectricCorporation providedinformation regarding modifications totheirECCSevaluation modelstoNRCStaff.References 1)and2)describethegenericeffectsofthemodelrevisions forbothlargeandsmallbreakLossOfCoolantAccidents (LOCA).Theattachment tothisletterprovidesinformation regarding theeffectsoftheECCSevaluation modelmodifications ontheGinnaUFSARChapter15.6.4LOCAanalysis.
| | This letter provides the annual report of Emergency Core Cooling System (ECCS)model revisions as they apply to R.E.Ginna.Zn References 1)and 2), Westinghouse Electric Corporation provided information regarding modifications to their ECCS evaluation models to NRC Staff.References 1)and 2)describe the generic effects of the model revisions for both large and small break Loss Of Coolant Accidents (LOCA).The attachment to this letter provides information regarding the effects of the ECCS evaluation model modifications on the Ginna UFSAR Chapter 15.6.4 LOCA analysis.Modifications to the model cause the large break LOCA Peak Clad Temperature (PCT)to increase by 2 F to 1889 F.9004i30i69 90032S PDR ADOCK 05000244 R PDC i p'gget5/44'/ |
| Modifications tothemodelcausethelargebreakLOCAPeakCladTemperature (PCT)toincreaseby2Fto1889F.9004i30i69 90032SPDRADOCK05000244RPDCip'gget5/44'/ | | Modifications to the model for small break LOCA do not affect calculated. |
| Modifications tothemodelforsmallbreakLOCAdonotaffectcalculated. | | peak clad temperature. |
| peakcladtemperature.
| | Very truly yours, Robert C.Mecredy Division Manager Nuclear Pr'oduction RWEi088 Enclosures xc: Mr.Allen R.Johnson (Mail Stop 14D1)Project Directorate I-3 Washington, D.C.20555 U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector ATT'O 10CFR50.46 ANNUAL PORT Effect of Westinghouse ECCS Evaluation Model Modifications on the LOCA Analysis Results Found in Chapter 15.6.4 of the R.E.Ginna (RG&E)Nuclear Power Plant Updated, Final Safety Analysis Report and.WCAP-11609 Containing the Steam Generator Tube Plu in Re ort for Ginna Nuclear Power Station The October 17, 1988 revision to 10CFR50.46 required applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory'ommission (NRC)of errors and changes in the ECCS Evaluation Models on an annual basis, when the errors and changes are not.significant. |
| Verytrulyyours,RobertC.MecredyDivisionManagerNuclearPr'oduction RWEi088Enclosures xc:Mr.AllenR.Johnson(MailStop14D1)ProjectDirectorate I-3Washington, D.C.20555U.S.NuclearRegulatory Commission RegionI475Allendale RoadKingofPrussia,PA19406GinnaSeniorResidentInspector ATT'O10CFR50.46ANNUALPORTEffectofWestinghouse ECCSEvaluation ModelModifications ontheLOCAAnalysisResultsFoundinChapter15.6.4oftheR.E.Ginna(RG&E)NuclearPowerPlantUpdated,FinalSafetyAnalysisReportand.WCAP-11609 Containing theSteamGenerator TubePluinReortforGinnaNuclearPowerStationTheOctober17,1988revisionto10CFR50.46 requiredapplicants andholdersofoperating licensesorconstruction permitstonotifytheNuclearRegulatory'ommission (NRC)oferrorsandchangesintheECCSEvaluation Modelsonanannualbasis,whentheerrorsandchangesarenot.significant.
| | Reference 1 defines a significant error or change as one which results in a calculated peak fuel cladding temperature dif f erent by more than 50'F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50'F.In References 2 and 3, information regarding modifications to the Westinghouse large break and small break LOCA ECCS Evaluation Models was submitted to the NRC.The following presents an assessment of the effect of the modifications to the Westinghouse ECCS Evaluation Models on the Loss-Of-Coolant Accident (LOCA)analyses results found in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna Nuclear Power Station", October 1987 and Chapter 15.6.4 of the R.E.Ginna (RG&E)Nuclear Power Plant Updated Final Safety Analysis Report.LARGE BREAK LOCA-EVALUATION MODEL CHANGES The large break LOCA analysis for R.E.Ginna (RG&E)was examined to assess the effect of the applicable modifications to the Westinghouse large break LOCA ECCS Evaluation Model on Peak Cladding Temperature (PCT)results reported.in WCAP-11609. |
| Reference 1definesasignificant errororchangeasonewhichresultsinacalculated peakfuelcladdingtemperature differentbymorethan50'Ffromthetemperature calculated forthelimitingtransient usingthelastacceptable model,orisacumulation ofchangesanderrorssuchthatthesumoftheabsolutemagnitudes oftherespective temperature changesisgreaterthan50'F.InReferences 2and3,information regarding modifications totheWestinghouse largebreakandsmallbreakLOCAECCSEvaluation Modelswassubmitted totheNRC.Thefollowing presentsanassessment oftheeffectofthemodifications totheWestinghouse ECCSEvaluation ModelsontheLoss-Of-Coolant Accident(LOCA)analysesresultsfoundinWCAP-11609, "SteamGenerator TubePlugging(SGTP)ReportforGinnaNuclearPowerStation", | | The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP)level of 12'-o.The large break analysis was subsequently reanalyzed and licensed for 15%tube plugging and.the results were documented in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna Nuclear Power Station", October 1987.The large break LOCA analysis results were calculated. |
| October1987andChapter15.6.4oftheR.E.Ginna(RG&E)NuclearPowerPlantUpdatedFinalSafetyAnalysisReport.LARGEBREAKLOCA-EVALUATION MODELCHANGESThelargebreakLOCAanalysisforR.E.Ginna(RG&E)wasexaminedtoassesstheeffectoftheapplicable modifications totheWestinghouse largebreakLOCAECCSEvaluation ModelonPeakCladdingTemperature (PCT)resultsreported.
| | using the 1981 version of the Westinghouse large break LOCA ECCS Evaluation Model which is documented in WCAP-9220-P-A (Reference 4).The current licensing basis analysis assumed.the following information important to the large break LOCA analysis NSSS power level-102-o of 1520 MWt Fuel Type-,14 X14 OFA Pellet Edge Configuration |
| inWCAP-11609.
| | -Chamfer Uniform Steam Generator Tube Plugging Level-15%i Nuclear Peaking Factors of 2.32 for the Total Peaking Factor and 1.66 for the Enthalpy Rise Peaking Factor. |
| TheLOCAsafetyanalysesreportedintheUFSARwereperformed forGinnawithauniformSteamGenerator TubePlugging(SGTP)levelof12'-o.Thelargebreakanalysiswassubsequently reanalyzed andlicensedfor15%tubepluggingand.theresultsweredocumented inWCAP-11609,"SteamGenerator TubePlugging(SGTP)ReportforGinnaNuclearPowerStation",
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| October1987.ThelargebreakLOCAanalysisresultswerecalculated.
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| usingthe1981versionoftheWestinghouse largebreakLOCAECCSEvaluation Modelwhichisdocumented inWCAP-9220-P-A (Reference 4).Thecurrentlicensing basisanalysisassumed.thefollowing information important tothelargebreakLOCAanalysisNSSSpowerlevel-102-oof1520MWtFuelType-,14X14OFAPelletEdgeConfiguration
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| -ChamferUniformSteamGenerator TubePluggingLevel-15%iNuclearPeakingFactorsof2.32fortheTotalPeakingFactorand1.66fortheEnthalpyRisePeakingFactor. | |
| ~~ | | ~~ |
| r~~orR.E.Ginna(R),thelimitingbreakreslkedfromthedoubleended.guillotine ruptureofthecoldlegpipingwithadischarge coefficient ofCD=0.4.Thecalculated peakcladdingtemperature was1887'Fincluding theapplicable penalties. | | r~~or R.E.Ginna (R), the limiting break reslked from the double ended.guillotine rupture of the cold leg piping with a discharge coefficient of CD=0.4.The calculated peak cladding temperature was 1887'F including the applicable penalties. |
| ThePCTvalueof1887'Fincludesa6'enaltytoaccountforanupperplenuminjection and.acorecrossflow penalty'of 10'F.Thefollowing modifications totheWestinghouse ECCSEvaluation Modelwereevaluated todetermine iftheywouldaffectthecurrentlicensing basis-largebreakLOCAanalysisresultsforR.E.Ginnafor15%SGTPfound.'nWCAP-11609.
| | The PCT value of 1887'F includes a 6'enalty to account for an upper plenum injection and.a core crossflow penalty'of 10'F.The following modifications to the Westinghouse ECCS Evaluation Model were evaluated to determine if they would affect the current licensing basis-large break LOCA analysis results for R.E.Ginna for 15%SGTP found.'n WCAP-11609. |
| 1981ECCSEvaluation Model:(NotMax-SILimited,)
| | 1981 ECCS Evaluation Model: (Not Max-SI Limited,)In the 1981 version of the Westinghouse ECCS Evaluation Model, a modification was made to delay downcomer overfilling. |
| Inthe1981versionoftheWestinghouse ECCSEvaluation Model,amodification wasmadetodelaydowncomer overfilling.
| | The delay corresponds to backfilling of the intact cold legs.Data from tests simulating cold leg injection during the post-large break LOCA reflood phase which have adequate safety injection flow to condense all of the-available steam flow show a significant amount of subcooled. |
| Thedelaycorresponds tobackfilling oftheintactcoldlegs.Datafromtestssimulating coldleginjection duringthepost-large breakLOCArefloodphasewhichhaveadequatesafetyinjection flowtocondenseallofthe-available steamflowshowasignificant amountofsubcooled.
| | liquid to be present in the cold leg pipe test section.This situation corresponds to the so-called maximum safety injection scenario of ECCS Evaluation Model analyses.The R.E.Ginna (RGGE)LOCA analysis performed with the Westinghouse 1981 large break LOCA ECCS Evaluation Model is not affected.by the WREFLOOD code modifications since the maximum safeguards safety injection flow assumption is not limiting.1981 ECCS Evaluation Model.: (Two-Loop Plants)In the 1981'version of the Westinghouse ECCS Evaluation Model, the pressurizer is modeled as being attached to the broken (faulted)loop in the SATAN-VI code for calculating large break blowdown behavior.Sensitivity studies were performed to determine if this pressurizer location in the noding scheme was the most limiting position.The results indicated that two-loop Westinghouse PWRs are sensitive'to the pressurizer nodal location and that in some cases modeling the pressurizer in the intact (non-faulted) loop resulted.in a slight increase in the calculated Peak Cladding Temperature (PCT).For a two loop plant, core cooling is provided by negative core flow and.the negative core flow period.lasts through most of the remaining blowdown period;The concern regarding pressurizer location relates to the negative core flow period which is crucial for core cooling in a two-loop plant.With the pressurizer on the broken loop, pressurizer flow is a large contributor to break flow (pump side), lessening the contribution from the upper plenum and leaving a large upper plenum inventory for negative core flow later in blowdown.For Ginna, a penalty of 2'F due to modeling the pressurizer in the intact (non-faulted) loop was assessed to be used in tracking margin to the 10CFR50.46 limit on the current licensing basis analysis for 15-o SGTP contained. |
| liquidtobepresentinthecoldlegpipetestsection.Thissituation corresponds totheso-called maximumsafetyinjection scenarioofECCSEvaluation Modelanalyses.
| | in WCAP-11609. |
| TheR.E.Ginna(RGGE)LOCAanalysisperformed withtheWestinghouse 1981largebreakLOCAECCSEvaluation Modelisnotaffected.
| | The PCT for the limiting C=0.4 break is 1887'F with applicable penalties, and thus adding the 2'F increase brings the PCT, to 1889'F which is well below the 10CFR50.46 limit. |
| bytheWREFLOODcodemodifications sincethemaximumsafeguards safetyinjection flowassumption isnotlimiting.
| | ks discussed abov modifications to the Nestgghouse large break LOCA ECCS Evaluation Model could affect the result by altering the~PCT.A.Analysis Calculated Result B.Modifications to Westinghouse ECCS Evaluation Model C.ECCS Evaluation Model Modifications Resultant PCT 1887'F 2oF 1889 F SMALL BREAK LOCA-EVALUATION MODEL CHANGES The small break.LOCA analysis for R.E.Ginna (RG&E)was also examined to assess the effect of the applicable modifications to the Westinghouse ECCS Evaluation Models on Peak Cladding Temperature (PCT)results reported in Chapter 15.6.4.1 of the UFSAR.The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP)level of 12%.The small break LOCA analysis was subsequently evaluated and licensed for a SGTP level increase from 12: to 15%.The small break LOCA event was not reanalyzed because the results are not sensitive to a SGTP level increase from 12%to 15:.The evaluation was documented, and transmitted in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna Nuclear Power Station", October 1987.The small break LOCA analysis results were calculated using the Westinghouse small break LOCA ECCS Evaluation Model documented in WCAP-8970 (Reference 12)which utilized the WFLASH computer code.For R.E.Ginna (RG&E), the limiting size small break resulted from a 6 inch equivalent diameter break in the cold leg.The calculated peak cladding temperature was 1092 F.The analysis assumed the following information important to the small break analyses: NSSS power level-102%of 1520 MWt Fuel Type-14 Z14 OFA Pellet Edge Configuration |
| 1981ECCSEvaluation Model.:(Two-Loop Plants)Inthe1981'version oftheWestinghouse ECCSEvaluation Model,thepressurizer ismodeledasbeingattachedtothebroken(faulted) loopintheSATAN-VIcodeforcalculating largebreakblowdownbehavior.
| | -Chamfer Uniform Steam Generator Tube Plugging Level-12-o Nuclear Peaking Factors of 2e32 for the Total Peaking Factor and 1.66 for the Enthalpy Rise Peaking Factor.600 GPM Total Auziliary Feedwater Flow.The following modification to the Westinghouse ECCS Evaluation Model were evaluated to determine if they would affect the current licensing basis small break LOCA analysis results for R.E.Ginna.WFLASH ECCS Evaluation Model Following the accident at Three Mile Island Unit 2, additional attention was focused on the small break LOCA and.Westinghouse submitted a report, WCAP-9600 (Reference 5), to the Nuclear Regulatory Commission (NRC)detailing the performance of the Westinghouse small break LOCA Evaluation Model which utilized the WFLASH computer code.In NUREG-0611 (Reference 6), the NRC staff questioned the validity of certain models in the WFLASH, computer |
| Sensitivity studieswereperformed todetermine ifthispressurizer locationinthenodingschemewasthemostlimitingposition. | | ~t1'V code and required censees to justify contin acceptance of the model.Section II.K.3.30 of NUREG-0737 (Reference 7), which clarified the NRC Post-TMI requirements regarding small break LOCA modeling, required that the licensees revise the small break LOCA ECCS models along the guidelines specified in NUREG-0611. |
| Theresultsindicated thattwo-loopWestinghouse PWRsaresensitive'to thepressurizer nodallocationandthatinsomecasesmodelingthepressurizer intheintact(non-faulted) loopresulted.
| | Following the issuance of NUREG-0737, Westinghouse and the Westinghouse Owners Group decided to develop the NOTRUMP (Reference 9)computer code for use in a new small break LOCA ECCS Evaluation Model (Reference 10).The NRC approved the use of NOTRUMP for small break LOCA ECCS analyses in May 1985.Since approval of the NOTRUMP small break LOCA ECCS Evaluation Model in 1985, the WFLASH computer code has not been maintained as.part of the Westinghouse ECCS Evaluation Model computer codes.In section II.K.3.31 of NUREG-0737, the NRC required that each licensee submit a new small break LOCA analysis using an NRC approved small break LOCA Evaluation Model.which satisfied the requirements of NUREG-0737 section II.K.3.30. |
| inaslightincreaseinthecalculated PeakCladdingTemperature (PCT).Foratwoloopplant,corecoolingisprovidedbynegativecoreflowand.thenegativecoreflowperiod.laststhroughmostoftheremaining blowdownperiod;Theconcernregarding pressurizer locationrelatestothenegativecoreflowperiodwhichiscrucialforcorecoolinginatwo-loopplant.Withthepressurizer onthebrokenloop,pressurizer flowisalargecontributor tobreakflow(pumpside),lessening thecontribution fromtheupperplenumandleavingalargeupperplenuminventory fornegativecoreflowlaterinblowdown.
| | NRC Generic Letter 83-35 (Reference 8)relaxed the requirements of Item II.K.3.31 by allowing a more generic response and providing a basis for retention of the existing small, break LOCA analyses.Provided that the previously existing model results were demonstrated to be conservative with respect to the new small break LOCA model approved under the requirements of NUREG-0737 II.K.3.30 (NOTRUMP), plant specific analyses using the new small break LOCA Evaluation Model would not be required.In WCAP-11145 (Reference 11), Westinghouse and the Westinghouse Owners Group demonstrated that the result's obtained from calculations with WFLASH were conservative relative to those obtained.with NOTRUMP.Compliance with Item II.K.3.31 of NUREG-0737 could be completed by referencing WCAP-11145 and supplying some plant specific information. |
| ForGinna,apenaltyof2'Fduetomodelingthepressurizer intheintact(non-faulted) loopwasassessedtobeusedintrackingmargintothe10CFR50.46 limitonthecurrentlicensing basisanalysisfor15-oSGTPcontained.
| | Westinghouse, therefore, has not been modifying, investigating, or evaluating proposed changes to the WFLASH.small break LOCA ECCS Evaluation Model.As discussed above, none of the modifications to the Westinghouse small break LOCA ECCS Evaluation Model would affect the small break LOCA analysis results by altering the PCT.A.Analysis Calculated Result B.Modifications to Westinghouse ECCS Evaluation Model C.ECCS Evaluation Model Modifications Resultant PCT 1092'F+0oF 1092'F CONCLUSION An evaluation of the effect of modifications to the Westinghouse ECCS Evaluation Model as reported in References 2 and 3 was performed for both the large break LOCA and small break LOCA analyses results found in WCAP-11609 and Chapter 15.6.4 of the R.E.Ginna Nuclear Power Plant Updated Final Safety Analysis Report.The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP)level of 12-o.The large break LOCA analysis was subsequently reanalyzed and licensed for an increase to 15: tube plugging.The small break LOCA analysis was evaluated and was not impacted by the increase in SGTP level to.The results of the ana es and evaluations for the increase in.SGTP level to 15-o were documented in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna.Nuclear Power Station".When the effects of the ECCS model changes were combined with the current plant analysis results, it was determined that compliance with the requirements of 10CFR50.46 would be maintained. |
| inWCAP-11609.
| | REFERENCES 2.3.4.5.6.7.8.9.10.12."Emergency Core Cooling Systems;Revisions to Acceptance Criteria", Federal Register, Vol.53, No.180, pp.35996-36005, dated September, 16, 1988.NS-NRC-89-3463,"10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models," Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), dated.October 5, 1989.NS-NRC-89-3464,"Correction of Errors.and Modifications to the NOTRUMP Code in the Westinghouse Small Break LOCA ECCS Evaluation Model Which Are Potentially Significant", Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), dated October 5, 1989.WCAP-9220-P-A, Revision 1 (Proprietary), WCAP-9221-A, Revision 1 (Non-Proprietary),"Westinghouse ECCS Evaluation Model 1981 Version", 1981, Eicheldinger, C.WCAP-10924-P-A (Proprietary), WCAP-12130-A (Non-Proprietary),"Westinghouse Large Break LOCA Best Estimate Methodology", Hochreiter, L.E., et.al., January 1987.Report on Small Break Accidents for Westinghouse Nuclear Steam Supply System", WCAP-9601 (Non-Proprietary), June 1979, WCAP-9600 (Proprietary), June 1979."Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants", NUREG-0611, January 1980."Clarification of TMI Action Plan Requirements", NUREG-0737, November 1980."Clarification of TMI Plan Item II.K.3.31,"NRC Generic Letter 83-85 from D.G.Eisenhut, November 2, 1983."NOTRUMP-A Nodal Transient Small Break and General Network Code", WCAP-10079-P-A (Proprietary), WCAP-10081-A (Non-'Proprietary), Lee, N., et.al., August 1985."Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), Lee, N., et.al., August 1985.WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary),"Westinghouse Emergency Core Cooling System Small Break October 1975 Model", April 1977.}} |
| ThePCTforthelimitingC=0.4breakis1887'Fwithapplicable penalties, andthusaddingthe2'FincreasebringsthePCT,to1889'Fwhichiswellbelowthe10CFR50.46 limit.
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| ksdiscussed abovmodifications totheNestgghouse largebreakLOCAECCSEvaluation Modelcouldaffecttheresultbyalteringthe~PCT.A.AnalysisCalculated ResultB.Modifications toWestinghouse ECCSEvaluation ModelC.ECCSEvaluation ModelModifications Resultant PCT1887'F2oF1889FSMALLBREAKLOCA-EVALUATION MODELCHANGESThesmallbreak.LOCAanalysisforR.E.Ginna(RG&E)wasalsoexaminedtoassesstheeffectoftheapplicable modifications totheWestinghouse ECCSEvaluation ModelsonPeakCladdingTemperature (PCT)resultsreportedinChapter15.6.4.1oftheUFSAR.TheLOCAsafetyanalysesreportedintheUFSARwereperformed forGinnawithauniformSteamGenerator TubePlugging(SGTP)levelof12%.ThesmallbreakLOCAanalysiswassubsequently evaluated andlicensedforaSGTPlevelincreasefrom12:to15%.ThesmallbreakLOCAeventwasnotreanalyzed becausetheresultsarenotsensitive toaSGTPlevelincreasefrom12%to15:.Theevaluation wasdocumented, andtransmitted inWCAP-11609, "SteamGenerator TubePlugging(SGTP)ReportforGinnaNuclearPowerStation",
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| October1987.ThesmallbreakLOCAanalysisresultswerecalculated usingtheWestinghouse smallbreakLOCAECCSEvaluation Modeldocumented inWCAP-8970 (Reference 12)whichutilizedtheWFLASHcomputercode.ForR.E.Ginna(RG&E),thelimitingsizesmallbreakresultedfroma6inchequivalent diameterbreakinthecoldleg.Thecalculated peakcladdingtemperature was1092F.Theanalysisassumedthefollowing information important tothesmallbreakanalyses:
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| NSSSpowerlevel-102%of1520MWtFuelType-14Z14OFAPelletEdgeConfiguration
| |
| -ChamferUniformSteamGenerator TubePluggingLevel-12-oNuclearPeakingFactorsof2e32fortheTotalPeakingFactorand1.66fortheEnthalpyRisePeakingFactor.600GPMTotalAuziliary Feedwater Flow.Thefollowing modification totheWestinghouse ECCSEvaluation Modelwereevaluated todetermine iftheywouldaffectthecurrentlicensing basissmallbreakLOCAanalysisresultsforR.E.Ginna.WFLASHECCSEvaluation ModelFollowing theaccidentatThreeMileIslandUnit2,additional attention wasfocusedonthesmallbreakLOCAand.Westinghouse submitted areport,WCAP-9600 (Reference 5),totheNuclearRegulatory Commission (NRC)detailing theperformance oftheWestinghouse smallbreakLOCAEvaluation ModelwhichutilizedtheWFLASHcomputercode.InNUREG-0611 (Reference 6),theNRCstaffquestioned thevalidityofcertainmodelsintheWFLASH,computer | |
| ~t1'Vcodeandrequiredcenseestojustifycontinacceptance ofthemodel.SectionII.K.3.30 ofNUREG-0737 (Reference 7),whichclarified theNRCPost-TMIrequirements regarding smallbreakLOCAmodeling, requiredthatthelicensees revisethesmallbreakLOCAECCSmodelsalongtheguidelines specified inNUREG-0611. | |
| Following theissuanceofNUREG-0737, Westinghouse andtheWestinghouse OwnersGroupdecidedtodeveloptheNOTRUMP(Reference 9)computercodeforuseinanewsmallbreakLOCAECCSEvaluation Model(Reference 10).TheNRCapprovedtheuseofNOTRUMPforsmallbreakLOCAECCSanalysesinMay1985.SinceapprovaloftheNOTRUMPsmallbreakLOCAECCSEvaluation Modelin1985,theWFLASHcomputercodehasnotbeenmaintained as.partoftheWestinghouse ECCSEvaluation Modelcomputercodes.InsectionII.K.3.31 ofNUREG-0737, theNRCrequiredthateachlicenseesubmitanewsmallbreakLOCAanalysisusinganNRCapprovedsmallbreakLOCAEvaluation Model.whichsatisfied therequirements ofNUREG-0737 sectionII.K.3.30. | |
| NRCGenericLetter83-35(Reference 8)relaxedtherequirements ofItemII.K.3.31 byallowingamoregenericresponseandproviding abasisforretention oftheexistingsmall,breakLOCAanalyses.
| |
| Providedthatthepreviously existingmodelresultsweredemonstrated tobeconservative withrespecttothenewsmallbreakLOCAmodelapprovedundertherequirements ofNUREG-0737 II.K.3.30 (NOTRUMP),
| |
| plantspecificanalysesusingthenewsmallbreakLOCAEvaluation Modelwouldnotberequired.
| |
| InWCAP-11145 (Reference 11),Westinghouse andtheWestinghouse OwnersGroupdemonstrated thattheresult'sobtainedfromcalculations withWFLASHwereconservative relativetothoseobtained.
| |
| withNOTRUMP.Compliance withItemII.K.3.31 ofNUREG-0737 couldbecompleted byreferencing WCAP-11145 andsupplying someplantspecificinformation.
| |
| Westinghouse, therefore, hasnotbeenmodifying, investigating, orevaluating proposedchangestotheWFLASH.smallbreakLOCAECCSEvaluation Model.Asdiscussed above,noneofthemodifications totheWestinghouse smallbreakLOCAECCSEvaluation ModelwouldaffectthesmallbreakLOCAanalysisresultsbyalteringthePCT.A.AnalysisCalculated ResultB.Modifications toWestinghouse ECCSEvaluation ModelC.ECCSEvaluation ModelModifications Resultant PCT1092'F+0oF1092'FCONCLUSION Anevaluation oftheeffectofmodifications totheWestinghouse ECCSEvaluation ModelasreportedinReferences 2and3wasperformed forboththelargebreakLOCAandsmallbreakLOCAanalysesresultsfoundinWCAP-11609 andChapter15.6.4oftheR.E.GinnaNuclearPowerPlantUpdatedFinalSafetyAnalysisReport.TheLOCAsafetyanalysesreportedintheUFSARwereperformed forGinnawithauniformSteamGenerator TubePlugging(SGTP)levelof12-o.ThelargebreakLOCAanalysiswassubsequently reanalyzed andlicensedforanincreaseto15:tubeplugging. | |
| ThesmallbreakLOCAanalysiswasevaluated andwasnotimpactedbytheincrease inSGTPlevelto.Theresultsoftheanaesandevaluations fortheincreasein.SGTPlevelto15-oweredocumented inWCAP-11609,"SteamGenerator TubePlugging(SGTP)ReportforGinna.NuclearPowerStation".
| |
| WhentheeffectsoftheECCSmodelchangeswerecombinedwiththecurrentplantanalysisresults,itwasdetermined thatcompliance withtherequirements of10CFR50.46 wouldbemaintained.
| |
| REFERENCES 2.3.4.5.6.7.8.9.10.12."Emergency CoreCoolingSystems;Revisions toAcceptance Criteria", | |
| FederalRegister, Vol.53,No.180,pp.35996-36005,datedSeptember, 16,1988.NS-NRC-89-3463, "10CFR50.46 AnnualNotification for1989ofModifications intheWestinghouse ECCSEvaluation Models,"LetterfromW.J.Johnson(Westinghouse) toT.E.Murley(NRC),dated.October5,1989.NS-NRC-89-3464, "Correction ofErrors.and Modifications totheNOTRUMPCodeintheWestinghouse SmallBreakLOCAECCSEvaluation ModelWhichArePotentially Significant",
| |
| LetterfromW.J.Johnson(Westinghouse) toT.E.Murley(NRC),datedOctober5,1989.WCAP-9220-P-A, Revision1(Proprietary),
| |
| WCAP-9221-A, Revision1(Non-Proprietary), | |
| "Westinghouse ECCSEvaluation Model1981Version", | |
| 1981,Eicheldinger, C.WCAP-10924-P-A (Proprietary), | |
| WCAP-12130-A (Non-Proprietary), | |
| "Westinghouse LargeBreakLOCABestEstimateMethodology", | |
| Hochreiter, L.E.,et.al.,January1987.ReportonSmallBreakAccidents forWestinghouse NuclearSteamSupplySystem",WCAP-9601 (Non-Proprietary), | |
| June1979,WCAP-9600(Proprietary),
| |
| June1979."GenericEvaluation ofFeedwater Transients andSmallBreakLoss-of-Coolant Accidents inWestinghouse DesignedOperating Plants",NUREG-0611, January1980."Clarification ofTMIActionPlanRequirements",
| |
| NUREG-0737, November1980."Clarification ofTMIPlanItemII.K.3.31, "NRCGenericLetter83-85fromD.G.Eisenhut, November2,1983."NOTRUMP-ANodalTransient SmallBreakandGeneralNetworkCode",WCAP-10079-P-A (Proprietary), | |
| WCAP-10081-A (Non-'Proprietary), | |
| Lee,N.,et.al.,August1985."Westinghouse SmallBreakECCSEvaluation ModelUsingtheNOTRUMPCode",WCAP-10054-P-A (Proprietary), | |
| WCAP-10081-A (Non-Proprietary), | |
| Lee,N.,et.al.,August1985.WCAP-8970 (Proprietary) andWCAP-8971 (Non-Proprietary), | |
| "Westinghouse Emergency CoreCoolingSystemSmallBreakOctober1975Model",April1977.}} | |
Similar Documents at Ginna |
|---|
Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999
Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated
IR 05000244/19990081999-10-14014 October 1999
Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy
ML17265A7651999-10-0808 October 1999
Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request
ML17265A7631999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000
ML17265A7641999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel
ML20212J3561999-09-30030 September 1999
Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21
ML20212J3801999-09-30030 September 1999
Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21
IR 05000244/19990051999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7571999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
IR 05000244/19992011999-09-24024 September 1999
Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld
ML17265A7461999-08-31031 August 1999
Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726
ML17265A7401999-08-26026 August 1999
Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components
ML17265A7411999-08-26026 August 1999
Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded
ML17265A7451999-08-23023 August 1999
Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d)
ML17250B3021999-08-23023 August 1999
Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806
ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
1999-09-30
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999
Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated
ML17265A7651999-10-0808 October 1999
Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request
ML17265A7631999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000
ML17265A7641999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel
ML20212J3561999-09-30030 September 1999
Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21
ML20212J3801999-09-30030 September 1999
Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21
IR 05000244/19990051999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7571999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7461999-08-31031 August 1999
Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726
ML17265A7401999-08-26026 August 1999
Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components
ML17265A7411999-08-26026 August 1999
Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded
ML17250B3021999-08-23023 August 1999
Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806
ML17265A7451999-08-23023 August 1999
Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d)
ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
ML17265A5111999-01-11011 January 1999
Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME Bp&V Code for ISI Program.Relief Requests 37,38 & 39 Encl
ML17265A5101999-01-11011 January 1999
Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl
1999-09-30
[Table view]
Category:UTILITY TO NRC
MONTHYEARML20059L4481990-09-18018 September 1990
Forwards Responses to Rer 900709-13 Team Visit Findings, Per .Responses Withheld (Ref 10CFR73.21)
ML17309A4491990-09-13013 September 1990
Forwards Slides Presented by Facility & Westinghouse to NRC in 900724 Meeting.Encl Withheld
ML17262A1331990-09-11011 September 1990
Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion
ML17261B1511990-08-29029 August 1990
Forwards Semiannual Radioactive Effluent Release Rept Jan- June 1990 & Rev 3 to Process Control Program for Ginna Station, Per Tech Specs 6.9.1.4 & 6.16,respectively
ML17261B1481990-08-28028 August 1990
Lists Understanding of Issues Util Planning to Address Re Containment Integrity,Per 900718 Telcon.Any Concerns or Action Items Different from Listed Submittal Should Be Provided to Util Prior to NRC 900905 & 06 Visit to Plant
IR 05000244/19880261990-08-20020 August 1990
Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented
ML17261B1501990-08-20020 August 1990
Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented
ML17261B1371990-08-17017 August 1990
Forwards Rev 0 to, Inservice Insp Rept for Third Interval (1990-1999),First Period,First Outage (1990) at Re Ginna Nuclear Power Plant.
ML17309A4481990-08-16016 August 1990
Responds to NRC 900717 Ltr Re Violations Noted in Insp Rept 50-244/90-80.Corrective Actions:Westinghouse Drawing E-2508 Approved & Issued to Central Records
ML17261B1351990-08-15015 August 1990
Provides Info & Assessment on Integrity of Connection of Containment Bldg to Foundation,Per 900808 Telcon.Util Believes That Existing Condition Not Safety Concern & No Reason Exists to Suspect Joint Will Not Perform Function
ML17261B1361990-08-14014 August 1990
Responds to Commitment Tracking Concerns Noted in Insp Rept 50-244/90-09 & Planned Corrective Actions.Util Confirms Commitments Dates for Implementation of Effective Shelf Life Program & Comprehensive Preventive Maint Program for Parts
ML17261B1331990-08-13013 August 1990
Recommits to Performing Enhanced primary-to-secondary Leak Rate Monitoring,Per NRC Bulletin 88-002
ML17261B1261990-07-30030 July 1990
Clarifies Commitment Made in 900316 RO Re Restoring Inoperable Fire Damper I-411-21-P.Util Plans to Design Removable Track Which Will Allow Charcoal Drawer to Be Manipulated.Definitive Schedule Will Be Provided in 60 Days
ML20055H9291990-07-23023 July 1990
Forwards Revised Page 6:4 of Plant Security Plan.Page Withheld Per 10CFR73.21
ML17261B1091990-07-20020 July 1990
Advises That Structural Evaluations of Containment Sys Being Performed in Response to Containment Integrity Insp
ML17309A4471990-07-17017 July 1990
Forwards Decommissioning Rept, for Plant,Per 10CFR50.33(k) & 50.75(b)
ML17261B0881990-07-13013 July 1990
Provides Update to Util 860616 Ltr Re Implementation of NUREG-0737,Item 6.2,Suppl 1, Emergency Response Capability.
ML17261B0921990-07-11011 July 1990
Responds to 900709 Request for Addl Info Re Containment Integrity Insp.Util Will Provide Preliminary Results by 900716 Re Where Groundwater Entering Annular Access Area. Meeting Proposed for Wk of 900723
ML20044B0481990-07-10010 July 1990
Discusses Review of Station Blackout Documentation,Per 10CFR50.63.Licensee Will Complete Enhancements to Station Blackout Documentation Identified in Attachment 1 as Indicated & Other Items Will Be Completed within 2 Yrs
ML17309A4461990-07-0909 July 1990
Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions
ML17261B0931990-07-0909 July 1990
Responds to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Util Completed All Actions Requested in Generic Ltr & Will Retain Documentation Verification for Min of 2 Yrs
ML17261B0901990-07-0909 July 1990
Advises That Licensee Will Install Containment Isolation Signal Going to Valve AOV 745 by End of 1992 Refueling Outage,Per Util 900608 Ltr Notifying of Condition Outside Design Basis of Plant Under 10CFR50.72 Criterion
ML17250B2151990-06-29029 June 1990
Forwards Application for Amend to License DPR-18, Reformatting Auxiliary Electrical Sys Tech Specs & Action Statements for Offsite & Onsite Power Sources Available for Plant Auxiliaries
ML17261B0851990-06-28028 June 1990
Forwards LER 89-016-02 Re 891117 Failure of Safety Injection Block/Unblock Switch Which Could Render Both Trains of Safety Injection Sys Inoperable.Also Reported Per Part 21
ML17250B1991990-06-0808 June 1990
Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983
ML17250B1951990-06-0505 June 1990
Responds to Generic Ltr 89-08, Erosion/Corrosion Induced Pipe Wall Thinning. Util Developed Erosion/Corrosion Program for Single & Two Phase Sys Consistent W/Requirements of NUREG-1344 & NUMARC 870611 Rept
ML17250B1851990-06-0101 June 1990
Forwards Application for Amend to License DPR-18,providing Guidance for Action Statements Associated W/Power Distribution Limit Specs
ML17261B0691990-06-0101 June 1990
Discusses Testing Frequency for Insp of Incore Neutron Monitoring Sys Thimble Tubes,Per NRC Bulletin 88-009.Thimble Tube Indicating Greatest Wear Recently Repositioned in Effort to Minimize Future Wear
ML17250B1811990-05-31031 May 1990
Forwards Addl Info Re Response to Notice of Violation from Insp Rept 50-244/90-04.Info Withheld (Ref 10CFR73.21)
ML17250B1801990-05-29029 May 1990
Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Util Does Not Need to Develop Enhanced Surveillance Program to Monitor Currently Installed Transmitter Based Upon Limited Installed Quantity
ML17250B1841990-05-29029 May 1990
Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (SS) Internal Pre-Load Bolting.... Valves Disassembled & 410 SS Studs Removed & Visually & Liquid Penetrant Examined
ML17261B0961990-05-29029 May 1990
Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-244/90-07.Corrective Actions:Worker Records Immediately Corrected Indicating That Worker Received No Significant Exposure for Time While Error Occurred
ML17250B1721990-05-18018 May 1990
Forwards Rev 1 to Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. Rept Summarizes Observations & Corrective Actions Resulting from Insps Performed During 1990 Outage
ML20042G8931990-05-0808 May 1990
Forwards Revised Emergency Operating Procedures,Including Rev 3 to AP-RHR.2,Rev 1 to ES-0.3 & Rev 11 to ES-1.3
ML17261B0611990-04-26026 April 1990
Advises of Changes to 890201 Commitments Made Under Programmed Enhancement Response to Generic Ltr 88-17.Changes Will Not Reduce Capability to Operate Safely in Reduced Inventory Condition or Scope of Programmed Enhancements
ML17261B0591990-04-17017 April 1990
Advises That No Interlocks Required for RHR motor-operated Valves 701 & 720 Based on Present Arrangement.Listed Failures Would Have to Occur in Order for Potential Overpressurization of RHR Sys to Occur
ML17261B0461990-04-12012 April 1990
Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys
ML17262A1431990-04-12012 April 1990
Responds to NRC 900202 Ltr Re Weaknesses Noted in Insp Rept 50-244/89-80.Corrective Actions:Procedure ES-0.3 Modified to Provide Guidance for Rapid Cooldown & Depressurization W/ & W/O Reactor & Vessel Instrumentation Sys
ML17250B1451990-04-0606 April 1990
Discusses Impact of SER Issuance for Inservice Insp & Inservice Testing Programs & Advises That Timing Will Not Affect Util Implementation Plans for Programs,Except as Listed
ML17250B1441990-04-0505 April 1990
Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-244/90-02.Corrective Actions:Positive Indicator Will Be Installed for All Check Valves to Enhance Visible Positive Position Verification Ability & to Avoid Confusion
ML17261B0401990-03-30030 March 1990
Advises That Final Results of Station Blackout Documentation Review Will Be Submitted to NRC on or About 900501
ML17261B0321990-03-28028 March 1990
Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp
ML17261B0331990-03-28028 March 1990
Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date
ML17261B0231990-03-26026 March 1990
Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated
ML17261B0281990-03-23023 March 1990
Forwards Addl Info Re Proposed Tech Spec Amend Concerning Use of Reconstituted Fuel,Per 900315 Telcon.During Fuel Assembly Reconstitution,Failed Fuel Rods Will Be Placed W/ Filler Rods
ML17261B0221990-03-22022 March 1990
Provides Revised Test Schedule for motor-operated Valve Diagnostic Test Program,Per IE Bulletin 85-003.NRC Notification of Changes to Valve Operability Program Required by Generic Ltr 89-10,dtd 890828
ML17261B0411990-03-20020 March 1990
Forwards Summary of Onsite Property Damage Coverage Currently in Force at Plant,Per 10CFR50.54(w)(4)
ML17261B0151990-03-19019 March 1990
Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Plants (USI A-47). Overfill Protection Provided Through Trip Bistables in Reactor Protection Racks Powered from 120-volt Instrument Buses
ML17261B0101990-03-15015 March 1990
Forwards Application for Amend to License DPR-18,allowing Use of Reconstituted Fuel Assemblies
ML20012D2831990-03-14014 March 1990
Responds to Issues Discussed During 900307 Telcon W/Util & Eg&G Re Inservice Testing Program Status & Relief Request. Valves 5960A & 5960B Will Be Disassembled to Verify Forward Flow.Relief Requests PR-8 & VR-25 Encl
1990-09-18
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ACCELERATED DISTjUBUTION DEMONSHRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9004130169 DOC.DATE: 90/03/28 NOTARIZED:
NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.
Rochester Gas 6 Electric Corp.RECIP.NAME RECIPIENT AFFILIATXON JOHNSON,A.R.
.Project Directorate I-3
SUBJECT:
Forwards annual rept of ECCS model revs as applicable to facility,per 10CFR50.46.
DISTRXBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE: TITLE: OR Submittal:
General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
DOCKET 05000244 05000244.S RECIPIENT ID CODE/NAME PDl-3 LA JOHNSON,A INTERNAL: NRR/DET/ECMB 9H NRR/DST 8E2 NRR/DST/SXCB 7E NUDOCS-ABSTRACT OGC/HDS2 RES/DSIR/EIB EXTERNAL: LPDR NSIC COPIES LTTR ENCL 1 1 5 5 1 1 1 1 1 1 1 1 1 0 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PD1-3 PD NRR/DOEA/OTS B11 NRR/DST/SELB 8 D NRR/DST/SRXB 8E EMB REG LE 01 NRC PDR COPIES LTTR ENCL 1"1 1 1 1 1 1 1 1 0 1 1 1 1 A N D D Ps9~<rmo7 R A NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTElCONTACT THE.DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM'DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDI TOTAL NUMBER OF COPIES REQUIRED: LTTR 21, ENCL 19 S A~fg 4 ,h g'l>>~)VI Al e~I PJ k C c f 1~)~<, iJ i/(*-P'(I Jl fl 0 i)l/IIIIIIIIIIIIII lillllllE/I i/ANO/j jIIIII/jjjjjIIII SZZZuI l(liltiiZikll IIIIIII/ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N Y.14649.0001 March 28, 1990 TELEP NON E AREA COOK 7ld 546-2700 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Mr.Allen R.Johnson PWR Project Directorate I-3 Washington, D.C.20555
Subject:
10CFR50.46 Annual Report ECCS Evaluation Model Revisions R.E.Ginna Nuclear Power Plant Docket No.50-244
Reference:
1)NS-NRC-89-3463,"10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models", Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), Dated.October 5, 1989.2)NS-NRC-89-3464,"Correction of Errors and Modifications to the NOTRUMP Code in the Westinghouse Small Break LOCA ECCS Evaluation Model Which Are Potentially Significant", Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), Dated October 5, 1989.
Dear Mr.Johnson:
This letter provides the annual report of Emergency Core Cooling System (ECCS)model revisions as they apply to R.E.Ginna.Zn References 1)and 2), Westinghouse Electric Corporation provided information regarding modifications to their ECCS evaluation models to NRC Staff.References 1)and 2)describe the generic effects of the model revisions for both large and small break Loss Of Coolant Accidents (LOCA).The attachment to this letter provides information regarding the effects of the ECCS evaluation model modifications on the Ginna UFSAR Chapter 15.6.4 LOCA analysis.Modifications to the model cause the large break LOCA Peak Clad Temperature (PCT)to increase by 2 F to 1889 F.9004i30i69 90032S PDR ADOCK 05000244 R PDC i p'gget5/44'/
Modifications to the model for small break LOCA do not affect calculated.
peak clad temperature.
Very truly yours, Robert C.Mecredy Division Manager Nuclear Pr'oduction RWEi088 Enclosures xc: Mr.Allen R.Johnson (Mail Stop 14D1)Project Directorate I-3 Washington, D.C.20555 U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector ATT'O 10CFR50.46 ANNUAL PORT Effect of Westinghouse ECCS Evaluation Model Modifications on the LOCA Analysis Results Found in Chapter 15.6.4 of the R.E.Ginna (RG&E)Nuclear Power Plant Updated, Final Safety Analysis Report and.WCAP-11609 Containing the Steam Generator Tube Plu in Re ort for Ginna Nuclear Power Station The October 17, 1988 revision to 10CFR50.46 required applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory'ommission (NRC)of errors and changes in the ECCS Evaluation Models on an annual basis, when the errors and changes are not.significant.
Reference 1 defines a significant error or change as one which results in a calculated peak fuel cladding temperature dif f erent by more than 50'F from the temperature calculated for the limiting transient using the last acceptable model, or is a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50'F.In References 2 and 3, information regarding modifications to the Westinghouse large break and small break LOCA ECCS Evaluation Models was submitted to the NRC.The following presents an assessment of the effect of the modifications to the Westinghouse ECCS Evaluation Models on the Loss-Of-Coolant Accident (LOCA)analyses results found in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna Nuclear Power Station", October 1987 and Chapter 15.6.4 of the R.E.Ginna (RG&E)Nuclear Power Plant Updated Final Safety Analysis Report.LARGE BREAK LOCA-EVALUATION MODEL CHANGES The large break LOCA analysis for R.E.Ginna (RG&E)was examined to assess the effect of the applicable modifications to the Westinghouse large break LOCA ECCS Evaluation Model on Peak Cladding Temperature (PCT)results reported.in WCAP-11609.
The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP)level of 12'-o.The large break analysis was subsequently reanalyzed and licensed for 15%tube plugging and.the results were documented in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna Nuclear Power Station", October 1987.The large break LOCA analysis results were calculated.
using the 1981 version of the Westinghouse large break LOCA ECCS Evaluation Model which is documented in WCAP-9220-P-A (Reference 4).The current licensing basis analysis assumed.the following information important to the large break LOCA analysis NSSS power level-102-o of 1520 MWt Fuel Type-,14 X14 OFA Pellet Edge Configuration
-Chamfer Uniform Steam Generator Tube Plugging Level-15%i Nuclear Peaking Factors of 2.32 for the Total Peaking Factor and 1.66 for the Enthalpy Rise Peaking Factor.
~~
r~~or R.E.Ginna (R), the limiting break reslked from the double ended.guillotine rupture of the cold leg piping with a discharge coefficient of CD=0.4.The calculated peak cladding temperature was 1887'F including the applicable penalties.
The PCT value of 1887'F includes a 6'enalty to account for an upper plenum injection and.a core crossflow penalty'of 10'F.The following modifications to the Westinghouse ECCS Evaluation Model were evaluated to determine if they would affect the current licensing basis-large break LOCA analysis results for R.E.Ginna for 15%SGTP found.'n WCAP-11609.
1981 ECCS Evaluation Model: (Not Max-SI Limited,)In the 1981 version of the Westinghouse ECCS Evaluation Model, a modification was made to delay downcomer overfilling.
The delay corresponds to backfilling of the intact cold legs.Data from tests simulating cold leg injection during the post-large break LOCA reflood phase which have adequate safety injection flow to condense all of the-available steam flow show a significant amount of subcooled.
liquid to be present in the cold leg pipe test section.This situation corresponds to the so-called maximum safety injection scenario of ECCS Evaluation Model analyses.The R.E.Ginna (RGGE)LOCA analysis performed with the Westinghouse 1981 large break LOCA ECCS Evaluation Model is not affected.by the WREFLOOD code modifications since the maximum safeguards safety injection flow assumption is not limiting.1981 ECCS Evaluation Model.: (Two-Loop Plants)In the 1981'version of the Westinghouse ECCS Evaluation Model, the pressurizer is modeled as being attached to the broken (faulted)loop in the SATAN-VI code for calculating large break blowdown behavior.Sensitivity studies were performed to determine if this pressurizer location in the noding scheme was the most limiting position.The results indicated that two-loop Westinghouse PWRs are sensitive'to the pressurizer nodal location and that in some cases modeling the pressurizer in the intact (non-faulted) loop resulted.in a slight increase in the calculated Peak Cladding Temperature (PCT).For a two loop plant, core cooling is provided by negative core flow and.the negative core flow period.lasts through most of the remaining blowdown period;The concern regarding pressurizer location relates to the negative core flow period which is crucial for core cooling in a two-loop plant.With the pressurizer on the broken loop, pressurizer flow is a large contributor to break flow (pump side), lessening the contribution from the upper plenum and leaving a large upper plenum inventory for negative core flow later in blowdown.For Ginna, a penalty of 2'F due to modeling the pressurizer in the intact (non-faulted) loop was assessed to be used in tracking margin to the 10CFR50.46 limit on the current licensing basis analysis for 15-o SGTP contained.
in WCAP-11609.
The PCT for the limiting C=0.4 break is 1887'F with applicable penalties, and thus adding the 2'F increase brings the PCT, to 1889'F which is well below the 10CFR50.46 limit.
ks discussed abov modifications to the Nestgghouse large break LOCA ECCS Evaluation Model could affect the result by altering the~PCT.A.Analysis Calculated Result B.Modifications to Westinghouse ECCS Evaluation Model C.ECCS Evaluation Model Modifications Resultant PCT 1887'F 2oF 1889 F SMALL BREAK LOCA-EVALUATION MODEL CHANGES The small break.LOCA analysis for R.E.Ginna (RG&E)was also examined to assess the effect of the applicable modifications to the Westinghouse ECCS Evaluation Models on Peak Cladding Temperature (PCT)results reported in Chapter 15.6.4.1 of the UFSAR.The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP)level of 12%.The small break LOCA analysis was subsequently evaluated and licensed for a SGTP level increase from 12: to 15%.The small break LOCA event was not reanalyzed because the results are not sensitive to a SGTP level increase from 12%to 15:.The evaluation was documented, and transmitted in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna Nuclear Power Station", October 1987.The small break LOCA analysis results were calculated using the Westinghouse small break LOCA ECCS Evaluation Model documented in WCAP-8970 (Reference 12)which utilized the WFLASH computer code.For R.E.Ginna (RG&E), the limiting size small break resulted from a 6 inch equivalent diameter break in the cold leg.The calculated peak cladding temperature was 1092 F.The analysis assumed the following information important to the small break analyses: NSSS power level-102%of 1520 MWt Fuel Type-14 Z14 OFA Pellet Edge Configuration
-Chamfer Uniform Steam Generator Tube Plugging Level-12-o Nuclear Peaking Factors of 2e32 for the Total Peaking Factor and 1.66 for the Enthalpy Rise Peaking Factor.600 GPM Total Auziliary Feedwater Flow.The following modification to the Westinghouse ECCS Evaluation Model were evaluated to determine if they would affect the current licensing basis small break LOCA analysis results for R.E.Ginna.WFLASH ECCS Evaluation Model Following the accident at Three Mile Island Unit 2, additional attention was focused on the small break LOCA and.Westinghouse submitted a report, WCAP-9600 (Reference 5), to the Nuclear Regulatory Commission (NRC)detailing the performance of the Westinghouse small break LOCA Evaluation Model which utilized the WFLASH computer code.In NUREG-0611 (Reference 6), the NRC staff questioned the validity of certain models in the WFLASH, computer
~t1'V code and required censees to justify contin acceptance of the model.Section II.K.3.30 of NUREG-0737 (Reference 7), which clarified the NRC Post-TMI requirements regarding small break LOCA modeling, required that the licensees revise the small break LOCA ECCS models along the guidelines specified in NUREG-0611.
Following the issuance of NUREG-0737, Westinghouse and the Westinghouse Owners Group decided to develop the NOTRUMP (Reference 9)computer code for use in a new small break LOCA ECCS Evaluation Model (Reference 10).The NRC approved the use of NOTRUMP for small break LOCA ECCS analyses in May 1985.Since approval of the NOTRUMP small break LOCA ECCS Evaluation Model in 1985, the WFLASH computer code has not been maintained as.part of the Westinghouse ECCS Evaluation Model computer codes.In section II.K.3.31 of NUREG-0737, the NRC required that each licensee submit a new small break LOCA analysis using an NRC approved small break LOCA Evaluation Model.which satisfied the requirements of NUREG-0737 section II.K.3.30.
NRC Generic Letter 83-35 (Reference 8)relaxed the requirements of Item II.K.3.31 by allowing a more generic response and providing a basis for retention of the existing small, break LOCA analyses.Provided that the previously existing model results were demonstrated to be conservative with respect to the new small break LOCA model approved under the requirements of NUREG-0737 II.K.3.30 (NOTRUMP), plant specific analyses using the new small break LOCA Evaluation Model would not be required.In WCAP-11145 (Reference 11), Westinghouse and the Westinghouse Owners Group demonstrated that the result's obtained from calculations with WFLASH were conservative relative to those obtained.with NOTRUMP.Compliance with Item II.K.3.31 of NUREG-0737 could be completed by referencing WCAP-11145 and supplying some plant specific information.
Westinghouse, therefore, has not been modifying, investigating, or evaluating proposed changes to the WFLASH.small break LOCA ECCS Evaluation Model.As discussed above, none of the modifications to the Westinghouse small break LOCA ECCS Evaluation Model would affect the small break LOCA analysis results by altering the PCT.A.Analysis Calculated Result B.Modifications to Westinghouse ECCS Evaluation Model C.ECCS Evaluation Model Modifications Resultant PCT 1092'F+0oF 1092'F CONCLUSION An evaluation of the effect of modifications to the Westinghouse ECCS Evaluation Model as reported in References 2 and 3 was performed for both the large break LOCA and small break LOCA analyses results found in WCAP-11609 and Chapter 15.6.4 of the R.E.Ginna Nuclear Power Plant Updated Final Safety Analysis Report.The LOCA safety analyses reported in the UFSAR were performed for Ginna with a uniform Steam Generator Tube Plugging (SGTP)level of 12-o.The large break LOCA analysis was subsequently reanalyzed and licensed for an increase to 15: tube plugging.The small break LOCA analysis was evaluated and was not impacted by the increase in SGTP level to.The results of the ana es and evaluations for the increase in.SGTP level to 15-o were documented in WCAP-11609,"Steam Generator Tube Plugging (SGTP)Report for Ginna.Nuclear Power Station".When the effects of the ECCS model changes were combined with the current plant analysis results, it was determined that compliance with the requirements of 10CFR50.46 would be maintained.
REFERENCES 2.3.4.5.6.7.8.9.10.12."Emergency Core Cooling Systems;Revisions to Acceptance Criteria", Federal Register, Vol.53, No.180, pp.35996-36005, dated September, 16, 1988.NS-NRC-89-3463,"10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models," Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), dated.October 5, 1989.NS-NRC-89-3464,"Correction of Errors.and Modifications to the NOTRUMP Code in the Westinghouse Small Break LOCA ECCS Evaluation Model Which Are Potentially Significant", Letter from W.J.Johnson (Westinghouse) to T.E.Murley (NRC), dated October 5, 1989.WCAP-9220-P-A, Revision 1 (Proprietary), WCAP-9221-A, Revision 1 (Non-Proprietary),"Westinghouse ECCS Evaluation Model 1981 Version", 1981, Eicheldinger, C.WCAP-10924-P-A (Proprietary), WCAP-12130-A (Non-Proprietary),"Westinghouse Large Break LOCA Best Estimate Methodology", Hochreiter, L.E., et.al., January 1987.Report on Small Break Accidents for Westinghouse Nuclear Steam Supply System", WCAP-9601 (Non-Proprietary), June 1979, WCAP-9600 (Proprietary), June 1979."Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants", NUREG-0611, January 1980."Clarification of TMI Action Plan Requirements", NUREG-0737, November 1980."Clarification of TMI Plan Item II.K.3.31,"NRC Generic Letter 83-85 from D.G.Eisenhut, November 2, 1983."NOTRUMP-A Nodal Transient Small Break and General Network Code", WCAP-10079-P-A (Proprietary), WCAP-10081-A (Non-'Proprietary), Lee, N., et.al., August 1985."Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", WCAP-10054-P-A (Proprietary), WCAP-10081-A (Non-Proprietary), Lee, N., et.al., August 1985.WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary),"Westinghouse Emergency Core Cooling System Small Break October 1975 Model", April 1977.
