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| {{#Wiki_filter:Attachment IVR.E.GinnaNuclearPowerPlantIncludedpages:4.0-15.0-205.0-21Mark-upofExistingGinnaStationTechnical Specifications 98i2070083 98ii24PDRADQCK05000244PPGR J | | {{#Wiki_filter:Attachment IV R.E.Ginna Nuclear Power Plant Included pages: 4.0-1 5.0-20 5.0-21 Mark-up of Existing Ginna Station Technical Specifications 98i2070083 98ii24 PDR ADQCK 05000244 P PGR J |
| DesignFeatures4.04.0DESIGNFEATURES4.1SiteLocationThesitefortheR.E.GinnaNuclearPowerPlantislocatedonthesouthshoreofLakeOntario,approximately 16mileseastofRochester, NewYork.Theexclusion areaboundarydistances fromtheplantshallbeasfollows:Direction N(including offshore)
| | Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The site for the R.E.Ginna Nuclear Power Plant is located on the south shore of Lake Ontario, approximately 16 miles east of Rochester, New York.The exclusion area boundary distances from the plant shall be as follows: Direction N (including offshore)NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW'W NNW Distance m 8000 8000 8000, 8000 747 640 503 450 450 450 503 915 945 701 8000 8000 4.2 Reactor Core 4.2.1 Fuel Assemblies cH o<z.i rc~loyj Z%<~~>~c"~AN The reactor shall contain 121 fuel asse blies.Each assembly shall consist of a matrix of zircalloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO,)as fuel material.Limited substitutions of zircon~urn a o or stainless steel filler rods for fuel rods, in accor ance with approved applications of fuel rod configurations, may e use.Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.A limite number of lead test assemblies that have not completed repr sentative testing may be placed in nonlimiting core.regions.eye e sp~(continued) |
| NNENEENEEESESESSESSSWSWWSWWWNW'WNNWDistancem800080008000,8000747640503450450450503915945701800080004.2ReactorCore4.2.1FuelAssemblies cHo<z.irc~loyjZ%<~~>~c"~ANThereactorshallcontain121fuelasseblies.Eachassemblyshallconsistofamatrixofzircalloy cladfuelrodswithaninitialcomposition ofnaturalorslightlyenricheduraniumdioxide(UO,)asfuelmaterial.
| | R.E.Ginna Nuclear Power Plant 4.0-1 Amendment No.61 11 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) b.The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC,.specifically those described in the following documents:" 1.WCAP-9272-P-A,"Westinghouse Reload Safety Evaluation Methodology," July 1985.(Methodology for LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)2.3.WCAP-9220-P-A,"Westinghouse ECCS Evaluation Model-1981 Version," Revision 1, February 1982.(Hethodolog for LCO 3.2.1.)WCAP-8385,"Power Distribution Control and Load Following Procedures |
| Limitedsubstitutions ofzircon~urn aoorstainless steelfillerrodsforfuelrods,inaccorancewithapprovedapplications offuelrodconfigurations, mayeuse.Fuelassemblies shallbelimitedtothosefueldesignsthathavebeenanalyzedwithapplicable NRCstaffapprovedcodesandmethodsandshownbytestsoranalysestocomplywithallfuelsafetydesignbases.Alimitenumberofleadtestassemblies thathavenotcompleted reprsentative testingmaybeplacedinnonlimiting core.regions.eyeesp~(continued)
| | -Topical Report,"'September 1974.(Methodology for LCO 3.2.3.)4.WCAP-8567-P-A,"Improved Thermal Design Procedure;" February 1989.(Hethodology for LCO 3.4.1 whe usin P.5.WCAP 11397-P-A,"Revised Thermal Design Procedure," April 1989.(Methodology for LCD 3.4.1'hen using RTDP.)~n 6.'CAP-'10054-P-A and WCAP-1008+"Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.(Methodology for LCO 3.2.1)p~U~'"~c.i>>.WCAP-10924-P-A, Volume 1,.1, and Adden 1,2,3,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Vo 1: Model Description and a>a so ," December 1988.(Hethodology for LCO 3.2;1)C.t.l 8.WCAP-10924-P-A, Volume 2,.2, and nda"Westinghouse Large-Break LOCA Best-stimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," December 8.(Methodology for LCO 3.2.1)(continued) |
| R.E.GinnaNuclearPowerPlant4.0-1Amendment No.61 11 Reporting Requirements 5.65.6Reporting Requirements 5.6.5COLR(continued) b.Theanalytical methodsusedtodetermine thecoreoperating limitsshallbethosepreviously reviewedandapprovedbytheNRC,.specifically thosedescribed inthefollowing documents:" | | R.E.Ginna Nuclear Power Plant 5.0-20 Amendment No.61 eporting Requirements 5.6 5.6 Reporting, Requirements. |
| 1.WCAP-9272-P-A, "Westinghouse ReloadSafetyEvaluation Methodology," | | 5:6.5 (g~d~Revs'Sia~)~wc WCAP-10924-P olume 1, 1, Adden um 4,"We t hous OCA Best-stimate Methodology Model ascription and Validation Model Revisions,"~ugus 990 Pldrdk l1%I (ethodology for LCO 3.2.1)COLR (continue 9.WCAP-10924-P-A, Rev.2 and WCAP-12071,"Westinghouse Large-Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped With Upper Plenum Injection, Addendum 1: Responses to NRC guestions," December 1988.Hethodolo y for LCO 3.2.1)C.The core operating limits shall be determined such that all'pplicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS)limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits)of the safety analysis are met.d.The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to" the NRC.5.6.6 Reactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS~RT PL a~b.RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: |
| July1985.(Methodology forLCO3.1.1,LCO3.1.3,LCO3.1.5,LCO3.1.6,LCO3.2.1,LCO3.2.2,LCO3.2.3,andLCO3.9.1.)2.3.WCAP-9220-P-A, "Westinghouse ECCSEvaluation Model-1981 Version,"
| | LCO 3.4.3,"RCS Pressure and Temperature (P/T)Limits" The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP)System, and the LTOP enable temperature shall be established arid documented in the PTLR for the following: |
| Revision1,February1982.(Hethodolog forLCO3.2.1.)WCAP-8385, "PowerDistribution ControlandLoadFollowing Procedures
| | LCO 3.4.6,"RCS Loops-MODE 4";LCO 3.4.7,"RCS Loops-MODE 5, Loops Filled";LCO 3.4.10,"Pressurizer Safety Valves";and LCO 3.4.12,"LTOP System." (continued) |
| -TopicalReport,"'September 1974.(Methodology forLCO3.2.3.)4.WCAP-8567-P-A, "Improved ThermalDesignProcedure;" | | R.E.Ginna Nuclear Power Plant 5.0-21 Amendment No.61 I k INSERT 1 WCAP-13677-P-A,"10 CFR 50.46 Evaluation Model Report: WCOBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLO Cladding Option," February 1994.(Methodology for LCO 3.2.1).INSERT 2 WCAP-12610-P-A,"VANTAGE+Fuel Assembly Reference Core Report," April 1995.(Methodology for LCO 3.2.1). |
| February1989.(Hethodology forLCO3.4.1wheusinP.5.WCAP11397-P-A, "RevisedThermalDesignProcedure,"
| | '1 f i Attachment V R.E.Ginna Nuclear Power Plant Proposed Ginna Station Technical Specifications Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The site for the R.E.Ginna Nuclear Power Plant is located on the south shore of Lake Ontario, approximately 16 miles east of Rochester, New York.The exclusion area boundary distances from the plant shall be as follows: Direction N (including offshore)NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Distance m 8000 8000 8000 8000 747 640 503 450 450 450 503 915 945 701 8000 8000 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 121 fuel assemblies. |
| April1989.(Methodology forLCD3.4.1'henusingRTDP.)~n6.'CAP-'10054-P-A andWCAP-1008+
| | Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO,)as fuel material.Limited substitutions of zircaloy, ZIRLO, or stainless steel filler rods for fuel rods, in accordance with NRC approved applications of fuel rod configurations, may be used.Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or cycle specific analyses to comply with all fuel safety design bases.A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.I (continued) |
| "Westinghouse SmallBreakECCSEvaluation ModelUsingtheNOTRUMPCode,"August1985.(Methodology forLCO3.2.1)p~U~'"~c.i>>.WCAP-10924-P-A, Volume1,.1,andAdden1,2,3,"Westinghouse Large-Break LOCABest-Estimate Methodology, Vo1:ModelDescription anda>aso,"December1988.(Hethodology forLCO3.2;1)C.t.l8.WCAP-10924-P-A, Volume2,.2,andnda"Westinghouse Large-Break LOCABest-stimateMethodology, Volume2:Application toTwo-LoopPWRsEquippedwithUpperPlenumInjection," | | R.E.Ginna Nuclear Power Plant 4.0-1 Amendment No.Q 4 |
| December8.(Methodology forLCO3.2.1)(continued)
| | eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) b.The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: |
| R.E.GinnaNuclearPowerPlant5.0-20Amendment No.61 eportingRequirements 5.65.6Reporting, Requirements. | | 1.WCAP-9272-P-A,"Westinghouse Reload Safety Evaluation Methodology," July 1985.(Methodology for LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)2.WCAP-13677-P-A,"10 CFR 50.46 Evaluation Model Report: WCOBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLOŽCladding Option," February 1994.(Methodology for LCO 3.2.1.)3.WCAP-8385,"Power Distribution Control and Load Following Procedures |
| 5:6.5(g~d~Revs'Sia~)~wcWCAP-10924-P olume1,1,Addenum4,"WethousOCABest-stimateMethodology Modelascription andValidation ModelRevisions," | | -Topical Report," September 1974.(Hethodology for LCO 3.2.3.)4.WCAP-12610-P-A,"VANTAGE+Fuel Assembly Reference Core Report," April 1995.(Methodology for LCO 3.2.1).5.WCAP 11397-P-A,"Revised Thermal Design Procedure," April 1989.(Methodology for LCO 3.4.1 when using RTDP.)6.WCAP-10054-P-A and WCAP-10081-A,"Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.(Methodology for LCO 3.2.1)7.WCAP-10924-P-A, Volume 1, Revision 1,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation Responses to NRC guestions," and Addenda 1,2,3, December 1988.(Hethodology for LCO 3.2.1)(continued) |
| ~ugus990Pldrdkl1%I(ethodology forLCO3.2.1)COLR(continue 9.WCAP-10924-P-A, Rev.2andWCAP-12071, "Westinghouse Large-Break LOCABestEstimateMethodology, Volume2:Application toTwo-LoopPWRsEquippedWithUpperPlenumInjection, Addendum1:Responses toNRCguestions," | | R.E.Ginna Nuclear Power Plant 5.0-20 Amendment No.g li r I eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) 8.WCAP-10924-P-A, Volume 2, Revision 2,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addendum 1, December 1988.(Methodology for LCO 3.2.1)9.WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation, Addendum 4: Model Revisions," March 1991.(Methodology for LCO 3.2.1)c~The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS)limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits)of the safety analysis are met.d.The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.5.6.6 Reactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS REPORT PTLR a 0 b.RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following: |
| December1988.Hethodolo yforLCO3.2.1)C.Thecoreoperating limitsshallbedetermined suchthatall'pplicable limits(e.g.,fuelthermalmechanical limits,corethermalhydraulic limits,Emergency CoreCoolingSystems(ECCS)limits,nuclearlimitssuchasSDH,transient analysislimits,andaccidentanalysislimits)ofthesafetyanalysisaremet.d.TheCOLR,including anymidcyclerevisions orsupplements, shallbeprovideduponissuanceforeachreloadcycleto"theNRC.5.6.6ReactorCoolantSstemRCSPRESSUREANDTEMPERATURE LIMITS~RTPLa~b.RCSpressureandtemperature limitsforheatup,cooldown, criticality, andhydrostatic testingaswellasheatupandcooldownratesshallbeestablished anddocumented inthePTLRforthefollowing:
| | LCO 3.4.3,"RCS Pressure and Temperature (P/T)Limits" The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP)System, and the LTOP enable temperature shall be established and documented in the PTLR for the following: |
| LCO3.4.3,"RCSPressureandTemperature (P/T)Limits"ThepoweroperatedreliefvalveliftsettingsrequiredtosupporttheLowTemperature Overpressure Protection (LTOP)System,andtheLTOPenabletemperature shallbeestablished ariddocumented inthePTLRforthefollowing:
| | LCO 3.4.6,"RCS Loops-MODE 4";LCO 3.4.7,"RCS Loops-MODE 5, Loops Filled";LCO 3.4.10,"Pressurizer Safety Valves";and LCO 3.4.12,"LTOP System." (continued) |
| LCO3.4.6,"RCSLoops-MODE4";LCO3.4.7,"RCSLoops-MODE5,LoopsFilled";LCO3.4.10,"Pressurizer SafetyValves";andLCO3.4.12,"LTOPSystem."(continued)
| | R.E.Ginna Nuclear Power Plant 5.0-21 Amendment No.g |
| R.E.GinnaNuclearPowerPlant5.0-21Amendment No.61 Ik INSERT1WCAP-13677-P-A, "10CFR50.46Evaluation ModelReport:WCOBRA/TRAC Two-LoopUpperPlenumInjection ModelUpdatestoSupportZIRLOCladdingOption,"February1994.(Methodology forLCO3.2.1).INSERT2WCAP-12610-P-A, "VANTAGE+FuelAssemblyReference CoreReport,"April1995.(Methodology forLCO3.2.1). | | .0~5 1 J~(((,}} |
| '1f iAttachment VR.E.GinnaNuclearPowerPlantProposedGinnaStationTechnical Specifications DesignFeatures4.04.0DESIGNFEATURES4.1SiteLocationThesitefortheR.E.GinnaNuclearPowerPlantislocatedonthesouthshoreofLakeOntario,approximately 16mileseastofRochester, NewYork.Theexclusion areaboundarydistances fromtheplantshallbeasfollows:Direction N(including offshore) | |
| NNENEENEEESESESSESSSWSWWSWWWNWNWNNWDistancem8000800080008000747640503450450450503915945701800080004.2ReactorCore4.2.1FuelAssemblies Thereactorshallcontain121fuelassemblies.
| |
| EachassemblyshallconsistofamatrixofzircaloyorZIRLOcladfuelrodswithaninitialcomposition ofnaturalorslightlyenricheduraniumdioxide(UO,)asfuelmaterial.
| |
| Limitedsubstitutions ofzircaloy, ZIRLO,orstainless steelfillerrodsforfuelrods,inaccordance withNRCapprovedapplications offuelrodconfigurations, maybeused.Fuelassemblies shallbelimitedtothosefueldesignsthathavebeenanalyzedwithapplicable NRCstaffapprovedcodesandmethodsandshownbytestsorcyclespecificanalysestocomplywithallfuelsafetydesignbases.Alimitednumberofleadtestassemblies thathavenotcompleted representative testingmaybeplacedinnonlimiting coreregions.I(continued)
| |
| R.E.GinnaNuclearPowerPlant4.0-1Amendment No.Q 4 | |
| eportingRequirements 5.65.6Reporting Requirements 5.6.5COLR(continued) b.Theanalytical methodsusedtodetermine thecoreoperating limitsshallbethosepreviously reviewedandapprovedbytheNRC,specifically thosedescribed inthefollowing documents:
| |
| 1.WCAP-9272-P-A, "Westinghouse ReloadSafetyEvaluation Methodology," | |
| July1985.(Methodology forLCO3.1.1,LCO3.1.3,LCO3.1.5,LCO3.1.6,LCO3.2.1,LCO3.2.2,LCO3.2.3,andLCO3.9.1.)2.WCAP-13677-P-A, "10CFR50.46Evaluation ModelReport:WCOBRA/TRAC Two-LoopUpperPlenumInjection ModelUpdatestoSupportZIRLOŽCladdingOption,"February1994.(Methodology forLCO3.2.1.)3.WCAP-8385, "PowerDistribution ControlandLoadFollowing Procedures
| |
| -TopicalReport,"September 1974.(Hethodology forLCO3.2.3.)4.WCAP-12610-P-A, "VANTAGE+FuelAssemblyReference CoreReport,"April1995.(Methodology forLCO3.2.1).5.WCAP11397-P-A, "RevisedThermalDesignProcedure," | |
| April1989.(Methodology forLCO3.4.1whenusingRTDP.)6.WCAP-10054-P-A andWCAP-10081-A, "Westinghouse SmallBreakECCSEvaluation ModelUsingtheNOTRUMPCode,"August1985.(Methodology forLCO3.2.1)7.WCAP-10924-P-A, Volume1,Revision1,"Westinghouse Large-Break LOCABest-Estimate Methodology, Volume1:ModelDescription andValidation Responses toNRCguestions,"
| |
| andAddenda1,2,3,December1988.(Hethodology forLCO3.2.1)(continued)
| |
| R.E.GinnaNuclearPowerPlant5.0-20Amendment No.g lirI eportingRequirements 5.65.6Reporting Requirements 5.6.5COLR(continued) 8.WCAP-10924-P-A, Volume2,Revision2,"Westinghouse Large-Break LOCABest-Estimate Methodology, Volume2:Application toTwo-LoopPWRsEquippedwithUpperPlenumInjection," | |
| andAddendum1,December1988.(Methodology forLCO3.2.1)9.WCAP-10924-P-A, Volume1,Revision1,Addendum4,"Westinghouse Large-Break LOCABest-Estimate Methodology, Volume1:ModelDescription andValidation, Addendum4:ModelRevisions,"
| |
| March1991.(Methodology forLCO3.2.1)c~Thecoreoperating limitsshallbedetermined suchthatallapplicable limits(e.g.,fuelthermalmechanical limits,corethermalhydraulic limits,Emergency CoreCoolingSystems(ECCS)limits,nuclearlimitssuchasSDH,transient analysislimits,andaccidentanalysislimits)ofthesafetyanalysisaremet.d.TheCOLR,including anymidcyclerevisions orsupplements, shallbeprovideduponissuanceforeachreloadcycletotheNRC.5.6.6ReactorCoolantSstemRCSPRESSUREANDTEMPERATURE LIMITSREPORTPTLRa0b.RCSpressureandtemperature limitsforheatup,cooldown, criticality, andhydrostatic testingaswellasheatupandcooldownratesshallbeestablished anddocumented inthePTLRforthefollowing:
| |
| LCO3.4.3,"RCSPressureandTemperature (P/T)Limits"ThepoweroperatedreliefvalveliftsettingsrequiredtosupporttheLowTemperature Overpressure Protection (LTOP)System,andtheLTOPenabletemperature shallbeestablished anddocumented inthePTLRforthefollowing:
| |
| LCO3.4.6,"RCSLoops-MODE4";LCO3.4.7,"RCSLoops-MODE5,LoopsFilled";LCO3.4.10,"Pressurizer SafetyValves";andLCO3.4.12,"LTOPSystem."(continued)
| |
| R.E.GinnaNuclearPowerPlant5.0-21Amendment No.g | |
| .0~51J~(((,}} | |
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1999-06-28
[Table view]
Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML17250B3061999-10-20020 October 1999
Proposed Tech Specs,Consisting of 1999 Changes to TS Bases
ML17265A6941999-06-28028 June 1999
Proposed Tech Specs,Revising ITS Associated with RCS Leakage Detection Instrumentation,As Result of Commitment Submitted as Part of Staff Review of Application of leak-before-break Status to Protions of RHR Piping
ML17265A6401999-05-12012 May 1999
Rev 11 to Technical Requirements Manual for Ginna Station.
ML17265A6261999-04-18018 April 1999
Rev 10 to Technical Requirements Manual (Trm), for Ginna Station
ML17311A0701999-04-14014 April 1999
Rev 10 to AP-PRZR.1, Abnormal Pressurizer Pressure.
ML17309A6521999-04-14014 April 1999
Rev 14 to AP-RCS.1, Reactor Coolant Leak.
ML17309A6531999-04-13013 April 1999
Rev 1 to FIG-2.0, Figure Sdm. with 990413 Ltr
ML17265A6111999-03-26026 March 1999
Rev 9 to Technical Requirements Manual for Ginna Station.
ML17265A5911999-03-0101 March 1999
Proposed Tech Specs Change Revising Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6)
ML17265A5571999-02-25025 February 1999
Rev 4 to Technical Requirements Manual (Trm).
ML17265A5491999-02-12012 February 1999
to EOP FR-H.5, Response to SG Low Level.
ML17265A5481999-02-12012 February 1999
to EOP ATT-22.0, Attachment Restoring Feed Flow.
ML17265A5461999-02-12012 February 1999
0 to EOP AP-TURB.1, Turbine Trip Without Rx Trip Required.
ML17309A6481999-01-25025 January 1999
Revised Ginna Station Emergency Operating Procedures. with 990125 Ltr
ML17265A5151999-01-14014 January 1999
Revised Emergency Operating Procedures,Including Rev 14 to AP-SW.1,rev 3 to ATT-5.2,rev 5 to ATT-8.0,rev 1 to ATT-14.6, Rev 17 to E-1,rev 16 to ECA-1.1,rev 26 to ES-1.3,rev 4 to FR-Z.2 & Index
ML17265A5081999-01-0808 January 1999
Rev 7 to Technical Requirements Manual, for Ginna Station
ML17265A4961998-12-18018 December 1998
Rev 6 to Technical Requirements Manual.
ML20198C0361998-12-14014 December 1998
Rev 11 to ECA-0.2, Loss of All AC Power Recovery with SI Required
ML20198C0121998-12-14014 December 1998
Rev 10 to AP-RCS.2, Loss of Reactor Coolant Flow
ML20198C0581998-12-14014 December 1998
Rev 5 to FR-Z.1, Response to High Containment Pressure
ML20198C0411998-12-14014 December 1998
Rev 16 to FR-C.1, Response to Inadequate Core Cooling
ML20198C0521998-12-14014 December 1998
Rev 13 to FR-S.1, Response to Reactor Restart/Atws
ML17265A4661998-11-24024 November 1998
Proposed Tech Specs 4.2.1,revising Description of Fuel Cladding Matl & Updating List of References Provided in TS 5.6.5 for COLR
ML20155G0301998-10-30030 October 1998
Rev 13 to EOP AP-CCW.1, Leakage Into Component Cooling Loop
ML17265A4091998-08-24024 August 1998
Rev 13 to AP-SW.1, Svc Water Leak. W/980824 Ltr
ML17265A3721998-07-16016 July 1998
Rev 14 to EOP AP-CW.1, Loss of Circ Water Pump. W/980716 Ltr
ML17265A3151998-06-0404 June 1998
Proposed Tech Specs Basis Re Main Steam Isolation Setpoint
ML17265A2941998-05-21021 May 1998
Tech Specs Consisting of Submittal Changes for 1998
ML17265A2861998-05-0606 May 1998
Rev 19 to EOP ECA-3.2, SGTR W/Loss of Reactor Coolant Saturated Recovery Desired & Updated ECA Index.W/980506 Ltr
ML17265A2461998-04-27027 April 1998
Proposed Tech Specs Revising Requirements Associated W/Sfp to Reflect Planned Mod to Storage Racks & Temporarily Addressing Boraflex Degradation within Pool
ML17265A1821998-02-20020 February 1998
Rev 1 to EWR 5111, MOV Qualification Program Plan, Calculation Assumption Verification Criteria.
ML17265A1491997-12-31031 December 1997
Rev 1 to Inservice Testing Program.
ML17264B0731997-10-14014 October 1997
Rev 4 to Technical Requirements Manual (TRM) for Ginna Station.
ML17264B0671997-10-0808 October 1997
Proposed Tech Specs,Correcting & Clarifying Info Re RCS Pressure & Temp Limits Rept Administrative Controls Requirements
ML17264B0441997-09-29029 September 1997
Proposed Tech Specs Revising Adminstrative Controls W/ Respect to Reactor Coolant Sys Pressure & Temp Limits Rept
ML17264B0391997-09-29029 September 1997
Proposed Tech Specs Revising Allowable Value & Trip Setpoint for High Steam Flow Input Into LCO Table 3.3.2-1,Function 4d (Main Steam Isolation) to Address Issues Identified in Rev Revised Setpoint Analysis Study
ML17265A1541997-09-16016 September 1997
Rev 1 to DA EE-92-089-21, Design Analysis Ginna Station Instrument Loop Performance Evaluation & Setpoint Verification.
ML17264B0031997-08-19019 August 1997
Proposed Tech Specs Allowing Testing of Three ECCS motor- Operated Valves in Mode 4 Which Currently Requires Entry Into LCO 3.0.3
ML17264A9991997-08-19019 August 1997
Proposed Tech Specs Correcting Specified Accumulator Borated Water Volume Values in SR 3.5.1.2 to Match Associated Accumulator Percent Level Values
ML17264A9791997-08-0505 August 1997
Rev 7 to AP-RCS.3, High Reactor Coolant Activity.
ML17264A9781997-08-0505 August 1997
Rev 12 to AP-RCS.1, Reactor Coolant Leak.
ML17264A9771997-08-0505 August 1997
Rev 11 to AP-RCP.1, RCP Seal Malfunction.
ML17264A9751997-08-0505 August 1997
Rev 14 to AP-ELEC.1, Loss of 12A & 12B Busses.
ML17264A9851997-08-0404 August 1997
Rev 2 to Summary Description of Compliance w/10CFR73 Amend Protection Against Malevolent Use of Vehicles at Nuclear Power Plants.
ML17264A9391997-07-0707 July 1997
Rev 3 to Technical Requirements Manual for Ginna Station, Inserting New Tabs Accordingly Per Previous Transmittal for Rev
ML17264A9341997-07-0101 July 1997
to Technical Requirements Manual (TRM) & inter-office Correspondence Dtd 970624
ML17264A9461997-06-20020 June 1997
to Inservice Insp (ISI) Program.
ML17264A9241997-06-0303 June 1997
Proposed Tech Specs Clarifying Issues Re Low Temperature Overpressure Protection
ML17311A0471997-05-22022 May 1997
Revised Eops,Including Procedures Index,Rev 5 to ATT-15.0, Rev 3 to ATT-15.2,rev 3 to AP-ELEC.3,rev 13 to ECA-0.1 & Rev 9 to ECA-0.2.W/970522 Ltr
ML17264A8671997-04-24024 April 1997
Proposed Tech Specs,Revising Rcs,Pt & Administrative Control Requirements
1999-06-28
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Text
Attachment IV R.E.Ginna Nuclear Power Plant Included pages: 4.0-1 5.0-20 5.0-21 Mark-up of Existing Ginna Station Technical Specifications 98i2070083 98ii24 PDR ADQCK 05000244 P PGR J
Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The site for the R.E.Ginna Nuclear Power Plant is located on the south shore of Lake Ontario, approximately 16 miles east of Rochester, New York.The exclusion area boundary distances from the plant shall be as follows: Direction N (including offshore)NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW'W NNW Distance m 8000 8000 8000, 8000 747 640 503 450 450 450 503 915 945 701 8000 8000 4.2 Reactor Core 4.2.1 Fuel Assemblies cH o<z.i rc~loyj Z%<~~>~c"~AN The reactor shall contain 121 fuel asse blies.Each assembly shall consist of a matrix of zircalloy clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO,)as fuel material.Limited substitutions of zircon~urn a o or stainless steel filler rods for fuel rods, in accor ance with approved applications of fuel rod configurations, may e use.Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.A limite number of lead test assemblies that have not completed repr sentative testing may be placed in nonlimiting core.regions.eye e sp~(continued)
R.E.Ginna Nuclear Power Plant 4.0-1 Amendment No.61 11 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) b.The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC,.specifically those described in the following documents:" 1.WCAP-9272-P-A,"Westinghouse Reload Safety Evaluation Methodology," July 1985.(Methodology for LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)2.3.WCAP-9220-P-A,"Westinghouse ECCS Evaluation Model-1981 Version," Revision 1, February 1982.(Hethodolog for LCO 3.2.1.)WCAP-8385,"Power Distribution Control and Load Following Procedures
-Topical Report,"'September 1974.(Methodology for LCO 3.2.3.)4.WCAP-8567-P-A,"Improved Thermal Design Procedure;" February 1989.(Hethodology for LCO 3.4.1 whe usin P.5.WCAP 11397-P-A,"Revised Thermal Design Procedure," April 1989.(Methodology for LCD 3.4.1'hen using RTDP.)~n 6.'CAP-'10054-P-A and WCAP-1008+"Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.(Methodology for LCO 3.2.1)p~U~'"~c.i>>.WCAP-10924-P-A, Volume 1,.1, and Adden 1,2,3,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Vo 1: Model Description and a>a so ," December 1988.(Hethodology for LCO 3.2;1)C.t.l 8.WCAP-10924-P-A, Volume 2,.2, and nda"Westinghouse Large-Break LOCA Best-stimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," December 8.(Methodology for LCO 3.2.1)(continued)
R.E.Ginna Nuclear Power Plant 5.0-20 Amendment No.61 eporting Requirements 5.6 5.6 Reporting, Requirements.
5:6.5 (g~d~Revs'Sia~)~wc WCAP-10924-P olume 1, 1, Adden um 4,"We t hous OCA Best-stimate Methodology Model ascription and Validation Model Revisions,"~ugus 990 Pldrdk l1%I (ethodology for LCO 3.2.1)COLR (continue 9.WCAP-10924-P-A, Rev.2 and WCAP-12071,"Westinghouse Large-Break LOCA Best Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped With Upper Plenum Injection, Addendum 1: Responses to NRC guestions," December 1988.Hethodolo y for LCO 3.2.1)C.The core operating limits shall be determined such that all'pplicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS)limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits)of the safety analysis are met.d.The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to" the NRC.5.6.6 Reactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS~RT PL a~b.RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3,"RCS Pressure and Temperature (P/T)Limits" The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP)System, and the LTOP enable temperature shall be established arid documented in the PTLR for the following:
LCO 3.4.6,"RCS Loops-MODE 4";LCO 3.4.7,"RCS Loops-MODE 5, Loops Filled";LCO 3.4.10,"Pressurizer Safety Valves";and LCO 3.4.12,"LTOP System." (continued)
R.E.Ginna Nuclear Power Plant 5.0-21 Amendment No.61 I k INSERT 1 WCAP-13677-P-A,"10 CFR 50.46 Evaluation Model Report: WCOBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLO Cladding Option," February 1994.(Methodology for LCO 3.2.1).INSERT 2 WCAP-12610-P-A,"VANTAGE+Fuel Assembly Reference Core Report," April 1995.(Methodology for LCO 3.2.1).
'1 f i Attachment V R.E.Ginna Nuclear Power Plant Proposed Ginna Station Technical Specifications Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The site for the R.E.Ginna Nuclear Power Plant is located on the south shore of Lake Ontario, approximately 16 miles east of Rochester, New York.The exclusion area boundary distances from the plant shall be as follows: Direction N (including offshore)NNE NE ENE E ESE SE SSE S SSW SW WSW W WNW NW NNW Distance m 8000 8000 8000 8000 747 640 503 450 450 450 503 915 945 701 8000 8000 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 121 fuel assemblies.
Each assembly shall consist of a matrix of zircaloy or ZIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO,)as fuel material.Limited substitutions of zircaloy, ZIRLO, or stainless steel filler rods for fuel rods, in accordance with NRC approved applications of fuel rod configurations, may be used.Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or cycle specific analyses to comply with all fuel safety design bases.A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.I (continued)
R.E.Ginna Nuclear Power Plant 4.0-1 Amendment No.Q 4
eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) b.The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.WCAP-9272-P-A,"Westinghouse Reload Safety Evaluation Methodology," July 1985.(Methodology for LCO 3.1.1, LCO 3.1.3, LCO 3.1.5, LCO 3.1.6, LCO 3.2.1, LCO 3.2.2, LCO 3.2.3, and LCO 3.9.1.)2.WCAP-13677-P-A,"10 CFR 50.46 Evaluation Model Report: WCOBRA/TRAC Two-Loop Upper Plenum Injection Model Updates to Support ZIRLOŽCladding Option," February 1994.(Methodology for LCO 3.2.1.)3.WCAP-8385,"Power Distribution Control and Load Following Procedures
-Topical Report," September 1974.(Hethodology for LCO 3.2.3.)4.WCAP-12610-P-A,"VANTAGE+Fuel Assembly Reference Core Report," April 1995.(Methodology for LCO 3.2.1).5.WCAP 11397-P-A,"Revised Thermal Design Procedure," April 1989.(Methodology for LCO 3.4.1 when using RTDP.)6.WCAP-10054-P-A and WCAP-10081-A,"Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.(Methodology for LCO 3.2.1)7.WCAP-10924-P-A, Volume 1, Revision 1,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation Responses to NRC guestions," and Addenda 1,2,3, December 1988.(Hethodology for LCO 3.2.1)(continued)
R.E.Ginna Nuclear Power Plant 5.0-20 Amendment No.g li r I eporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 COLR (continued) 8.WCAP-10924-P-A, Volume 2, Revision 2,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 2: Application to Two-Loop PWRs Equipped with Upper Plenum Injection," and Addendum 1, December 1988.(Methodology for LCO 3.2.1)9.WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4,"Westinghouse Large-Break LOCA Best-Estimate Methodology, Volume 1: Model Description and Validation, Addendum 4: Model Revisions," March 1991.(Methodology for LCO 3.2.1)c~The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS)limits, nuclear limits such as SDH, transient analysis limits, and accident analysis limits)of the safety analysis are met.d.The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.5.6.6 Reactor Coolant S stem RCS PRESSURE AND TEMPERATURE LIMITS REPORT PTLR a 0 b.RCS pressure and temperature limits for heatup, cooldown, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3,"RCS Pressure and Temperature (P/T)Limits" The power operated relief valve lift settings required to support the Low Temperature Overpressure Protection (LTOP)System, and the LTOP enable temperature shall be established and documented in the PTLR for the following:
LCO 3.4.6,"RCS Loops-MODE 4";LCO 3.4.7,"RCS Loops-MODE 5, Loops Filled";LCO 3.4.10,"Pressurizer Safety Valves";and LCO 3.4.12,"LTOP System." (continued)
R.E.Ginna Nuclear Power Plant 5.0-21 Amendment No.g
.0~5 1 J~(((,
