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| | issue date = 10/31/1988 | | | issue date = 10/31/1988 |
| | title = Reduced Temp & Pressure Operation for Donald C. Cook Nuclear Plant Unit 1 Licensing Rept, Including FSAR mark-ups | | | title = Reduced Temp & Pressure Operation for Donald C. Cook Nuclear Plant Unit 1 Licensing Rept, Including FSAR mark-ups |
| | author name = AUGUSTINE D B, CECCHETT D L | | | author name = Augustine D, Cecchett D |
| | author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. | | | author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| | addressee name = | | | addressee name = |
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| {{#Wiki_filter:MESTINGHOUSECLASSIIIHCAP-11902REDUCEDTEMPERATUREANOPRESSUREOPERATIONFORDONALDC.COOKNUCLEARPLANTUNIT1LICENSINGREPORT0.L.Cecchett0.B.AugustineOctober1988eWESTINGHOUSEELECTRICCORPORATIONEnergySystemsBusinessUnitP.O.Box355Pittsburgh,Pennsy1vania15230pop/g/'980a:1d/100588 cumulativefatigueusagefactorwascalculatedtobe0.69whichislessthanthe1.0Codelimit.Thisvalueincludesthethermalstressduetothenon-linearportionofthethermalgradientforconservatism.Asubstantialreductioninthemaximumrangeofstressintensitycouldbeachievedifthenon-linearthermalgradientcontributionwereneglectedaspermittedbytheASHECode.3.10.1.2ReactorVesselIntegrityAnevaluationoftheimpactofreratingonreactorvesselintegrityforneutronembrittlementwasperformed.Neutronfluencechangesforthereratingwerecalculated.Usingtheserevisedfluencesandthosefromotherrelevantsystemsparametersassociatedwiththererating,theassessmentincludedreviewofsurveillancecapsulewithdrawalschedules,thescheduleofapplicabilityoftheplantheatupandcooldownlimits,and10CFR50-~~~~~~~~~~~~AppendixGanalyses,includingaverificationofplantspecificmaterialproperties.ArevisiontothecalculationsusedinthesubmittaltotheNRCformeeting-therequirementsofthepressurizedthermalshock(PTS)rulewereperformed.Finally,aninitialevaluationoftheimpactofreratingonthePTSriskofvesselfailurewascarriedouttoconfirmtheapplicabilityofthescreeningcriteriainthePTSrulefortheCookNuclearPlantreactorvessels.ThereratingaffectsthePTStransientinitiatingtemperaturewhichislowerthanthatusedinthegenericPTSriskanalyseswhichsupportthescreeningcriteria.~Neutrontransportcalculationshavebeencompletedataconservativelyhighpowerlevelof3600HWt,andreducedtemperature/pressureconditionsforCookNuclearPlantUnit1.Resultsoftheseanalyseshaveindicatedthatinadditiontotheincreasedreactorpower,operationoftheunitswithreducedcoolanttemperatures,particularlyinthedowncomerregions,alsohasasignificantimpactonthefastneutronexposureratesincidentonthepressurevessel.Alsoimpactedaretherelationshipsamongtheneutronexposureratesatsurveillancecapsulelocationsandthoseatpositionswithinthepressurevesselwall;i.e.,capsuleleadfactors.Agd~~9tzpgQg78739a:1d/0929883.10-5 | | {{#Wiki_filter:MESTINGHOUSE CLASS III HCAP-11902 REDUCED TEMPERATURE ANO PRESSURE OPERATION FOR DONALD C. COOK NUCLEAR PLANT UNIT 1 LICENSING REPORT |
| ~I Toprovideaboundingevaluationofthesevariouseffects,calculationswereperformedforthecurrentlicensedconditionsaswellasforupratedpowerwithbothmaximumandminimumdowncomertemperatures.TheresultsofthesestudiesaresummarizedinTables3.10.1-1and3.10.1-2.AnexaminationofTables3.10.1-1and3.10.1-2showsthattheimpactofhigherpoweroperationandchangedcoolanttemperaturesisgreaterontheabsolutemagnitudeoftheneutronexposureratethanontherelativebehaviorofneutronfluxdistributionsasreflectedincapsuleleadfactors.TwosetsofdatahavebeenprovidedtoestablishupperandlowerboundexposuresforCookNuclearPlantUnit1.Theupperlimitconditionsareconsistentwithassuminganupratingattheonsetofthenextfuelcyclewithcontinuedoperationataninlettemperatureof547'Fforunit1.Lowerboundfluenceestimateswerebasedonnoupratinginpowerlevel,butwithfutureoperationoccurringatreducedT;i.e.,theinlettemperaturewasassumedavg'obe512'F.Theselowerboundfluenceestimatesarelowerthanthoseatcurrentoperatingconditions.ReviewoftheheatupandcooldowncurvesthatwerepreviouslygeneratedbySouthwestResearchInstituteindicatethatthesecurvesweregeneratedinaccordancewithRegulatoryGuide1.99-Revision1.PerNRCGenericLetter88-11,datedJuly12,1988,allutilitiesmustsubmittotheNRCbyNovember1988theresultsoftheirtechnicalanalysisrelativetotheimplementationofRegulatoryGuide1.99-Revision2,whichwasofficiallyissuedinMay1988.Giventhisregulatorychange,theeffectofreratingshouldbeincorporatedIIintofuturecalculationsthatwillbeperformedforrevisingheatupandcooldowncurvesinaccordancewiththislatestrevisionofRegulatoryGuide1.99.Theeffectofthereratingisdeemedtonotbesignificant.Thechangesinthesystemsparametersassociatedwiththereratinghavebeenjudgedtonothaveanysignificantimpacton10CFRPart50-AppendixGanalysis.8139e:1d/0930883.10-6IZpa~+g7 S | | : 0. L. Cecchett |
| TheCookNuclearPlantUnit1reactorvesselbeltlineregionmaterialpropertieswereverifiedagainstthelatestavailableinformationinvariousindustrydata.basesandsurveillancecapsulereports.ThepropertiesdefinedfromthelatestinformationareconsistentwiththoseusedinpriorutilitysubmittalstotheNRCrelativetomeetingtherequirementsofthePTSrule.SincethecoreloadingpatternwiIIbechangingasaresultofthererating,anupdatetothePTSsubmittalwillberequiredasstatedinthePTSRule.RevisedcalculationswereperformedforthereratingusingthecurrentPTSRulemethodologyandthelatestprocedurespecifiedbyRegulatoryGuide1.99-Revision2.AsstatedinNRCGenericLetter88-11,thestaffispresentlyconsideringanamendmenttothePTSRule,10CFR50.61,thatwillreplacetheequationsforRTptsgiveninparagraph(b)(2)withthecalculationprocedureinSectionC.lofRevision2toRegulatoryGuide1.99,buttheywillnotchangethescreeningcriteria.AlltheRTptsvaluesremainbelowtheNRCscreeningvaluesforPTSusingtheprojectedfluencevaluesthatarebaseduponreratedconditionsthroughthelicenseexpiration.ThehighestRTptsvalue(265'F)wascalculatedatthecircumferentialweldoftheCookNuclearPlantUnitIreactorvessel,usingthemethodologyofRegulatoryGuide1.99Revision2.OnthebasisofprobabilisticworkdescribedinthePTSRule,theNRCstaff'oncludedthatPHRvesselswithconservativelycalculatedvaluesofRTndt(i.e.,RTpts)lessthan270'Fforplatematerialandaxialwelds,andlessthan300'FforcircumferentialweldspresentanacceptablylowriskofvesselfailurefromPTSevents.Thisevaluation,however,didnottakeintoaccounttheimpactofrerating,whichcausespotentialPTStransientscenariostobeginfromalowersystemtemperature.81390:1d/0929883.10"7 Aninitialevaluationwasperformedtodeterminetheimpactofreratingonthe~~~~~~~applicabilityofthePTSscreeningcriteriaintermsofriskofvesselfailure.AprobabilisticfracturemechanicssensitivitystudyoflimitingPTStransientcharacteristics,startingfromaloweroperatingtemperature,showedthattheconditionalprobabilityofreactorvesselfailurewillnotbeadverselyaffected.Therefore,theoverallriskofvesselfailurewillnotbeadverselyimpactedmeaningthatthatthescreeningcriteriainthePTSRulearestillapplicablefortheCookNuclearPlantUnit1reactorvesselrelativetoreratedconditions.813S@:Id/0929883.10-8 TABLE3.10.1-1FASTNEUTRON(E>1.0MeV)FLUENCEPROJECTIONSFORCOOKNUCLEARPLANTUNIT1Unit122.89EFPYAllplates;Weld9-442Welds2-442B,2-442C,3-442A,3-442CUerBound1.84x101.19x10LowerBound1.55x101.01x10Weids2-442A,3-442B5.92x105.01x108139e:1d/0929883.10-9 I
| | : 0. B. Augustine October 1988 WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit P.O. Box 355 Pittsburgh, Pennsy1vania 15230 e pop /g |
| TABLE3.1.0.1-2SURVEILLANCECAPSULELEADFACTORSFORCOOKNUCLEARPLANTUNIT14'aosules40'apsulesUnit1BaseCase(3250Wt,536'FDowncomer)1.34.2Unit1AtUpratedPower(3588HMt,547'FDowncomer)1.34.2Unit1AtUpratedPower(3588HMt,512'FDowncomer)1.34'8139e:1d/0929883.10"10pffQQIPNCJtjI2PV~<7 E
| | /'980a:1d/100588 |
| TABLE4.1-1SYSTEMDESIGNANDOPERATINGPARAMETERSiPlantdesignlife,yearsNumberofheatcransferloopsDesignpressure,psigNominaloperatingpressure,psigTocalsystemvolumeincludingpressurizerandsurgeline(ambientconditions),ft(estimated)3Systemliquidvolume,includingpressurizerandsurgeline(ambientconditions),fcSystemliquidvolume,includingpressurizer3max.,guaranteedpower,fc(escimac'ed)TocalReactorheatoutput(100%power)Btu/hr402485223512,50011,89211,78011,089x10(Unit1)6(3250MVt)11,641x10(Unit2)6(3411HPt)Unit1BoundingConditionsforReracingLower/UpperUnit2Reactorvesselcoolanttemperatureatfullpower:Inlet,nominal,F0Outlet,nominal,F0Coolanttemperatureriseinvessel0acfullpower,avg.,FTotalcoolantflowrace,lb/hrx106Sceampressureacfullpo~er,psiaSteamTemp.Qfullpower,FTotalReactorCoolantVolumeatambientcondicions,ft3514.9/545.2579.1/607.564.2/62.3139.0/133.9618/820489'/521.112,438541.3606.464.8134.6820521.112,4704.1-25July1990}}
| | |
| | c umulative fatigue usage factor was calculated to be 0.69 which is less than the 1.0 Code limit. |
| | This value includes the thermal stress due to the non-linear portion of the thermal gradient for conservatism. A substantial reduction in the maximum range of stress intensity could be achieved if the non-linear thermal gradient contribution were neglected as permitted by the ASHE Code. |
| | 3.10.1.2 Reactor Vessel Integrity An evaluation of the impact of rerating on reactor vessel integrity for neutron embrittlement was performed. Neutron fluence changes for the rerating were calculated. Using these revised fluences and those from other relevant systems parameters associated with the rerating, the assessment included review of surveillance capsule withdrawal schedules, the schedule of applicability of the plant heatup and cooldown limits, and 10 CFR 50- |
| | ~ |
| | Appendix G analyses, including a verification of plant specific material |
| | ~ ~ ~ ~ ~ ~ |
| | properties. A revision to the calculations used in the submittal to the NRC |
| | ~ |
| | ~ ~ ~ |
| | for meeting- the requirements of the pressurized thermal shock (PTS) rule were |
| | ~ |
| | performed. Finally, an initial evaluation of the impact of rerating on the PTS risk of vessel failure was carried out to confirm the applicability of the screening criteria in the PTS rule for the Cook Nuclear Plant reactor vessels. The rerating affects the PTS transient initiating temperature which is lower than that used in the generic PTS risk analyses which support the screening criteria. |
| | ~ |
| | Neutron transport calculations have been completed at a conservatively high power level of 3600 HWt, and reduced temperature/pressure conditions for Cook Nuclear Plant Unit 1. Results of these analyses have indicated that in addition to the increased reactor power, operation of the units with reduced coolant temperatures, particularly in the downcomer regions, also has a significant impact on the fast neutron exposure rates incident on the pressure vessel. Also impacted are the relationships among the neutron exposure rates at surveillance capsule locations and those at positions within the pressure vessel wall; i.e., capsule lead factors. |
| | Agd~~ 9 tz p gQg7 8739a:1d/092988 3.10-5 |
| | |
| | ~ I To provide a bounding evaluation of these various effects, calculations were performed for the current licensed conditions as well as for uprated power with both maximum and minimum downcomer temperatures. The results of these studies are summarized in Tables 3.10.1-1 and 3.10.1-2. |
| | An examination of Tables 3.10.1-1 and 3.10.1-2 shows that the impact of higher power operation and changed coolant temperatures is greater on the absolute magnitude of the neutron exposure rate than on the relative behavior of neutron flux distributions as reflected in capsule lead factors. |
| | Two sets of data have been provided to establish upper and lower bound exposures for Cook Nuclear Plant Unit 1. The upper limit conditions are consistent with assuming an uprating at the onset of the next fuel cycle with continued operation at an inlet temperature of 547'F for unit 1. Lower bound fluence estimates were based on no uprating in power level, but with future operation occurring at reduced T ; i.e., the inlet temperature was assumed avg'o be 512'F. These lower bound fluence estimates are lower than those at current operating conditions. |
| | Review of the heatup and cooldown curves that were previously generated by Southwest Research Institute indicate that these curves were generated in accordance with Regulatory Guide 1.99-Revision 1. Per NRC Generic Letter 88-11, dated July 12, 1988, all utilities must submit to the NRC by November 1988 the results of their technical analysis relative to the implementation of Regulatory Guide 1.99-Revision 2, which was officially issued in May 1988. |
| | Given this regulatory change, the effect of rerating should be incorporated II into future calculations that will be performed for revising heatup and cooldown curves in accordance with this latest revision of Regulatory Guide 1.99. The effect of the rerating is deemed to not be significant. |
| | The changes in the systems parameters associated with the rerating have been judged to not have any significant impact on 10 CFR Part 50 - Appendix G analysis. |
| | IZ pa~ +g 7 8139e:1d/093088 3.10-6 |
| | |
| | S The Cook Nuclear Plant Unit 1 reactor vessel beltline region material properties were verified against the latest available information in various industry data. bases and surveillance capsule reports. The properties defined from the latest information are consistent with those used in prior utility submittals to the NRC relative to meeting the requirements of the PTS rule. |
| | Since the core loading pattern wi II be changing as a result of the rerating, an update to the PTS submittal will be required as stated in the PTS Rule. |
| | Revised calculations were performed for the rerating using the current PTS Rule methodology and the latest procedure specified by Regulatory Guide 1.99-Revision 2. As stated in NRC Generic Letter 88-11, the staff is presently considering an amendment to the PTS Rule, 10 CFR 50.61, that will replace the equations for RTpts given in paragraph (b) (2) with the calculation procedure in Section C.l of Revision 2 to Regulatory Guide 1.99, but they will not change the screening criteria. |
| | All the RTpts values remain below the NRC screening values for PTS using the projected fluence values that are based upon rerated conditions through the license expiration. The highest RTpts value (265'F) was calculated at the circumferential weld of the Cook Nuclear Plant Unit I reactor vessel, using the methodology of Regulatory Guide 1.99 Revision 2. |
| | On the basis of probabi listic work described in the PTS Rule, the NRC staff that PHR vessels with conservatively calculated values of RTndt 'oncluded (i.e., RTpts) less than 270'F for plate material and axial welds, and less than 300'F for circumferential welds present an acceptably low risk of vessel failure from PTS events. This evaluation, however, did not take into account the impact of rerating, which causes potential PTS transient scenarios to begin from a lower system temperature. |
| | 81390:1d/092988 3.10"7 |
| | |
| | An initial |
| | ~ ~ ~ ~ |
| | evaluation was performed to determine the impact of rerating on the screening criteria in terms of risk of vessel |
| | ~ ~ ~ |
| | applicability of the PTS failure. A probabilistic fracture mechanics sensitivity study of limiting PTS transient characteristics, starting from a lower operating temperature, showed that the conditional probability of reactor vessel failure will not be adversely affected. Therefore, the overall risk of vessel failure will not be adversely impacted meaning that that the screening criteria in the PTS Rule are still applicable for the Cook Nuclear Plant Unit 1 reactor vessel relative to rerated conditions. |
| | 813S@: I d/092988 3.10-8 |
| | |
| | TABLE 3.10.1-1 FAST NEUTRON (E > 1.0 MeV) FLUENCE PROJECTIONS FOR COOK NUCLEAR PLANT UNIT 1 Unit 1 22.89 EFPY U er Bound Lower Bound All plates; Weld 9-442 1.84 x 10 1.55 x 10 Welds 2-442B, 2-442C, 1.19 x 10 1.01 x 10 3-442A, 3-442C Weids 2-442A, 3-442B 5.92 x 10 5.01 x 10 8139e:1d/092988 3.10-9 |
| | |
| | I TABLE 3.1.0.1-2 SURVEILLANCE CAPSULE LEAD FACTORS FOR COOK NUCLEAR PLANT UNIT 1 4'aosules 40'apsules Unit 1 Base Case 1.3 4.2 (3250 Wt, 536'F Downcomer) |
| | Unit 1 At Uprated Power 1.3 4.2 (3588 HMt, 547'F Downcomer) |
| | Unit 1 At Uprated Power 1.3 4' (3588 HMt, 512'F Downcomer) p ff QIPN Q C Jt j I2 P V~<7 8139e:1d/092988 3.10"10 |
| | |
| | E TABLE 4.1-1 SYSTEM DESIGN AND OPERATING PARAMETERS Plant design life, years 40 Number of heat cransfer loops Design pressure, psig 2485 Nominal operating pressure, psig 2235 Tocal system volume including pressurizer and surge line (ambient conditions), ft3 (estimated) 12,500 System liquid volume, including pressurizer and surge line (ambient conditions), fc 11,892 System liquid volume, including pressurizer max.,guaranteed power, fc 3 (escimac'ed) 11,780 6 |
| | Tocal Reactor heat output (100% power) Btu/hr 11,089 x 10 (Unit 1) |
| | (3250 MVt) i 6 11,641 x 10 (Unit 2) |
| | (3411 HPt) |
| | Unit 1 Unit 2 Bounding Conditions for Reracing Lower/Upper Reactor vessel coolant temperature at full power: |
| | Inlet, nominal, 0 F 514.9/545.2 541. 3 Outlet, nominal, 0 F 579.1/607.5 606.4 Coolant temperature rise in vessel ac full power, avg., 0 F 64.2/62.3 64.8 Total coolant flow race, lb/hr x 10 6 139.0/133.9 134.6 Sceam pressure ac full po~er, psia 618/820 820 Steam Temp. Q full power, F 489 '/521.1 521.1 Total Reactor Coolant Volume at a mbient condicions, ft3 12,438 12,470 4.1-25 July 1990}} |
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ML17326A0871999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Unit 1.With 990812 Ltr
ML17326A0861999-07-31031 July 1999
Monthly Operating Rept for July 1999 for DC Cook Nuclear Plant,Units 2.With 990812 Ltr
ML17326A0741999-07-29029 July 1999
LER 99-018-00:on 990629,determined That Valve Yokes May Yield Under Combined Stress of Seismic Event & Static,Valve Closed,Stem Thrust.Caused by Inadequate Design of Associated Movs.Operability Determinations Were Performed for Valves
ML17326A0661999-07-26026 July 1999
LER 99-017-00:on 990625,noted That Improperly Installed Fuel Oil Return Relief Valve Rendered EDG Inoperable.Caused by Personnel Error.Fuel Oil Return Valve Was Replaced with Valve in Correct Orientation.With 990722 Ltr
ML17326A0651999-07-22022 July 1999
LER 98-014-03:on 980310,noted That Response to high-high Containment Pressure Procedure Was Not Consistent with Analysis of Record.Caused by Inadequate Interface with W. FRZ-1 Will Be Revised to Be Consistent with New Analysis
ML17326A0491999-07-13013 July 1999
LER 99-016-00:on 990615,TS Requirements for Source Range Neutron Flux Monitors Not Met.Caused by Failure to Understand Design Basis of Plant.Procedures Revised.With 990713 Ltr
ML17326A0331999-07-0101 July 1999
LER 99-004-01:on 971030,failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed Was Noted.Caused by Ineffective Mgt of Tss.Chemistry Personnel Have Been Instructed on Requirement to Follow TS as Written
ML17326A0511999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 2.With 990709 Ltr
ML17326A0501999-06-30030 June 1999
Monthly Operating Rept for June 1999 for DC Cook Nuclear Plant,Unit 1.With 990709 Ltr
ML17326A0151999-06-18018 June 1999
LER 99-014-00:on 990521,determined That Boron Injection Tank Manway Bolts Were Not Included in ISI Program,Creating Missed Exam for Previous ISI Interval.Caused by Programmatic Weakness.Isi Program & Associated ISI Database Modified
ML17325B6421999-06-0101 June 1999
LER 99-013-00:on 990327,safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Have Led to ECCS Pump Failure.Caused by Inadequate Original Design Application of Si.Throttle Valves Will Be Developed
ML17325B6311999-06-0101 June 1999
LER 99-S03-00:on 990430,vital Area Barrier Degradation Was Noted.Caused by Inadequate Insp & Maint of Vital Area Barrier.Repairs & Mods Were Made to Barriers to Eliminate Degraded & Nonconforming Conditions
ML17326A0061999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Dcp.With 990609 Ltr
ML17326A0071999-05-31031 May 1999
Monthly Operating Rept for May 1999 for DC Cook Nuclear Plant,Unit 2.With 990609 Ltr
ML17325B6351999-05-28028 May 1999
LER 99-S02-00:on 990428,vulnerability in Safeguard Sys That Could Allow Unauthorized Access to Protected Area Was Noted. Caused by Inadequate Original Plant Design.Mods Were Made to Wall Opening to Eliminate Nonconforming Conditions
ML17265A8231999-05-24024 May 1999
LER 98-037-01:on 990422,determined That Ice Condenser Bypass Leakage Exceeds Design Basis Limit.Caused by Pressure Seal Required by Revised W Design Not Incorporated Into Aep Design.Numerous Matl Condition Walkdowns & Assessments Made
ML17325B6001999-05-20020 May 1999
LER 99-012-00:on 990420,concluded That Auxiliary Bldg ESF Ventilation Sys Not Capable of Maintaining ESF Room Temps post-accident.Caused by Inadequate Control of Sys Design Inputs.Comprehensive Action Plan Being Developed
ML17325B5861999-05-10010 May 1999
LER 99-002-00:on 990415,discovered That TS 4.0.5 Requirements Were Not Met Due to Improperly Performed Test. Caused by Incorrect Interpretation of ASME Code.App J Testing Will Be Completed & Procedures Will Be Revised
ML17325B5811999-05-0404 May 1999
LER 99-011-00:on 990407,air Sys for EDG Will Not Support Long Operability.Caused by Original Design Error.Temporary Mod to Supply Makeup Air Capability in Modes 5 & 6 Was Prepared
ML17325B5771999-05-0303 May 1999
LER 99-010-00:on 990401,RCS Leak Detection Sys Sensitivity Not in Accordance with Design Requirements Occurred.Caused by Inadequate Original Design of Containment Sump Level. Evaluation Will Be Performed to Clearly Define Design
ML17335A5301999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 1.With 990508 Ltr
ML17335A5291999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for DC Cook Nuclear Plant,Unit 2.With 990508 Ltr
ML17325B5581999-04-16016 April 1999
LER 99-006-00:on 990115,personnel Identified Discrepancy Between TS 3.9.7 Impact Energy Limit & Procedure 12 Ohp 4030.STP.046.Caused by Lack of Design Basis Control.Placed Procedure 12 Ohp 4030.STP.046 on Administrative Hold
ML17325B5471999-04-12012 April 1999
LER 99-009-00:on 990304,as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit Occurred.Cause Investigation for Condition Has Not Been Completed.Update to LER Will Be Submitted,Upon Completion of Investigation
ML17325B5321999-04-0707 April 1999
LER 99-S01-00:on 990308,discovered That Lock for Vital Gate Leading to Plant 4KV Switchgear Area Was Nonconforming & Vulnerable to Unauthorized Access.Caused by Inadequate Gate Design & Inadequate Procedures.Mods Are Being Made to Gate
ML17325B5161999-04-0101 April 1999
LER 99-007-00:on 981020,calculations Showed That Divider Barrier Between Upper & Lower Containment Vols Were Overstressed.Engineers Are Currently Working on Analyses of Loads & Stress on Enclosures
ML17325B5491999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant Unit 2.With 990408 Ltr
ML17325B5441999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for DC Cook Nuclear Plant,Unit 1.With 990408 Ltr
ML17325B5221999-03-29029 March 1999
LER 99-001-00:on 960610,degraded Component Cooling Water Flow to Containment Main Steam Line Penetrations,Identified on 990226.Caused by Inadequate Understanding of Design Basis.Additional Investigations Ongoing
ML17325B4801999-03-18018 March 1999
LER 99-004-00:on 971030,failure to Perform TS Surveillance Analyses of Rc Chemistry with Fuel Removed Was Noted.Cause of Event Is Under Investigation.Corrected Written Job Order Activities Used to Control SD Chemistry Sampling
ML17325B4741999-03-18018 March 1999
LER 99-005-00:on 940512,determined That Rt Breaker Manual Actuations During Rod Drop Testing Were Not Previously Reported.Caused by Lack of Training.Addl Corrective Actions,Including Preventative Actions May Be Developed
ML17325B5671999-03-0202 March 1999
Summary of Unit 1 Steam Generator Layup Chemistry from 980101 to 990218.
ML17325B4631999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Power Station,Unit 2.With 990308 Ltr
ML17325B4621999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for DC Cook Nuclear Plant,Unit 1.With 990308 Ltr
ML17325B4571999-02-24024 February 1999
LER 99-003-00:on 990107,CR Pressurization Sys Surveillance Test Did Not Test Sys in Normal Operating Condition.Caused by Failure to Recognize Door 12DR-AUX415 as Part of CR Pressure Boundary.Performed Walkdown of Other Doors
1999-09-30
[Table view] |
Text
MESTINGHOUSE CLASS III HCAP-11902 REDUCED TEMPERATURE ANO PRESSURE OPERATION FOR DONALD C. COOK NUCLEAR PLANT UNIT 1 LICENSING REPORT
- 0. L. Cecchett
- 0. B. Augustine October 1988 WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit P.O. Box 355 Pittsburgh, Pennsy1vania 15230 e pop /g
/'980a:1d/100588
c umulative fatigue usage factor was calculated to be 0.69 which is less than the 1.0 Code limit.
This value includes the thermal stress due to the non-linear portion of the thermal gradient for conservatism. A substantial reduction in the maximum range of stress intensity could be achieved if the non-linear thermal gradient contribution were neglected as permitted by the ASHE Code.
3.10.1.2 Reactor Vessel Integrity An evaluation of the impact of rerating on reactor vessel integrity for neutron embrittlement was performed. Neutron fluence changes for the rerating were calculated. Using these revised fluences and those from other relevant systems parameters associated with the rerating, the assessment included review of surveillance capsule withdrawal schedules, the schedule of applicability of the plant heatup and cooldown limits, and 10 CFR 50-
~
Appendix G analyses, including a verification of plant specific material
~ ~ ~ ~ ~ ~
properties. A revision to the calculations used in the submittal to the NRC
~
~ ~ ~
for meeting- the requirements of the pressurized thermal shock (PTS) rule were
~
performed. Finally, an initial evaluation of the impact of rerating on the PTS risk of vessel failure was carried out to confirm the applicability of the screening criteria in the PTS rule for the Cook Nuclear Plant reactor vessels. The rerating affects the PTS transient initiating temperature which is lower than that used in the generic PTS risk analyses which support the screening criteria.
~
Neutron transport calculations have been completed at a conservatively high power level of 3600 HWt, and reduced temperature/pressure conditions for Cook Nuclear Plant Unit 1. Results of these analyses have indicated that in addition to the increased reactor power, operation of the units with reduced coolant temperatures, particularly in the downcomer regions, also has a significant impact on the fast neutron exposure rates incident on the pressure vessel. Also impacted are the relationships among the neutron exposure rates at surveillance capsule locations and those at positions within the pressure vessel wall; i.e., capsule lead factors.
Agd~~ 9 tz p gQg7 8739a:1d/092988 3.10-5
~ I To provide a bounding evaluation of these various effects, calculations were performed for the current licensed conditions as well as for uprated power with both maximum and minimum downcomer temperatures. The results of these studies are summarized in Tables 3.10.1-1 and 3.10.1-2.
An examination of Tables 3.10.1-1 and 3.10.1-2 shows that the impact of higher power operation and changed coolant temperatures is greater on the absolute magnitude of the neutron exposure rate than on the relative behavior of neutron flux distributions as reflected in capsule lead factors.
Two sets of data have been provided to establish upper and lower bound exposures for Cook Nuclear Plant Unit 1. The upper limit conditions are consistent with assuming an uprating at the onset of the next fuel cycle with continued operation at an inlet temperature of 547'F for unit 1. Lower bound fluence estimates were based on no uprating in power level, but with future operation occurring at reduced T ; i.e., the inlet temperature was assumed avg'o be 512'F. These lower bound fluence estimates are lower than those at current operating conditions.
Review of the heatup and cooldown curves that were previously generated by Southwest Research Institute indicate that these curves were generated in accordance with Regulatory Guide 1.99-Revision 1. Per NRC Generic Letter 88-11, dated July 12, 1988, all utilities must submit to the NRC by November 1988 the results of their technical analysis relative to the implementation of Regulatory Guide 1.99-Revision 2, which was officially issued in May 1988.
Given this regulatory change, the effect of rerating should be incorporated II into future calculations that will be performed for revising heatup and cooldown curves in accordance with this latest revision of Regulatory Guide 1.99. The effect of the rerating is deemed to not be significant.
The changes in the systems parameters associated with the rerating have been judged to not have any significant impact on 10 CFR Part 50 - Appendix G analysis.
IZ pa~ +g 7 8139e:1d/093088 3.10-6
S The Cook Nuclear Plant Unit 1 reactor vessel beltline region material properties were verified against the latest available information in various industry data. bases and surveillance capsule reports. The properties defined from the latest information are consistent with those used in prior utility submittals to the NRC relative to meeting the requirements of the PTS rule.
Since the core loading pattern wi II be changing as a result of the rerating, an update to the PTS submittal will be required as stated in the PTS Rule.
Revised calculations were performed for the rerating using the current PTS Rule methodology and the latest procedure specified by Regulatory Guide 1.99-Revision 2. As stated in NRC Generic Letter 88-11, the staff is presently considering an amendment to the PTS Rule, 10 CFR 50.61, that will replace the equations for RTpts given in paragraph (b) (2) with the calculation procedure in Section C.l of Revision 2 to Regulatory Guide 1.99, but they will not change the screening criteria.
All the RTpts values remain below the NRC screening values for PTS using the projected fluence values that are based upon rerated conditions through the license expiration. The highest RTpts value (265'F) was calculated at the circumferential weld of the Cook Nuclear Plant Unit I reactor vessel, using the methodology of Regulatory Guide 1.99 Revision 2.
On the basis of probabi listic work described in the PTS Rule, the NRC staff that PHR vessels with conservatively calculated values of RTndt 'oncluded (i.e., RTpts) less than 270'F for plate material and axial welds, and less than 300'F for circumferential welds present an acceptably low risk of vessel failure from PTS events. This evaluation, however, did not take into account the impact of rerating, which causes potential PTS transient scenarios to begin from a lower system temperature.
81390:1d/092988 3.10"7
An initial
~ ~ ~ ~
evaluation was performed to determine the impact of rerating on the screening criteria in terms of risk of vessel
~ ~ ~
applicability of the PTS failure. A probabilistic fracture mechanics sensitivity study of limiting PTS transient characteristics, starting from a lower operating temperature, showed that the conditional probability of reactor vessel failure will not be adversely affected. Therefore, the overall risk of vessel failure will not be adversely impacted meaning that that the screening criteria in the PTS Rule are still applicable for the Cook Nuclear Plant Unit 1 reactor vessel relative to rerated conditions.
813S@: I d/092988 3.10-8
TABLE 3.10.1-1 FAST NEUTRON (E > 1.0 MeV) FLUENCE PROJECTIONS FOR COOK NUCLEAR PLANT UNIT 1 Unit 1 22.89 EFPY U er Bound Lower Bound All plates; Weld 9-442 1.84 x 10 1.55 x 10 Welds 2-442B, 2-442C, 1.19 x 10 1.01 x 10 3-442A, 3-442C Weids 2-442A, 3-442B 5.92 x 10 5.01 x 10 8139e:1d/092988 3.10-9
I TABLE 3.1.0.1-2 SURVEILLANCE CAPSULE LEAD FACTORS FOR COOK NUCLEAR PLANT UNIT 1 4'aosules 40'apsules Unit 1 Base Case 1.3 4.2 (3250 Wt, 536'F Downcomer)
Unit 1 At Uprated Power 1.3 4.2 (3588 HMt, 547'F Downcomer)
Unit 1 At Uprated Power 1.3 4' (3588 HMt, 512'F Downcomer) p ff QIPN Q C Jt j I2 P V~<7 8139e:1d/092988 3.10"10
E TABLE 4.1-1 SYSTEM DESIGN AND OPERATING PARAMETERS Plant design life, years 40 Number of heat cransfer loops Design pressure, psig 2485 Nominal operating pressure, psig 2235 Tocal system volume including pressurizer and surge line (ambient conditions), ft3 (estimated) 12,500 System liquid volume, including pressurizer and surge line (ambient conditions), fc 11,892 System liquid volume, including pressurizer max.,guaranteed power, fc 3 (escimac'ed) 11,780 6
Tocal Reactor heat output (100% power) Btu/hr 11,089 x 10 (Unit 1)
(3250 MVt) i 6 11,641 x 10 (Unit 2)
(3411 HPt)
Unit 1 Unit 2 Bounding Conditions for Reracing Lower/Upper Reactor vessel coolant temperature at full power:
Inlet, nominal, 0 F 514.9/545.2 541. 3 Outlet, nominal, 0 F 579.1/607.5 606.4 Coolant temperature rise in vessel ac full power, avg., 0 F 64.2/62.3 64.8 Total coolant flow race, lb/hr x 10 6 139.0/133.9 134.6 Sceam pressure ac full po~er, psia 618/820 820 Steam Temp. Q full power, F 489 '/521.1 521.1 Total Reactor Coolant Volume at a mbient condicions, ft3 12,438 12,470 4.1-25 July 1990
