|
|
| Line 17: |
Line 17: |
|
| |
|
| =Text= | | =Text= |
| {{#Wiki_filter:MESTINGHOUSE CLASSIIIHCAP-11902 REDUCEDTEMPERATURE ANOPRESSUREOPERATION FORDONALDC.COOKNUCLEARPLANTUNIT1LICENSING REPORT0.L.Cecchett0.B.Augustine October1988eWESTINGHOUSE ELECTRICCORPORATION EnergySystemsBusinessUnitP.O.Box355Pittsburgh, Pennsy1vania 15230pop/g/'980a:1d/100588 cumulative fatigueusagefactorwascalculated tobe0.69whichislessthanthe1.0Codelimit.Thisvalueincludesthethermalstressduetothenon-linear portionofthethermalgradientforconservatism. | | {{#Wiki_filter:MESTINGHOUSE CLASS II I 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 e WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit P.O.Box 355 Pittsburgh, Pennsy1vania 15230 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. |
| Asubstantial reduction inthemaximumrangeofstressintensity couldbeachievedifthenon-linear thermalgradientcontribution wereneglected aspermitted bytheASHECode.3.10.1.2ReactorVesselIntegrity Anevaluation oftheimpactofreratingonreactorvesselintegrity forneutronembrittlement wasperformed.
| | 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. |
| Neutronfluencechangesforthereratingwerecalculated.
| | Neutron fluence changes for the rerating were calculated. |
| Usingtheserevisedfluencesandthosefromotherrelevantsystemsparameters associated withthererating, theassessment includedreviewofsurveillance capsulewithdrawal schedules, thescheduleofapplicability oftheplantheatupandcooldownlimits,and10CFR50-~~~~~~~~~~~~AppendixGanalyses, including averification ofplantspecificmaterialproperties.
| | 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. |
| Arevisiontothecalculations usedinthesubmittal totheNRCformeeting-therequirements ofthepressurized thermalshock(PTS)rulewereperformed.
| | 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,aninitialevaluation oftheimpactofreratingonthePTSriskofvesselfailurewascarriedouttoconfirmtheapplicability ofthescreening criteriainthePTSrulefortheCookNuclearPlantreactorvessels.ThereratingaffectsthePTStransient initiating temperature whichislowerthanthatusedinthegenericPTSriskanalyseswhichsupportthescreening criteria. | | 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 |
| ~Neutrontransport calculations havebeencompleted ataconservatively highpowerlevelof3600HWt,andreducedtemperature/pressure conditions forCookNuclearPlantUnit1.Resultsoftheseanalyseshaveindicated thatinadditiontotheincreased reactorpower,operation oftheunitswithreducedcoolanttemperatures, particularly inthedowncomer regions,alsohasasignificant impactonthefastneutronexposureratesincidentonthepressurevessel.Alsoimpactedaretherelationships amongtheneutronexposureratesatsurveillance capsulelocations andthoseatpositions withinthepressurevesselwall;i.e.,capsuleleadfactors.Agd~~9tzpgQg78739a:1d/092988 3.10-5 | | ~I T o 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. |
| ~I Toprovideaboundingevaluation ofthesevariouseffects,calculations wereperformed forthecurrentlicensedconditions aswellasforupratedpowerwithbothmaximumandminimumdowncomer 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. |
| Theresultsofthesestudiesaresummarized inTables3.10.1-1and3.10.1-2.
| | 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. |
| Anexamination ofTables3.10.1-1and3.10.1-2showsthattheimpactofhigherpoweroperation andchangedcoolanttemperatures isgreaterontheabsolutemagnitude oftheneutronexposureratethanontherelativebehaviorofneutronfluxdistributions asreflected incapsuleleadfactors.Twosetsofdatahavebeenprovidedtoestablish upperandlowerboundexposures forCookNuclearPlantUnit1.Theupperlimitconditions areconsistent withassuminganupratingattheonsetofthenextfuelcyclewithcontinued operation ataninlettemperature of547'Fforunit1.Lowerboundfluenceestimates werebasedonnoupratinginpowerlevel,butwithfutureoperation occurring atreducedT;i.e.,theinlettemperature wasassumedavg'obe512'F.Theselowerboundfluenceestimates arelowerthanthoseatcurrentoperating conditions.
| | 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.8139e:1d/093088 3.10-6 IZ pa~+g 7 S |
| Reviewoftheheatupandcooldowncurvesthatwerepreviously generated bySouthwest ResearchInstitute indicatethatthesecurvesweregenerated inaccordance withRegulatory Guide1.99-Revision 1.PerNRCGenericLetter88-11,datedJuly12,1988,allutilities mustsubmittotheNRCbyNovember1988theresultsoftheirtechnical analysisrelativetotheimplementation ofRegulatory Guide1.99-Revision 2,whichwasofficially issuedinMay1988.Giventhisregulatory change,theeffectofreratingshouldbeincorporated IIintofuturecalculations thatwillbeperformed forrevisingheatupandcooldowncurvesinaccordance withthislatestrevisionofRegulatory Guide1.99.Theeffectofthereratingisdeemedtonotbesignificant.
| | 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. |
| Thechangesinthesystemsparameters associated withthereratinghavebeenjudgedtonothaveanysignificant impacton10CFRPart50-AppendixGanalysis.
| | 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'oncluded that PHR vessels with conservatively calculated values of RTndt (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. |
| 8139e:1d/093088 3.10-6IZpa~+g7 S | | 81390:1d/092988 3.10"7 An initial evaluation was performed to determine the impact of rerating on the~~~~~~~applicability of the PTS screening criteria in terms of risk of vessel 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. |
| TheCookNuclearPlantUnit1reactorvesselbeltlineregionmaterialproperties wereverifiedagainstthelatestavailable information invariousindustrydata.basesandsurveillance capsulereports.Theproperties definedfromthelatestinformation areconsistent withthoseusedinpriorutilitysubmittals totheNRCrelativetomeetingtherequirements ofthePTSrule.SincethecoreloadingpatternwiIIbechangingasaresultofthererating, anupdatetothePTSsubmittal willberequiredasstatedinthePTSRule.Revisedcalculations wereperformed forthereratingusingthecurrentPTSRulemethodology andthelatestprocedure specified byRegulatory Guide1.99-Revision2.AsstatedinNRCGenericLetter88-11,thestaffispresently considering anamendment tothePTSRule,10CFR50.61,thatwillreplacetheequations forRTptsgiveninparagraph (b)(2)withthecalculation procedure inSectionC.lofRevision2toRegulatory Guide1.99,buttheywillnotchangethescreening criteria.
| | 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 All plates;Weld 9-442 Welds 2-442B, 2-442C, 3-442A, 3-442C U er Bound 1.84 x 10 1.19 x 10 Lower Bound 1.55 x 10 1.01 x 10 Weids 2-442A, 3-442B 5.92 x 10 5.01 x 10 8139e:1d/092988 3.10-9 I |
| AlltheRTptsvaluesremainbelowtheNRCscreening valuesforPTSusingtheprojected fluencevaluesthatarebaseduponreratedconditions throughthelicenseexpiration.
| | TABLE 3.1.0.1-2 SURVEILLANCE CAPSULE LEAD FACTORS FOR COOK NUCLEAR PLANT UNIT 1 4'aosules 40'apsules Unit 1 Base Case (3250 Wt, 536'F Downcomer) 1.3 4.2 Unit 1 At Uprated Power (3588 HMt, 547'F Downcomer) 1.3 4.2 Unit 1 At Uprated Power (3588 HMt, 512'F Downcomer) 1.3 4'8139e:1d/092988 3.10"10 p ff Q QIPN C Jt j I 2 P V~<7 E |
| ThehighestRTptsvalue(265'F)wascalculated atthecircumferential weldoftheCookNuclearPlantUnitIreactorvessel,usingthemethodology ofRegulatory Guide1.99Revision2.Onthebasisofprobabilisticworkdescribed inthePTSRule,theNRCstaff'oncluded thatPHRvesselswithconservatively calculated valuesofRTndt(i.e.,RTpts)lessthan270'Fforplatematerialandaxialwelds,andlessthan300'Fforcircumferential weldspresentanacceptably lowriskofvesselfailurefromPTSevents.Thisevaluation, however,didnottakeintoaccounttheimpactofrerating, whichcausespotential PTStransient scenarios tobeginfromalowersystemtemperature.
| | TABLE 4.1-1 SYSTEM DESIGN AND OPERATING PARAMETERS i Plant design life, years Number of heat cransfer loops Design pressure, psig Nominal operating pressure, psig Tocal system volume including pressurizer and surge line (ambient conditions), ft (estimated) 3 System liquid volume, including pressurizer and surge line (ambient conditions), fc System liquid volume, including pressurizer 3 max.,guaranteed power, fc (escimac'ed) |
| 81390:1d/092988 3.10"7 Aninitialevaluation wasperformed todetermine theimpactofreratingonthe~~~~~~~applicability ofthePTSscreening criteriaintermsofriskofvesselfailure.Aprobabilistic fracturemechanics sensitivity studyoflimitingPTStransient characteristics, startingfromaloweroperating temperature, showedthattheconditional probability ofreactorvesselfailurewillnotbeadversely affected. | | Tocal Reactor heat output (100%power)Btu/hr 40 2485 2235 12,500 11,892 11,780 11,089 x 10 (Unit 1)6 (3250 MVt)11,641 x 10 (Unit 2)6 (3411 HPt)Unit 1 Bounding Conditions for Reracing Lower/Upper Unit 2 Reactor vessel coolant temperature at full power: Inlet, nominal, F 0 Outlet, nominal, F 0 Coolant temperature rise in vessel 0 ac full power, avg., F Total coolant flow race, lb/hr x 10 6 Sceam pressure ac full po~er, psia Steam Temp.Q full power, F Total Reactor Coolant Volume at a mbient condicions, ft 3 514.9/545.2 579.1/607.5 64.2/62.3 139.0/133.9 618/820 489'/521.1 12,438 541.3 606.4 64.8 134.6 820 521.1 12,470 4.1-25 July 1990}} |
| Therefore, theoverallriskofvesselfailurewillnotbeadversely impactedmeaningthatthatthescreening criteriainthePTSRulearestillapplicable fortheCookNuclearPlantUnit1reactorvesselrelativetoreratedconditions. | |
| 813S@:Id/0929883.10-8 TABLE3.10.1-1FASTNEUTRON(E>1.0MeV)FLUENCEPROJECTIONS FORCOOKNUCLEARPLANTUNIT1Unit122.89EFPYAllplates;Weld9-442Welds2-442B,2-442C,3-442A,3-442CUerBound1.84x101.19x10LowerBound1.55x101.01x10Weids2-442A,3-442B5.92x105.01x108139e:1d/092988 3.10-9 I | |
| TABLE3.1.0.1-2 SURVEILLANCE CAPSULELEADFACTORSFORCOOKNUCLEARPLANTUNIT14'aosules 40'apsules Unit1BaseCase(3250Wt,536'FDowncomer) 1.34.2Unit1AtUpratedPower(3588HMt,547'FDowncomer) 1.34.2Unit1AtUpratedPower(3588HMt,512'FDowncomer) 1.34'8139e:1d/092988 3.10"10pffQQIPNCJtjI2PV~<7 E
| |
| TABLE4.1-1SYSTEMDESIGNANDOPERATING PARAMETERS iPlantdesignlife,yearsNumberofheatcransferloopsDesignpressure, psigNominaloperating
| |
| : pressure, psigTocalsystemvolumeincluding pressurizer andsurgeline(ambientconditions),
| |
| ft(estimated) 3Systemliquidvolume,including pressurizer andsurgeline(ambientconditions), | |
| fcSystemliquidvolume,including pressurizer 3max.,guaranteed power,fc(escimac'ed)
| |
| TocalReactorheatoutput(100%power)Btu/hr402485223512,50011,89211,78011,089x10(Unit1)6(3250MVt)11,641x10(Unit2)6(3411HPt)Unit1BoundingConditions forReracingLower/Upper Unit2Reactorvesselcoolanttemperature atfullpower:Inlet,nominal,F0Outlet,nominal,F0Coolanttemperature riseinvessel0acfullpower,avg.,FTotalcoolantflowrace,lb/hrx106Sceampressureacfullpo~er,psiaSteamTemp.Qfullpower,FTotalReactorCoolantVolumeatambientcondicions, ft3514.9/545.2 579.1/607.5 64.2/62.3 139.0/133.9 618/820489'/521.112,438541.3606.464.8134.6820521.112,4704.1-25July1990}}
| |
Similar Documents at Cook |
|---|
Category:REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)
MONTHYEARML17335A5481999-09-30030 September 1999
Non-proprietary DC Cook Nuclear Plant Units 1 & 2 Mods to Containment Sys W SE (Secl 99-076,Rev 3).
ML17333A4941996-05-31031 May 1996
Rev 1 to DCP Unit 2 3,600 Mwt Uprating Program Licensing Program Rept.
ML17333A2071995-11-30030 November 1995
Nonproprietary Presentation Matls from 950810 Meeting W/ Aepsc,Nrc & Westinghouse Re Application of Revised Pressure Boundary Limits for Hej Sleeved Tubes.
ML17333A2221995-11-29029 November 1995
Nonproprietary Response to NRC Questions on Backup Functions Environ Allowance Terms.
ML17332A9921995-09-30030 September 1995
Repair Boundary for Parent Tube Indications within Upper Joint Zone of Hybrid Expansion Joint Sleeved Tubes.
ML17332A9471995-09-30030 September 1995
Rev 1 to DC Cook Unit 1 Steam Generator Tube Plugging Criteria for Indications at Tube Support Plates.
ML17331B2421994-02-28028 February 1994
Nonproprietary F* Tube Plugging Criterion for Tubes W/ Degradation in Tubesheet Roll Expansion Region of DC Cook Unit 1 Sgs.
ML17334B4681993-02-28028 February 1993
Evaluation of PTS for DC Cook Unit 2.
ML17329A4231992-03-31031 March 1992
Nonproprietary DC Cook Unit 1 Steam Generator Tube Plugging Criteria for Indications at Tube Support Plates.
ML20029C1401991-01-31031 January 1991
Nonproprietary Structural Evaluation of DC Cook Nuclear Plant Units 1 & 2 Pressurizer Surge Line,Considering Effects of Thermal Stratification.
ML20043C0711990-05-31031 May 1990
Westinghouse Revised Thermal Design Procedure Instrument Uncertainty Methodolgy for American Electric Power DC Cook Unit 2 Nuclear Power Station.
ML17328A2941990-01-31031 January 1990
Analysis of Capsule U from American Electric Power Co DC Cook Unit 1 Reactor Vessel Radiation Surveillance Program.
ML17328A7281989-09-30030 September 1989
Suppl 1 to, Rerated Power & Revised Temp & Pressure Operation for DC Cook Nuclear Plant,Units 1 & 2,Licensing Rept.
ML17325B1371988-11-30030 November 1988
Aep Reactor Core Thermal-Hydraulic Analysis Using Cobra IIIC/MIT-2 Computer Code.
ML17325A9671988-10-31031 October 1988
Reduced Temp & Pressure Operation for Donald C. Cook Nuclear Plant Unit 1 Licensing Rept, Including FSAR mark-ups
ML20207G0071988-08-31031 August 1988
Rev 1 to Criticality Safety Analysis,Dc Cook Spent Fuel Storage Racks W/15 X 15 & 17 X 17 Fuel Enrichments Up to 5.0%
ML20207G0221988-07-31031 July 1988
Thermal-Hydraulic Analysis of DC Cook Spent Fuel Pool
ML17325A9081988-07-31031 July 1988
Containment Integrity Analysis for Donald C Cook Nuclear Plant Units 1 & 2.
ML20207F9981988-07-31031 July 1988
Rev 0 to Final Rept Criticality Safety Analysis,Dc Cook New Fuel Storage Vault W/5.0% Enriched 17 X 17 Fuel
ML20238A9621987-09-0202 September 1987
Investigation on Undervoltage Trip Attachments for Cause of Misoperation at DC Cook Unit 2 on 851029,EQ/P(86)-143
ML20207P8901987-01-15015 January 1987
Nonproprietary Rev 2 to DC Cook Unit 1 Limiting Break K(Z) Loca/Eccs Analysis
ML20214J9241986-09-30030 September 1986
Nonproprietary NRC Presentation Rept on Steam Generator Tube Integrity for DC Cook Unit 2,Sept 1986
ML17324A7541986-02-28028 February 1986
Nonproprietary American Electric Power DC Cook Unit 2 Rdf RTD Installation Safety Evaluation.
ML20151Z0461986-01-31031 January 1986
NRC Presentation Rept on Steam Generator Tube Integrity for DC Cook Unit 2
ML20154H9841985-12-31031 December 1985
Nonproprietary, Spray Additive Tank Deletion Analysis
ML20151U8871985-11-30030 November 1985
Limiting Break K(Z) Loca/Eccs Analysis
ML17326B0101985-07-31031 July 1985
Nonproprietary XN-NF-85-28, DC Cook Unit 2,Cycle 6 Sar.
ML20116K2131985-04-30030 April 1985
Suppl 2 to Rev 2 to DC Cook,Unit 2 Cycle 5 5% Steam Generator Tube Plugging Limiting Break LOCA-ECCS Analysis: K(Z) Curve
ML20100G5111985-04-0404 April 1985
Suppl 1 to Rev 2 to DC Cook Unit 2 Cycle 5 5% Steam Generator Tube Plugging Limiting Break Loca/Eccs Analysis: K(Z) Curve
ML20151R7131985-03-31031 March 1985
Rev 1 to Evaluation of Acceptability of Reactor Vessel Head Lift Rig,Reactor Vessel Internals Lift Rig,Load Cell & Load Cell Linkage to Requirements of NUREG-0612 for Indiana & Michigan Electric Co,Dc Cook 1 & 2
ML20113A4261985-02-28028 February 1985
Rev 0 to Suppl 4-NP to Westinghouse Technical Support Complex Design & Verification & Validation Process for DC Cook Nuclear Plant
ML20096C9241984-08-21021 August 1984
Mechanical Design Rept Suppl for DC Cook Unit 1 Extended Burnup Fuel Assemblies
ML17326B1341984-08-0707 August 1984
Nonproprietary Version of Rev 2 to DC Cook,Unit 2 Cycle 5,5%.Steam Generator Tube Plugging Limiting Break Loca/Eccs Analysis.
ML20090J2371984-07-17017 July 1984
Generic Mechanical Design Rept,Exxon 17x17 Fuel Assembly
ML20090K4811984-05-22022 May 1984
Nonproprietary Rev 1 to DC Cook Unit 2 Cycle 5 5% Steam Generator Tube Plugging Limiting Break Loca/Eccs Analysis
ML20081D0021984-03-31031 March 1984
Potential Radiological Consequences of Incidents Involving High Exposure Fuel,Dc Cook,Unit 2
ML17320A9461984-03-0303 March 1984
Nonproprietary Rev 2 to Plant Transient Analysis for DC Cook Unit 2 Reactor at 3,425 Mwt Operation W/5% Steam Generator Tube Plugging.
ML17320A9441984-03-0101 March 1984
Nonproprietary DC Cook Unit 2 Cycle 5 - 5% Steam Generator Tube Plugging Limiting Break Loca/Eccs Analysis.
ML17320A9411984-02-23023 February 1984
Suppl 1 to DC Cook Unit 2,Cycle 5 Sar.
ML17320A9451983-10-24024 October 1983
DC Cook Unit 2,Cycle 5 Sar.
ML20081M2851983-10-21021 October 1983
Exxon Nuclear DNB Correlation for PWR Fuel Designs
ML20095D2311983-08-0303 August 1983
DC Cook Unit 1 LOCA-ECCS Analysis for Extended Exposure
ML17320A5561983-04-30030 April 1983
Evaluation of Acceptability of Reactor Vessel Head Lift Rig,Reactor Vessel Internals Lift Rig,Load Cell & Load Cell Linkage to Requirements of NUREG-0612.
ML17319B6661982-11-18018 November 1982
Suppl 1 to DC Cook Unit 2 Cycle 4 Limiting Break LOCA-ECCS Analysis Using Exem/Pwr.
ML17319B6701982-11-17017 November 1982
DC Cook Unit 2 Potential Radiological Consequences of Incidents Involving High Exposure Fuel.
ML17319B4111982-04-30030 April 1982
DC Cook Unit 2,Cycle 4 Sar.
ML17319B4121982-04-30030 April 1982
DC Cook Unit 2 LOCA ECCS Analysis Using Exem/Pwr Large Break Results.
ML19350C3911981-02-12012 February 1981
LOCA ECCS Reanalysis for DC Cook Unit 1 Using ENC Wrem 11A PWR ECCS Evaluation Model.
ML17321A6611980-10-31031 October 1980
Rev 0 to Man-Machine Interface Design Basis Document:Info Coding for Computer Display Sys.
ML17329A5671975-11-30030 November 1975
American Electric Power Co DC Cook Unit 2 Reactor Vessel Radiation Surveillance Program.
1999-09-30
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17335A5641999-10-18018 October 1999
LER 99-024-00:on 990708,literal TS Requirements Were Not Met by Accumlator Valve Surveillance.Caused by Misjudgement Made in Conversion from Initial DC Cook TS to W Std Ts.Submitted License Amend Request.With 991018 Ltr
ML17335A5531999-10-0707 October 1999
LER 99-023-00:on 990907,inadequate TS Surveillance Testing of ESW Pump ESF Response Time Noted.Caused by Inadequate Understanding of Plant Design Basis.Surveillance Tests Will Be Revised & Implemented
ML17335A5631999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 1.With 991012 Ltr
ML17335A5621999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for DC Cook Nuclear Plant,Unit 2.With 991012 Ltr
ML17335A5481999-09-30030 September 1999
Non-proprietary DC Cook Nuclear Plant Units 1 & 2 Mods to Containment Sys W SE (Secl 99-076,Rev 3).
ML17335A5451999-09-28028 September 1999
Rev 1 to Containment Sump Level Design Condition & Failure Effects Analysis for Potential Draindown Scenarios.
ML17326A1291999-09-17017 September 1999
LER 99-022-00:on 990609,electrical Bus Degraded Voltage Setpoints Too Low for Safety Related Loads,Was Discovered. Caused by Lack of Understanding of Design of Plant.No Immediate Corrective Actions Necessary
ML17326A1481999-09-17017 September 1999
Independent Review of Control Rod Insertion Following Cold Leg Lbloca,Dc Cook,Units 1 & 2.
ML17326A1211999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 2.With 990915 Ltr
ML17326A1201999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Cook Nuclear Plant, Unit 1.With 990915 Ltr
ML17326A1121999-08-27027 August 1999
LER 99-021-00:on 990728,determined That GL 96-01 Test Requirements Were Not Met in Surveillance Tests.Caused by Failure to Understand Full Extent of GL Requirements. Surveillance Procedures Will Be Revised or Developed
ML17326A1011999-08-26026 August 1999
LER 99-020-00:on 990727,EDGs Were Declared Inoperable.Caused by Inadequate Protection of Air Intake,Exhaust & Room Ventilation Structures from Tornado Missile Hazards. Implemented Compensatory Measures in Form of ACs
ML17326A0911999-08-16016 August 1999
LER 99-019-00:on 990716,noted Victoreen Containment Hrrms Not Environmentally Qualified to Withstand post-LOCA Conditions.Caused by Inadequate Design Control.Reviewing Options to Support Hrrms Operability in Modes 1-4
ML17326A0771999-08-0404 August 1999
LER 98-029-01:on 980422,noted That Fuel Handling Area Ventilation Sys Was Inoperable.Caused by Original Design Deficiency.Radiological Analysis for Spent Fuel Handling Accidents in Auxiliary Bldg Will Be Redone by 990830
ML17335A5461999-08-0202 August 1999
Rev 0 to Evaluation of Cook Recirculation Sump Level for Reduced Pump Flow Rates.
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 II I 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 e WESTINGHOUSE ELECTRIC CORPORATION Energy Systems Business Unit P.O.Box 355 Pittsburgh, Pennsy1vania 15230 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 T o 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.8139e:1d/093088 3.10-6 IZ pa~+g 7 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'oncluded that PHR vessels with conservatively calculated values of RTndt (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~~~~~~~applicability of the PTS screening criteria in terms of risk of vessel 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 All plates;Weld 9-442 Welds 2-442B, 2-442C, 3-442A, 3-442C U er Bound 1.84 x 10 1.19 x 10 Lower Bound 1.55 x 10 1.01 x 10 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 (3250 Wt, 536'F Downcomer) 1.3 4.2 Unit 1 At Uprated Power (3588 HMt, 547'F Downcomer) 1.3 4.2 Unit 1 At Uprated Power (3588 HMt, 512'F Downcomer) 1.3 4'8139e:1d/092988 3.10"10 p ff Q QIPN C Jt j I 2 P V~<7 E
TABLE 4.1-1 SYSTEM DESIGN AND OPERATING PARAMETERS i Plant design life, years Number of heat cransfer loops Design pressure, psig Nominal operating pressure, psig Tocal system volume including pressurizer and surge line (ambient conditions), ft (estimated) 3 System liquid volume, including pressurizer and surge line (ambient conditions), fc System liquid volume, including pressurizer 3 max.,guaranteed power, fc (escimac'ed)
Tocal Reactor heat output (100%power)Btu/hr 40 2485 2235 12,500 11,892 11,780 11,089 x 10 (Unit 1)6 (3250 MVt)11,641 x 10 (Unit 2)6 (3411 HPt)Unit 1 Bounding Conditions for Reracing Lower/Upper Unit 2 Reactor vessel coolant temperature at full power: Inlet, nominal, F 0 Outlet, nominal, F 0 Coolant temperature rise in vessel 0 ac full power, avg., F Total coolant flow race, lb/hr x 10 6 Sceam pressure ac full po~er, psia Steam Temp.Q full power, F Total Reactor Coolant Volume at a mbient condicions, ft 3 514.9/545.2 579.1/607.5 64.2/62.3 139.0/133.9 618/820 489'/521.1 12,438 541.3 606.4 64.8 134.6 820 521.1 12,470 4.1-25 July 1990
