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Category:GENERAL EXTERNAL TECHNICAL REPORTS
MONTHYEARML17265A1361997-12-23023 December 1997
Rev 2 to RCS Pressure & Temp Limits Rept (Ptlr).
ML17264B1011997-08-29029 August 1997
Rev 0 to Leak-Before-Break Evaluation of Portions of RHR Sys at Re Ginna Nuclear Power Station.
ML17264A8681997-04-23023 April 1997
Rev 0 to Evaluation of Ginna RCS Coolant Temp to Support LTOPs Requirements.
ML17264A8511997-03-19019 March 1997
Rg&E Re Ginna Nuclear Power Plant Spent Fuel Pool Re-racking Licensing Rept.
ML17264A7931997-01-31031 January 1997
Rev 1 to Final Rept, Re Ginna Nuclear Power Plant Probabilistic Safety Assessment.
ML17264A6101996-09-23023 September 1996
Rev 0 to Design Analysis Operability Evaluation for 857 A/B/C Ginna Station.
ML17264A6121996-09-23023 September 1996
Rev 2 to Design Analysis Ginna Station Pressure Locking Evaluation for MOVs 852 A&B.
ML17309A6051996-09-13013 September 1996
Rev 2 to RCS Pressure & Temp Limits Rept (Ptlr).
ML17264A6791996-05-24024 May 1996
Rev 1 to RCS Pressure & Temperature Limits Rept (Ptlr).
ML17264A4001996-02-24024 February 1996
Rev 0 to RCS Pressure & Temp Limits Rept.
ML17264A2791995-12-0808 December 1995
Re Ginna NPP RCS Pressure & Temp Limits Rept Cycle 25, Draft B
ML17264A1051995-05-0404 May 1995
Rev 0 to Final Exam Rept for 1995 SG Eddy Current Insp at Ginna Nuclear Power Station, Dtd 950503
ML17263B0391995-04-18018 April 1995
Summary Exam Rept for 1995 SG Eddy Current Insp,Rev 0.
ML17264A3411995-03-15015 March 1995
Low Temp Overpressure Analysis Summary Rept.
ML17263A8351994-11-0707 November 1994
Rev 1 to Fission Product Barrier Evaluation.
ML17263A8331994-10-11011 October 1994
Rev 1 to Re Ginna EALs Technical Bases.
ML17263A8311994-09-26026 September 1994
Draft Rev C to Design Criteria Ginna Station Containment Structural Mods Wbs 4.
ML17263A7941994-09-15015 September 1994
Safety Evaluation of Ginna SG Replacement.
ML17059A3611994-07-0606 July 1994
Emergency Action Level Verification & Validation Rept.
ML17263B0481994-06-30030 June 1994
Criticality Analysis of Plant Fresh & Spent Fuel Racks & Consolidated Rod Storage Canisters.
ML17311A0181994-05-13013 May 1994
New York State EAL Upgrade Project Verification & Validation Rept.
ML17263A8291994-03-30030 March 1994
Draft Rev a to Safety Evaluation SEV-1019, Containment Structural Mods Wbs 4.
ML17263A4651993-05-17017 May 1993
Radial Displacement & Rebar Strain Measurements for EWR #5181,Rev A.
ML17262B1201992-11-30030 November 1992
Re Ginna Boric Acid Storage Tank Boron Concentration Reduction Study.
ML17262B0831992-07-31031 July 1992
Recommended Info for Inclusion in Section 15.6.4 of FSAR for Re Ginna Nuclear Plant.
ML17262A8391992-04-30030 April 1992
Rev 0 to Summary Exam Rept for 1992 SG Eddy Current Insp at Re Ginna Nuclear Power Station.
ML17262A5601991-06-18018 June 1991
Rev 1 to Design Verification Ginna Station Containment Foundation Mat Analysis.
ML17262A4691991-04-25025 April 1991
Rev 0 Summary Exam Rept for 1991 Steam Generator Eddy Current Insp.
ML17262A4521991-04-22022 April 1991
Control Room Heatup Analysis.
ML17262A3781991-02-28028 February 1991
Nonproprietary Re Ginna Low Temp Overpressure Protection Sys Setpoint Phase II Evaluation, Final Rept
ML17262A4141991-02-26026 February 1991
Safety Analysis,Ginna Station Updated FSAR Section 6.2.4 & Tables 6.2-13,6.2-14 & 6.2-15 Changes.
ML17262A3681991-02-15015 February 1991
Simulation Facility Certification Rept.
ML17262A4431990-10-0404 October 1990
Rev 0 to Design Analysis Ginna Station Containment Mat Design Water Level Elevation 265 ft,0 Inches.
ML17262A4401990-10-0404 October 1990
Rev 0 to Design Verification Ginna Station Containment Foundation Mat Analysis.
ML17262A1931990-10-0303 October 1990
Rev 1 to Safety Analysis Ginna Station Updated FSAR Table 6.2-13 Changes.
ML17262A1761990-08-30030 August 1990
Voltage Simulation for Case EOF LOC4 LOCA Simulation for 50/50 Mode - Circuit 767 Details 12B Transformer Feeding Bus 12B.
ML17262A1771990-07-27027 July 1990
Rev 1 to Design Analysis EWR 4525-1, Fault Current Analysis of Power Distribution Sys.
ML17262A1781990-07-24024 July 1990
Rev 1 to Design Analysis EWR 4525-2, Adequacy of Electric Sys Voltages.
ML17250B1761990-05-0808 May 1990
Rev 1 Summary Exam Rept for 1990 Steam Generator Eddy Current Insp.
ML17261B0201990-03-14014 March 1990
Design Criteria Ginna Station Steam Generator Containment Penetration.
ML17251A4811989-02-28028 February 1989
Ultrasonic Indication Sizing Technique Development. Related Info Encl
ML17251A4771988-06-17017 June 1988
Rev 0 to Differential Pressure Thrust Calculation Methodology.
ML17261A6571987-10-31031 October 1987
Steam Generator Tube Plugging Increase Licensing Rept for Ginna Nuclear Power Station.
ML17261A5521987-07-14014 July 1987
Supplemental Rept to Dcrdr Final Summary Rept for Re Ginna Station.
ML17251A4741987-04-0101 April 1987
Rev 0 to Safety Analysis,Ginna Station PORV Block Valves.
ML17251A4721987-03-10010 March 1987
Rev 0 to Design Criteria,Ginna Station PORV Block Valves Replacement.
ML17251A9191986-12-18018 December 1986
Rev 0 to Implementation Rept EWR 2799, Reactor Vessel Level Monitoring Sys.
ML17251A6171986-03-0101 March 1986
1986 Steam Generator Eddy Current Exam Summary Rept.
ML17254A7031985-12-31031 December 1985
Vols 1 & 2 to Dcrdr Final Summary Rept Program Implementation,Re Ginna Nuclear Power Plant.
ML17254A6911985-12-16016 December 1985
Reinforced Masonry Wall Evaluation,Evaluation of Control Bldg Reinforced Walls.
1997-08-29
[Table view]
Category:TEXT-SAFETY REPORT
MONTHYEARML17265A7601999-10-0505 October 1999
Part 21 Rept Re W2 Switch Supplied by W Drawn from Stock, Did Not Operate Properly After Being Installed on 990409. Switch Returned to W on 990514 for Evaluation & Root Cause Analysis
ML17265A7621999-09-30030 September 1999
Monthly Operating Rept for Sept 1999 for Re Ginna Npp.With 991008 Ltr
ML17265A7531999-09-23023 September 1999
Part 21 Rept Re Corrective Action & Closeout of 10CFR21 Rept of Noncompliance Re Unacceptable Part for 30-4 Connector. Unacceptable Parts Removed from Stock & Scrapped
ML17265A7541999-09-22022 September 1999
LER 99-011-00:on 990823,small Tears Were Discovered in Flexible Duct Work Connector at Inlet of CR HVAC Sys Return Air Fan (AKF08).Caused by in-leakage Greater than That Assumed.Implemented Temporary Mod 99-029.With 990922 Ltr
ML17265A7471999-08-31031 August 1999
Monthly Operating Rept for Aug 1999 for Re Ginna Npp.With 990909 Ltr
ML17265A7431999-08-24024 August 1999
LER 99-004-01:on 990412,discovered That Containment Recirculation Fan Chevron Separator Vanes Were Installed Backwards.Caused by Improper Assembly by Mfg.Moisture Separator Vanes Were Dismantled & Correctly re-installed
ML17265A7341999-07-31031 July 1999
Monthly Operating Rept for July 1999 for Re Ginna Npp.With 990806 Ltr
ML17265A7291999-07-29029 July 1999
Interim Part 21 Rept Re safety-related DB-25 Breaker Mechanism Procured from W Did Not Pas Degradatin Checks When Drawn from Stock to Be Installed Into BUS15/03A.Holes Did Not line-up & Tripper Pan Bent
ML17265A7181999-07-23023 July 1999
LER 99-007-01:on 990423,reactor Trip Occurred Due to Instrument & Control Technicians Inadvertently Pulling Fuses from Wrong Nuclear Instrument Channel.Setpoint Adjustments Were Completed by Different Crew of Technicians
ML17265A7081999-07-22022 July 1999
LER 98-003-02:on 980904,actuations of CR Emergency Air Treatment Sys Was Noted Due to Invalid Causes.Caused by Various Degraded Components in CR RM Sys.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored
ML17265A7131999-07-22022 July 1999
Special Rept:On 990407,radiation Monitor RM-14A Was Declared Inoperable.Caused by Failed Communication Link from TSC to Plant Process Computer Sys.Communication Link Was re-established & RM-14A Was Declaed Operable on 990521
ML17265A7031999-07-19019 July 1999
LER 99-S01-00:on 990617,determined That Temporary Unescorted Access Had Been Granted to Contractor Employee.Caused by Incomplete Info Re Circumstances of Individual Military Separation.Individual Access Was Revoked.With 990719 Ltr
ML17265A7211999-07-19019 July 1999
ISI Rept for Third Interval (1990-1999) Third Period, Second Outage (1999) at Re Ginna Npp.
ML17265A7021999-07-15015 July 1999
LER 99-010-00:on 990615,ventilation Isolation of Auxiliary Bldg Occurred When Auxiliary Bldg Gas Radiation Monitor R-14 Reached High Alarm Setpoint.Cr Operators Rest Auxiliary Bldg Ventilation Isolation Signal.With 990715 Ltr
ML17265A7661999-06-30030 June 1999
1999 Rept of Facility Changes,Tests & Experiments Conducted Without Prior NRC Approval for Jan 1998 Through June 1999, Per 10CFR50.59.With 991020 Ltr
ML17265A7011999-06-30030 June 1999
Monthly Operating Rept for June 1999 for Re Ginna Npp.With 990712 Ltr
ML17265A6851999-06-21021 June 1999
LER 99-001-01:on 990222,deficiencies in NSSS Vendor steam- Line Brake Mass & Energy Release Analysis Results in Plant Being Outside Design Bases Occurred.Caused by Deficiencies in W.Temporary Administrative Replaced.With 990621 Ltr
ML17265A6761999-06-16016 June 1999
Part 21 Rept Re Defects & noncompliances,10CFR21(d)(3)(ii), Which Requires Written Notification to NRC on Identification of Defect or Failure to Comply. Relays Were Returned to Eaton for Evaluation & Root Cause Analysis
ML17265A6661999-06-0202 June 1999
LER 99-009-00:on 990503,instrumentation Declared Inoperable in Multiple Channels Resulted in Condition Prohibited by Ts. Caused by Unanticipated High Frequency AC Voltage Ripple. Entered TS LCO 3.0.3.With 990602 Ltr
ML17265A6681999-05-31031 May 1999
Monthly Operating Rept for May 1999 for Re Ginna Nuclear Power Plant.With 990608 Ltr
ML17265A6651999-05-27027 May 1999
Interim Rept Re W2 Control Switch,Procured from W,Did Not Operate Satisfactorily When Drawn from Stock to Be Installed in Main Control Board for 1C2 Safety Injection Pump. Estimated That Evaluation Will Be Completed by 991001
ML17309A6541999-05-27027 May 1999
LER 99-008-00:on 990427,overtemperature Delta T Reactor Trip Occurred Due to Faulted Bistable During Calibr of Redundant Channel.Plant Was Stabilized in Mode 3 & Faulted Bistable Was Subsequently Replaced.With 990527 Ltr
ML17265A6631999-05-24024 May 1999
LER 99-007-00:on 990423,technicians Inadvertently Pulled Fuses from Wrong Nuclear Instrument Cahnnel,Causing Reactor Trip,Due to High Range Flux Trip.Caused by Personnel Error. Labeling Scheme Improved
ML17265A6601999-05-21021 May 1999
LER 99-006-00:on 990421,start of turbine-driven Auxiliary Feedwater Pump Was Noted.Caused by MOV Being Left in Open Position.Closed Manual Isolation Valve to Secure Steam to Pump.With 990521 Ltr
ML17265A6591999-05-17017 May 1999
Part 21 Rept Re Relay Deficiency Detected During pre-installation Testing.Caused by Incorrectly Wired Relay Coil.Relays Were Returned to Eaton Corp for Investigation. Relays Were Repaired & Retested
ML17265A6441999-05-13013 May 1999
LER 99-005-00:on 990413,undervoltage Signal of Safeguards Bus During Testing Resulted in Automatic Start of B Edg. Caused by Personnel Error.Blown Fuse Was Replaced & Offsite Power Was Restored to Safeguards Bus 17.With 990513 Ltr
ML17265A6431999-05-12012 May 1999
LER 99-004-00:on 990412,discovered That Containment Recirculation Fan Moisture Separator Vanes Were Incorrectly Installed,Per 10CFR21.Caused by Improper Assembly by Mfg. Subject Vanes Were Dismantled & Correctly re-installed
ML17265A6381999-05-0707 May 1999
Part 21 Rept Re Replacement Turbocharger Exhaust Turbine Side Drain Port Not Functioning as Design Intended.Caused by Manufacturing Deficiency.Turbocharger Was Reaasembled & Reinstalled on B EDG
ML17265A6391999-04-30030 April 1999
Monthly Operating Rept for Apr 1999 for Re Ginna Nuclear Power Plant.With 990510 Ltr
ML17265A6361999-04-23023 April 1999
Part 21 Rept Re Power Supply That Did Not Work Properly When Drawn from Stock & Installed in -25 Vdc Slot.Power Supply Will Be Sent to Vendor to Perform Failure Mode Assessment.Evaluation Will Be Completed by 991001
ML17265A6301999-04-18018 April 1999
Rev 1 to Cycle 28 COLR for Re Ginna Npp.
ML17265A6251999-04-15015 April 1999
Special Rept:On 990309,halon Systems Were Removed from Svc & Fire Door F502 Was Blocked Open.Caused by Mods Being Made to CR Emergency Air Treatment Sys.Continuous Fire Watch Was Established with Backup Fire Suppression Equipment
ML17265A6551999-04-0909 April 1999
Initial Part 21 Rept Re Mfg Deficiency in Replacement Turbocharger for B EDG Supplied by Coltec Industries. Deficiency Consisted of Missing Drain Port in Intermediate Casing.Required Oil Drain Port Machined Open
ML17265A6291999-03-31031 March 1999
Rev 0 to Cycle 28 COLR for Re Ginna Npp.
ML17265A6241999-03-31031 March 1999
Monthly Operating Rept for Mar 1999 for Ginna Station.With 990409 Ltr
ML17265A6141999-03-31031 March 1999
LER 99-003-00:on 990301,two Main Steam non-return Check Valves Were Declared Inoperable Due to Exceedance of Acceptance Criteria.Caused by Changes in Methodology & Matls.Packing Gland Torque Will Be Adjusted.With 990331 Ltr
ML17265A6131999-03-29029 March 1999
LER 99-002-00:on 990227,discovered That Surveillance Had Not Been Performed at Frequency,Per Ts.Caused by Personnel Error.Procedure O-6.13 Will Be Evaluated for Enhancement Documentation of Completion of ITS Srs.With 990329 Ltr
ML17265A6061999-03-24024 March 1999
LER 99-001-00:on 990222,plant Was Noted Outside Design Basis.Caused by Deficiencies in NSSS Vendor Slb Mass & Energy Release.Placed Temporary Administrative Restriction 40 Degrees F Max on Screenhouse Bay Temp
ML17265A5661999-03-0101 March 1999
Rev 26 to QA Program for Station Operation.
ML17265A5961999-02-28028 February 1999
Monthly Operating Rept for Feb 1999 for Ginna Nuclear Power Plant.With 990310 Ltr
ML17265A5371999-01-31031 January 1999
Monthly Operating Rept for Jan 1999 for Re Ginna Nuclear Power Plant.With 990205 Ltr
ML17265A5951998-12-31031 December 1998
Rg&E 1998 Annual Rept.
ML17265A5001998-12-21021 December 1998
Rev 26 to QA Program for Station Operation.
ML17265A4951998-12-21021 December 1998
LER 98-005-00:on 981120,loss of 34.5 Kv Offsite Power Circuit 751,resulted in Automatic Start of B Edg.Caused by Faulted Cable Splice.Performed Appropriate Actions of Abnormal Procedure AP-ELEC.1.With 981221 Ltr
ML17265A4931998-12-17017 December 1998
LER 98-004-00:on 971030,determined That Improperly Performed Surveillance Resulted in Condition Prohibited by Ts.Caused by Procedure non-adherence.Appropriate Calibr Procedures Were Properly Performed with 24 H of Condition Discovery
ML17265A4761998-11-30030 November 1998
Monthly Operating Rept for Nov 1998 for Re Ginna Nuclear Power Plant.With 981210 Ltr
ML17265A4691998-11-25025 November 1998
LER 98-003-01:on 980904,actuations of CR Emergency Air Treatment Systems (Creats) Occurred.Caused by Radon build-up During Temp Inversion.Creats Actuation Signal Was Reset & Normal Ventilation Was Restored to CR
ML17265A4531998-10-31031 October 1998
Monthly Operating Rept for Oct 1998 for Re Ginna Nuclear Power Plant.With 981110 Ltr
ML17265A4271998-10-0505 October 1998
LER 98-003-00:on 980904,actuations of CR Emergency Air Treatment Sys Occurred.Caused by Radon build-up During Temp Inversion.Air Samples Were Taken & Determined That Source of Radiation Was Naturally Occurring Radon.With 981005 Ltr
ML17265A4291998-09-30030 September 1998
Monthly Operating Rept for Sept 1998 for Re Ginna Nuclear Power Plant.With 981009 Ltr
1999-09-30
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Text
DEMONSTRATION OFTHECONFORMANCE OFEXXONNUCLEARCOMPANYFUELTOTHEWESTINGHOUSE K(Z)OPERATING ENVELOPEFORTHEROBERTE.GINNANUCLEARPOWERPLANTWestinghouse ElectricCorporation NuclearTechnology DivisionNuclearSafetyDepartment Safeguards Engineering andDevelopment September 19858512200249 851216PDRADOCK05000244',P'DR I.ntroduction Thisdocumentreportstheresultsofasensitivity studythatwasperformed inordertodemonstrate conformance ofExxonNuclearCompanynuclearfuelintheRobertE.GinnanuclearpowerplanttotheWestinghouse K(z)operating envelope.
Inparticular, theresultsof.thisanalysisshowthatforskewedtothetoppowershapes,inadditiontothepowershapepeakedatthemid-coreelevation, thattheworstpeakcladdingtemperature (PCT)intheunlikelyeventofaLoss-Of-Coolant-Accident (LOCA)remainsbelowthe2200deg-Flimitasspecified byAppendixKof10CFR50.46.
II.MethodofAnalsisThesensitivity studywasperformed usingtheLOCTAcomputercodeoftheWestinghouse 1981LargeBreakLOCAEvaluation Model(WEM)tocalculate thePCTforExxonfuelforthreepowershapes.Thepower.shapesinvestigated Iwerepeakedat6.0ft.,8.0ft.,andat10.5ft.ThepowershapesusedintheLOCAanalysesareshowninFigures1-3.ThepeakpowerofeachpowershapeislimitedbythecurrentK(z)envelopefortheRobertE.Ginna(RGE)powerplant.ThecurrentK(z)envelopeforRGEassumesamaximumtotalpeakingfactorof2.32,andahotchannelenthalpyrisefactorof1.66.
Thefueldesignparameters fortheExxonfuelwereobtainedfromtheExxonNuclearCompanythroughathree-party proprietary agreement betweenWestinghouse, Rochester Gas6Electric, an'dExxon.Thefuelparameters specifictoeachpowershapeweregenerated byExxonandtransmitted toWestinghouse.
Thefuelparameters, whichincludedfuelpellettemperatures andgappressures, werethenusedasinputineachoftheLOCTAcalculations.
TheresultsoftheLOCTAcalculations aresummarized inthefollowing table:'IComarisonofExxonFuelPeakCladdinTemeraturesPowerShapePeakPCTOFPCTElevation PCTTimesec.6~08'10'1781159815287.257'510.001065..14'Theseresultsdemonstrate thatfortheRobertE.GinnaUnit,thatthechoppedcosinepowershape(i.e.6.0ft.peakedshape)generates themostlimitingpeakcladtemperature.
Figures4-6showthecladtemperature responseforthepeaknodeforthe6.0,8.0,and10.5ft.powershapesrespectively.
Animportant observation oftheseresultsisthatforthetop-skewed shapes,thepeakcladdingtemperature occursduringtheblowdownphase.Thisisimportant becausemostoftheAppendixKprescribed analytical modelshavetheirgreatestinfluence duringtherefloodphase.Apeak
'~cladtemperature whichoccursduringrefloodissensitive tocore-wide andsystem-wide hydraulic phenomena, whileablowdownpeakisastrongerfunctionofinitialfuelstoredenergy.Acomparison ofthepeakcladtemperatures duringtheblowdownandrefloodphasesforeachofthesepowershapesprovidesamoreconclusive demonstration thatthechoppedcosinepowershapeproducesthemostlimitingLOCAresults.ComarisonofPeakCladdinTemeraturesDurinBlowdownPowerShapePeakPCT-Blowdown opPCTElevation Timesec.6.08'10'1635159815286~007'510.005'5'4'Thecomparison ofpeakcladtemperatures duringblowdownshowsthatthehighestPCToccursforthechoppedcosinepowershape.Thisisreasonable.,
because6.0ft.shapepermitsthehighesttotallocalpeakingfactor(2.32).ThefactthatthePCTforthetop-skewed shapesoccursbelowthepeakpowerlocationisduetothebetterheattransferathigherelevations thatoccursduringtheperiodofnegativecoreflow.
ComarisonofPeakCladdinTemeraturesDurinRefloodPowerShapePeakPCT-Reflood OFPCTElevation Timesec.6~08~010'1781158514587'5F0010.001067468Thecomparison ofrefloodpeakcladtemperatures showsanevenwidermarginbetweenthechoppedcosineshapeandthetop-skewed powershapes.Znadditiontoshowingthatthechoppedcosinepowershape,isthe"worst"powershapeforaLOCAanalysisofRGEwithExxonfuel,italsodemonstrates alargemargintothe2200deg-Flimitforthetop-skewed shapesforthisplant.III.UseofNon-Exxon FuelHdraulicsThissensitivity studywasperformed byre-calculating thecladVItemperature responseforExxonforthethreepowershapesusingtheLOCTAcomputercode.Thehotassemblyhydraulics wasnotre-calculated fortheExxonfuel.Theblowdownandrefloodhydraulic transients aregenerated usingtheSATAN,WREFLOODandCOCOcomputercodes.Existingblowdownandrefloodhydraulics fromapreviousRGEplantspecificanalysiswithW-OFAfuelwereusedashotassemblyboundaryconditions fortheLOCTAcalculations.
UseoftheOFAhydraulics arejustifiable forthissensitivity studyonthefollowing basis:
(l)TheW-OFArodsizeissmallerthantheExxonrodsize.ThiswillresultinaslowerrefloodofafullcoreofW-OFAfuel.Thus,foragivenpowershape,theuseoftheOFAcorerefloodrateprovidesaconservative'eflood ratefortheExxonfuelheat-upcalculation.
(2)Previoussensitivities withtop-skewed powershapesinSATANhavehotshownalargeeffectontheend-of-blowdown fueltemperatures.
Thus,theuseofSATANchoppedcosinepowershaperesultsfortop-skewed LOCTAcalculations canbeconsidered sufficiently accurateforaroughsensitivity study.Becauseofthewidemarginbetweenthecosineandthetop-skewed shapePCTsinthisanalysis, re-calculation oftheSATANtransient isnotexpectedtochangetheconclusions ofthisanalysis.
IV.Conclusions TheWestinghouse LargeBreakLOCAEvaluation Modelcladheat-upcomputercode,LOCTA,wasusedtoanalyzeExxonfuelforthreepowershapes.Theresultsconfirmed thatthepowershapepeaked..at thecenterofthecoreproducesthehighestpeakcladdingtemperature.
ThisresultfortheExxonfuelisconsistent withpowershapestudiesperformed byWestinghouse withthesamecomputercodesforWestinghouse fuel.Theresultsofthisstudydemonstrate thattheExxonfuelintheRobertE.Ginnanuclearpowerplantconformstothecurrentoperating K(z)envelopefortop-skewed powershapes.Whiletheentiretransient wasnotrecalculated forthisanalysis, acompletere-analysis wouldnotbeexpectedtochangetheconclusions ofthissensitivity study.\
0.03.04.25,.4.75 5.255.75-6.25 6.757.257.759.012.1.S4.04.55.05.56.06.57.07.58.010.5COREHEIGHT(FEET)Figure1.AxialPowerShapePeakedat6.0ft.(ChoppedCosinePowerShape) 0.01.02.03.04.05.06.07.08.09.010.011.012.0
.51.52.53.54.55.56.57.58.59.510.511.5COREHEIGHT(FEET)Figure2.AxialPowerShapePeakedat8.0ft; 2.S2.0M~1.SC51.0.5.0.01.02.03.04.05.06.07.08.09.010.011.012.0
.51.52.53.54.55.56.57.58.59.510.511.5COREHEIGHT(FEET)Figure3.AxialPowerShapePeakedat10.5ft.
c~tlctl~clvbttt~ovtv[5tovlt5vttl51u0tO.~Cl(IC.C0tlte~OvltSvttlll~0~vC1[vt<<01~00IutlfiC00t'llItlJt~tot5tll~ItW>.OCC5'AO.Ovoltd.O50vAOS<lvl~55C>Figure4.CladTemperature ResponseforPCTLocationforthe6.0ft.PowerShape.
~Oll~lC~CllilCt~00lulltOOC~SwltlS'IUOt0'OCClCO,O~CaatOol~SsstlCCAOArt.lllit.wOS 000~VOS'i~tCSlitl~CitySy$0tll+lVlC000.0C\SSOO.ONXol000.00.088SINSISCCIFigure5.-CladTemperature ResponseforPCTLocationfor8.0ft.PowerShape
~OltV<C,CIVVVt<<OV<Vt<VOVtV5Vatt5<VDV0,~Ott<C10.5VtvvVOvtV5vVVtC<JDAVC.1t<<V,<<0<
VODOvv51~0.00<1<<Vtvv~O,DOA<~<W~0000.0o1500.0XE<Co1000.0lr0.0CII<t<5CCISg8Figure6.CladTemperature ResponseforPCTLocationforthe10.5ft.PowerShape-11
