ML15314A658

Wolf Creek - Cycle 21 Core Operating Limits Report, Revision 1

ML15314A658
Person / Time
Site:	Wolf Creek
Issue date:	11/04/2015
From:	Hafenstine C R Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 15-0081
Download: ML15314A658 (17)

Text

LFCREEKNIUCLEAR OPERATING CORPORATIONNovember 4, 2015Cynthia R. HafenstineManager Regulatory AffairsRA 15-0081U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 21 CoreOperating Limits Report, Revision 1Gentlemen:Enclosed is Revision 1 of the Wolf Creek Generating Station Cycle 21 Core Operating LimitsReport (COLR). Revision 1 incorporates changes associated with the implementation ofAmendment No. 213. This document is being submitted pursuant to Section 5.6.5 of the WolfCreek Generating Station Technical Specifications.This letter contains no commitments. If you have any questions concerning this matter, pleasecontact me at (620) 364-4204.Sincerely,Cynthia R. HafenstineCRH/rltEnclosurecc: M. L. Dapas (NRC), wleC. F. Lyon (NRC), w/eN. H. Taylor (NRC), w/eSenior Resident Inspector (NRC), w/eP.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831An Equal Opportunity Employer M/F/HCINET W@LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 1WOLF CREEK GENERATING STATIONCYCLE 21CORE OPERATING LIMITS REPORTRevision 1September 2015Prepared by:Reviewed by:Approved by:Jeff Blair DateKeith Colussy DateGregory S. KinnDatePage 1 of 16 C R ICI¢K ~Wolf Creek Generating StatiOncce2W NULFA CPR EEK CRORTO Core Operating Limits Report'NUCEAROPERTIN CORORAIONRevision I1.0 CORE OPERATING LIMITS REPORTThe CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating StationCycle 21 has been prepared in accordance with the requirements of TechnicalSpecification 5.6.5.The core operating limits that are included in the COLR affect the following TechnicalSpecifications:2.1.1 Reactor Core Safety Limits3.1 .1 Shutdown Margin (SDM)3.1.3 Moderator Temperature Coefficient (MTC)3.1 .4 Rod Group Alignment Limits3.1.5 Shutdown Bank Insertion Limits3.1.6 Control Bank Insertion Limits3.1.8 PHYSICS TESTS Exceptions -MODE 23.2.1 Heat Flux Hot Channel Factor (FQ(z)) (Fo Methodology)3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (Fr)3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control(RAOC) Methodology)3.3.1 Reactor Trip System (RTS) Instrumentation3.4.1 RCS Pressure, Temperature, and Flow Departure from NucleateBoiling (DNB) Limits3.9.1 Boron ConcentrationThe portions of the Technical Specification Bases affected by the report are listedbelow:ASA B 3.4.1 RCS Pressure, Temperature, and Flow Departure from NucleateBoiling (DNB) LimitsPage 2 of 16 W0LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.0 OPERATING LIMITSThe cycle-specific parameter limits for the specifications listed in Section 1.0 arepresented in the subsections below:2.1 Reactor Core Safety Limits (SL 2.1.1)In MODES 1 and 2, the combination of THERMAL POWER, Reactor CoolantSystem (RCS) highest loop average temperature, and pressurizer pressure shallnot exceed the limits in Figure 2.1.680660640LI-6206005805600.00.2 0.4 0.6 0.8 1.0Fraction of Rated Thermal Power1.2Figure 2.1Reactor Core Safety LimitsPage 3 of 16

'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.2 Moderator Temperature Coefficient (MTC') (LCO 3.1.3, SR 3.1.3.2)The MTC shall be less positive than the limit provided in Figure 2.2.The MTC shall be less negative than -50 pcm/°F.The 300 PPM MTC Surveillance limit is -41 pcm/°F (equilibrium, all rodswithdrawn, RATED THERMAL POWER condition).The 60 PPM MTC Surveillance limit is -46 pcm/°F (equilibrium, all rodswithdrawn, RATED THERMAL POWER condition).UI-zw0_a.0UNACCEPTABLEOPERATION6.0, 70%ACCEPTABLEOPERATION0 10 20 30 40 50 60% of RATED THERMAL POWER70 80 90 100Figure 2.2Moderator Temperature Coefficient Vs.THERMAL POWER (%)Page 4 of 16 W@LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.3 Shutdown Bank Insertion Limits (LCO 3.1.5)The shutdown banks shall be fully withdrawn (i.e., positioned within the intervalof > 222 and < 231 steps withdrawn).2.4 Control Bank Insertion Limits (LCO 3.1 .6)The Control Bank insertion, sequence, and overlap limits are specified in Figure2.4.(FULLY WITHDRAWN)220 r--- I-- F1- --__ -Z--200180160STE 140PSW 120TH 100DRA 80WN60402000(FULLY INSERTED)20 40 60 80THERMAL POWER (Percent)100Figure 2.4Control Bank Insertion, Sequence, and Overlap Limits Vs.THERMAL POWER (%) -Four Loop OperationFully withdrawn shall be the condition where control banks are at a position within theinterval of> 222 and < 231 steps withdrawn.Page 5 of 16

'WSLF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.5 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)Methodoloaqy) (LCO 3.2.3)The indicated AXIAL FLUX DIFFERENCE (AFD) allowed operational space isdefined by Figure 2.5.110F90RATED 80THER70MALP 60ER5040UNACCEPTABLEOPERATION( -29 ,50 )ACCEPTABLEOPERATIONUNACCEPTABLEOPERATION( 24,50 )! I I I I I I I ! I I I I I-40-30-20 -10AXIAL FLUX0 10 20DIFFERENCE (% AI)30 40Figure 2.5AXIAL FLUX DIFFERENCE Limits as aFunction of THERMAL POWER (%)Page 6 of 16 W LF C~lli=KWolf Creek Generating StatiOnyce2WNUCLFA CPR EEK CRORTO Core Operating Limits Report'NUCEAROPERTIN CORORAIONRevision 12.6 Heat Flux Hot Channel Factor (FoIZY)(F0 Methodologqy) (LCO 3.2.1, SR 3.2.1 .2)for P>O0.5FQ(Z)<_C.Q5*K(Z), for P _ 0.5where ~ -THERMAL POWERRATED THERMAL POWERCFQ = FlrTPF~P= Fo(Z) limit at RATED THERMAL POWER (RTP)= 2.50, andK(Z) = as defined in Figure 2.6.Fo'(Z) is the measured value of F0(Z), inferred from a power distributionmeasurement obtained with the Movable Incore Detector System (MIDS) or thePower Distribution Monitoring System.Measurement uncertainty is applied as follows.FffZ)= when F oM(Z) is obtained from MIDS.Fg(Z) = when FoM(Z) is obtained from PDMS.Manufacturing tolerances are accounted for in the 1 .03 Engineering uncertaintyfactor. Measurement uncertainty for MIDS is accounted for in the 1.05 factor.PDMS measurement uncertainty is accounted for in the UQU factor, and it isdetermined by PDMS.where, W(Z) = a cycle dependent function that accounts for power distributiontransients encountered during normal operation (see Appendix A).When using the PDMS, F oY(Z) uses that is determined from an that reflects full-power steady-state conditions rather than current conditions.See Appendix A for: FQ Penalty Factor.Page 7 of 16 WCLF CREEKTNUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 11.2ir1.00.C. 0.8S0.z106z 0.2+ +FQRW= = 2.50Bevation (ft) K(Z)-0.0 1.0006.0 1.00012.0 0.9250.0l I024681012CORE HEIGHT (FT)Figure 2.6K(Z) -Normalized Peaking Factor Vs. Core HeightPage 8 of 16 W LF CRI¢ICKWolf Creek Generating StatiOncce2WNULEAR CPR EEK CRORTO Core Operating Limits ReportOPERTIN CORORAIONRevision I2.7 Nuclear Enthalpy Rise Hot Channel Factor (Ft) (LCO 3.2.2)shall be limited by the following relationship:FNJ _< FLfft[1.O + -P]Where, F7p = limit at RATED THERMAL POWER (RTP)= 1.650= power factor multiplier for = 0.3P = THERM/AL POWERRATED THERMAL POWERF[H = is the measured value of inferred from a powerdistribution measurement obtained with the Movable IncoreDetector System (MIDS) or the Power Distribution MonitoringSystem (PDMS). Measurement uncertainty is applied asfollows.When is obtained from MIDS, the measured value ismultiplied by 1 .04.When is obtained from PDMS, the measured value isincreased by an uncertainty factor (UAH), and the factor isdetermined by PDMS, with a lower limit of 4%.Page 9 of 16 W6FCREEKPRTNG CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision I2.8 Reactor Trip~ System Overtemperature AT Setpoint Parameter Values (LCO3.3.1, Table 3.3.1-1, Note 1)ParameterOvertemperature AT reactor trip setpointOvertemperature AT reactor trip setpoint TavgcoefficientOvertemperature AT reactor trip setpoint pressurecoefficientNominal Tavg at RTPNominal RCS operating pressureMeasured RCS AT lead/lag constantMeasured RCS AT lag constantMeasured RCS average temperature lead/lagMeasurdRtaergaeprauela/lg=teValueK1 = 1.10K2 = 0.01 37/°FK3 = 0.000671/psigT' < 586.5°FP' >2235 psig1 = 6 sec1t2 = 3 sec13= 2 sec"14 =16 sec15= 4 secconstantf1(AI) = -0.0227 {23% + (qt-qb)} when (qt-qb) < -23% RTP0% of RTPwhen -23% RTP < (qt-qb,) < 5% RTP0.0184 {(qt-qb) -5%} when (qt-qb,) > 5% RTPWhere, qt and qb are percent RTP in the upper and lower halves of the core,respectively, and qt + qb is the total THERMAL POWER in percent RTP.Page 10 of 16 W0LF CREEKrNUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.9 Reactor Trip System Overpower AT Setpoint Parameter Values (LCO 3.3.1,Table 3.3.1-1, Note 2)ParameterOverpower AT reactor trip setpointOverpower AT reactor trip setpoint Tavgrate/lag coefficientOverpower AT reactor trip setpoint Tavg heatupcoefficientIndicated mavg at RTP (calibration temperaturefor AT instrumentation)Measured RCS AT lead/lag constantMeasured RCS AT lag constantMeasured RCS average temperature lead/lagconstantMeasured RCS average temperature rate/lagconstantValueK4= 1.10I5 =O.02/0F for increasing Tavg= 0/°F for decreasing Tavg0.001 28/°F for T > T"= 0/OF forT<T,,T" _< 586.5°F11= 6 sec1;2 = 3 sec13= 2 sec16= 0 sec;7= 10 secf2(AI) = 0% RTP for all AlPage 11 of 16

'W LF CREEK7NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling ('DNB)Limits (LCO 3.4.1)ParameterPressurizer pressureRCS average temperatureRCS total flow rateIndicated ValuePressure _2220 psigTavg<590.5 °FFlow > 371,000 gpm2.11 Boron Concentration (LCO 3.9.1)The refueling boron concentration shall be greater than or equal to 2300 PPM.2.12 SHUTDOWN MARGIN (LCO 3.1.1, 3.1.4, 3.1.5, 3.1.6, & 3.1.8)The SHUTDOWN MARGIN shall be greater than or equal to 1300 pcm(1.3% Ak/k).2.13 Departure from Nucleate Boilinq Ratio (DNBR) Limits (B 3.4.1, ASA)Safety Analysis DNBR Limit 1.76WRB-2 Design Limit DNBR 1.23Page 12 of 16 W@LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 1APPENDIX AA. Input relating to LCO 3.2.1:W() Q( Z )steadystateW(z) 1EQ (Z)stcadystate 0.5'wher ~ = THERMAL POWERRATED THERMAL POWERfor P > 0.5for P <_ 0.5FQ(Z)maxtransient =Maximum (FQ('Z) xp) calculated over the entire range of power shapesanalyzed for Condition I operations (p = power at which maximumoccurs).FQ(Z)stea'stae = (FQ(Z) xp) calculated at full power (p = 1.0) equilibrium conditions.The W(z) values are generated at full power equilibrium conditions (P = 1.0). W(z)values specific to part-power conditions may also be generated; these can beused for part-power surveillance measurements, rather than the full-power W(z)values. For these part-power W(z) values, the FQ(Z)steady state (denominator inabove equations) is generated at the specific anticipated surveillance conditions.W(Z) values are issued in controlled reports which will be provided on request.Page 13 of 16

• W#LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 1Input relating to SR 3.2.1.2Cycle Burnup(MWD/MTU)> 0 to <150348546743 to 84568654885290499247944596439840100381023610434> 10631Cycle Burnup(MWD/MTU)F0o(Z) Penalty Factor(%)3.082.762.382.002.062.262.442.402.382.362.312.232.152.062.00F o(7) Exclusion Zone(% [INCORE meshpoints])Top I:Bottom10 [7] I: 10 [7]<8 8000> 8,000Page 14 of 16 CREEK ~ ~Wolf Creek yleGenerating 21StationWUCLFA CPR EEK CRORTO Core Operating Limits Report'NUCEAROPERTIN CORORAIONRevision 1B. App~roved Analytical Methods for Determining Core Op~eratinaq LimitsThe analytical methods used to determine the core operating limits shall be thosepreviously reviewed and approved by the NRC, specifically those described in thefollowing documents.1. WCNOC Topical Report TR 90-0025 WO1, "Core Thermal Hydraulic AnalysisMethodology for the Wolf Creek Generating Station." (ET 90-0140, ET 92-0103)NRC Safety Evaluation Report dated October 29, 1992, for the "Core ThermalHydraulic Analysis Methodology for the Wolf Creek Generating Station."2. WCAP-1 1397-P-A, "Revised Thermal Design Procedure," April 1989.NRC Safety Evaluation Report dated January 17, 1989, for the "Acceptancefor Referencing of Licensing Topical Report WCAP-1 1397, Revised ThermalDesign Procedure."3. WCNOC Topical Report NSAG-006, "Transient Analysis Methodology for the WolfCreek Generating Station" (ET-91-0026, ET 92-0142, WM 93-001 0, WM~ 93-0028).NRC Safety Evaluation Report dated September 30, 1993, for the "TransientAnalysis Methodology for the Wolf Creek Generating ,Station."EPRI Topical Report NP-7450(A), "RETRAN-3D -A Program for TransientThermal-Hydraulic Analysis of Complex Fluid Flow Systems," including NRCSafety Evaluation Report dated January 25, 2001, "Safety Evaluation Reporton EPRI Topical Report NP-7450(P), Revision 4, "RETRAN-3D -A Programfor Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,"(TAC No. MA431 1)." RETRAN-3D code is only utilized in the RETRAN-02mode.4. WCAP-1 0216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control -FQSurveillance Technical Specification," February 1994.NRC Safety Evaluation Report dated November 26, 1993, "Acceptance forReferencing of Revised Version of Licensing Topical Report WCAP-1 021 6-P,Rev. 1, Relaxation of Constant Axial Offset Control -F0 SurveillanceTechnical ,Specification" (TAC No. M88206).5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for theWolf Creek Generating Station" (ET 92-0032, ET 93-0017).NRC ,Safety Evaluation Report dated March 10, 1993, for the "Reload SafetyEvaluation Methodology for the Wolf Creek Generating Station."6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to TechnicalSpecification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054).Page 15 of 16

..... C R I KWolf Creek Generating Station'NUCLEAR OPERATING CORPORATION RoeO evisiong LimisRpr7. WCAP-1 6009-P-A, "Realistic Large Break LOCA Evaluation MethodologyUsing Automated Statistical Treatment of Uncertainty Method (ASTRUM),"Revision 0, January 2005.NRC letter dated November 5, 2004,"Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation MethodologyUsing Automated Statistical Treatment of Uncertainty Method (ASTRUM)"(TAC NO. MB9483)."8. WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport CodePARAGON," August 2004.NRC Safety Evaluation dated March 18, 2004, "Final Safety Evaluation forWestinghouse Topical Report WCAP-1 6045-P, Revision 0, "Qualification ofthe Two-Dimensional Transport Code PARAGON."9. WCAP-1 6045-P-A, Addendum I-A, "Qualification of the NEXUS Nuclear DataMethodology," August 2007.NRC Safety Evaluation dated February 23, 2007, "Final Safety Evaluation forWestinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16045-P-A, Addendum 1, "Qualification of the NEXUS Nuclear DataMethodology" (TAC NO. MC9606)."10. WCAP 1 0965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"September 1986.NRC letter dated June 23, 1986, "Acceptance for Referencing of TopicalReport WCAP 10965-P and WCAP 10966-NP."11. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.NRC Safety Evaluation Reports dated July 1, 1991, "Acceptance forReferencing of Topical Report WCAP-1 2610, 'VANTAGE+ Fuel AssemblyReference Core Report' (TAC NO. 77258)."NRC Safety Evaluation Report dated September 15, 1994, "Acceptance forReferencing of Topical Report WCAP-12610, Appendix B, Addendum 1,'Extended Burnup Fuel Design Methodology and ZIRLO Fuel PerformanceModels' (TAC NO. M8641 6)."12. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and ThermalOvertemperature AT Trip Function." September 1986.NRC Safety Evaluation Report dated April 17, 1986, "Acceptance forReferencing of Licensing Topical Report WCAP-8745(P)/8746(NP), 'DesignBases for the Thermal Overpower AT and Thermal Overtemperature AT TripFunctions."'"Page 16 of 16 LFCREEKNIUCLEAR OPERATING CORPORATIONNovember 4, 2015Cynthia R. HafenstineManager Regulatory AffairsRA 15-0081U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Cycle 21 CoreOperating Limits Report, Revision 1Gentlemen:Enclosed is Revision 1 of the Wolf Creek Generating Station Cycle 21 Core Operating LimitsReport (COLR). Revision 1 incorporates changes associated with the implementation ofAmendment No. 213. This document is being submitted pursuant to Section 5.6.5 of the WolfCreek Generating Station Technical Specifications.This letter contains no commitments. If you have any questions concerning this matter, pleasecontact me at (620) 364-4204.Sincerely,Cynthia R. HafenstineCRH/rltEnclosurecc: M. L. Dapas (NRC), wleC. F. Lyon (NRC), w/eN. H. Taylor (NRC), w/eSenior Resident Inspector (NRC), w/eP.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831An Equal Opportunity Employer M/F/HCINET W@LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 1WOLF CREEK GENERATING STATIONCYCLE 21CORE OPERATING LIMITS REPORTRevision 1September 2015Prepared by:Reviewed by:Approved by:Jeff Blair DateKeith Colussy DateGregory S. KinnDatePage 1 of 16 C R ICI¢K ~Wolf Creek Generating StatiOncce2W NULFA CPR EEK CRORTO Core Operating Limits Report'NUCEAROPERTIN CORORAIONRevision I1.0 CORE OPERATING LIMITS REPORTThe CORE OPERATING LIMITS REPORT (COLR) for Wolf Creek Generating StationCycle 21 has been prepared in accordance with the requirements of TechnicalSpecification 5.6.5.The core operating limits that are included in the COLR affect the following TechnicalSpecifications:2.1.1 Reactor Core Safety Limits3.1 .1 Shutdown Margin (SDM)3.1.3 Moderator Temperature Coefficient (MTC)3.1 .4 Rod Group Alignment Limits3.1.5 Shutdown Bank Insertion Limits3.1.6 Control Bank Insertion Limits3.1.8 PHYSICS TESTS Exceptions -MODE 23.2.1 Heat Flux Hot Channel Factor (FQ(z)) (Fo Methodology)3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (Fr)3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control(RAOC) Methodology)3.3.1 Reactor Trip System (RTS) Instrumentation3.4.1 RCS Pressure, Temperature, and Flow Departure from NucleateBoiling (DNB) Limits3.9.1 Boron ConcentrationThe portions of the Technical Specification Bases affected by the report are listedbelow:ASA B 3.4.1 RCS Pressure, Temperature, and Flow Departure from NucleateBoiling (DNB) LimitsPage 2 of 16 W0LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.0 OPERATING LIMITSThe cycle-specific parameter limits for the specifications listed in Section 1.0 arepresented in the subsections below:2.1 Reactor Core Safety Limits (SL 2.1.1)In MODES 1 and 2, the combination of THERMAL POWER, Reactor CoolantSystem (RCS) highest loop average temperature, and pressurizer pressure shallnot exceed the limits in Figure 2.1.680660640LI-6206005805600.00.2 0.4 0.6 0.8 1.0Fraction of Rated Thermal Power1.2Figure 2.1Reactor Core Safety LimitsPage 3 of 16

'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.2 Moderator Temperature Coefficient (MTC') (LCO 3.1.3, SR 3.1.3.2)The MTC shall be less positive than the limit provided in Figure 2.2.The MTC shall be less negative than -50 pcm/°F.The 300 PPM MTC Surveillance limit is -41 pcm/°F (equilibrium, all rodswithdrawn, RATED THERMAL POWER condition).The 60 PPM MTC Surveillance limit is -46 pcm/°F (equilibrium, all rodswithdrawn, RATED THERMAL POWER condition).UI-zw0_a.0UNACCEPTABLEOPERATION6.0, 70%ACCEPTABLEOPERATION0 10 20 30 40 50 60% of RATED THERMAL POWER70 80 90 100Figure 2.2Moderator Temperature Coefficient Vs.THERMAL POWER (%)Page 4 of 16 W@LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.3 Shutdown Bank Insertion Limits (LCO 3.1.5)The shutdown banks shall be fully withdrawn (i.e., positioned within the intervalof > 222 and < 231 steps withdrawn).2.4 Control Bank Insertion Limits (LCO 3.1 .6)The Control Bank insertion, sequence, and overlap limits are specified in Figure2.4.(FULLY WITHDRAWN)220 r--- I-- F1- --__ -Z--200180160STE 140PSW 120TH 100DRA 80WN60402000(FULLY INSERTED)20 40 60 80THERMAL POWER (Percent)100Figure 2.4Control Bank Insertion, Sequence, and Overlap Limits Vs.THERMAL POWER (%) -Four Loop OperationFully withdrawn shall be the condition where control banks are at a position within theinterval of> 222 and < 231 steps withdrawn.Page 5 of 16

'WSLF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.5 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC)Methodoloaqy) (LCO 3.2.3)The indicated AXIAL FLUX DIFFERENCE (AFD) allowed operational space isdefined by Figure 2.5.110F90RATED 80THER70MALP 60ER5040UNACCEPTABLEOPERATION( -29 ,50 )ACCEPTABLEOPERATIONUNACCEPTABLEOPERATION( 24,50 )! I I I I I I I ! I I I I I-40-30-20 -10AXIAL FLUX0 10 20DIFFERENCE (% AI)30 40Figure 2.5AXIAL FLUX DIFFERENCE Limits as aFunction of THERMAL POWER (%)Page 6 of 16 W LF C~lli=KWolf Creek Generating StatiOnyce2WNUCLFA CPR EEK CRORTO Core Operating Limits Report'NUCEAROPERTIN CORORAIONRevision 12.6 Heat Flux Hot Channel Factor (FoIZY)(F0 Methodologqy) (LCO 3.2.1, SR 3.2.1 .2)for P>O0.5FQ(Z)<_C.Q5*K(Z), for P _ 0.5where ~ -THERMAL POWERRATED THERMAL POWERCFQ = FlrTPF~P= Fo(Z) limit at RATED THERMAL POWER (RTP)= 2.50, andK(Z) = as defined in Figure 2.6.Fo'(Z) is the measured value of F0(Z), inferred from a power distributionmeasurement obtained with the Movable Incore Detector System (MIDS) or thePower Distribution Monitoring System.Measurement uncertainty is applied as follows.FffZ)= when F oM(Z) is obtained from MIDS.Fg(Z) = when FoM(Z) is obtained from PDMS.Manufacturing tolerances are accounted for in the 1 .03 Engineering uncertaintyfactor. Measurement uncertainty for MIDS is accounted for in the 1.05 factor.PDMS measurement uncertainty is accounted for in the UQU factor, and it isdetermined by PDMS.where, W(Z) = a cycle dependent function that accounts for power distributiontransients encountered during normal operation (see Appendix A).When using the PDMS, F oY(Z) uses that is determined from an that reflects full-power steady-state conditions rather than current conditions.See Appendix A for: FQ Penalty Factor.Page 7 of 16 WCLF CREEKTNUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 11.2ir1.00.C. 0.8S0.z106z 0.2+ +FQRW= = 2.50Bevation (ft) K(Z)-0.0 1.0006.0 1.00012.0 0.9250.0l I024681012CORE HEIGHT (FT)Figure 2.6K(Z) -Normalized Peaking Factor Vs. Core HeightPage 8 of 16 W LF CRI¢ICKWolf Creek Generating StatiOncce2WNULEAR CPR EEK CRORTO Core Operating Limits ReportOPERTIN CORORAIONRevision I2.7 Nuclear Enthalpy Rise Hot Channel Factor (Ft) (LCO 3.2.2)shall be limited by the following relationship:FNJ _< FLfft[1.O + -P]Where, F7p = limit at RATED THERMAL POWER (RTP)= 1.650= power factor multiplier for = 0.3P = THERM/AL POWERRATED THERMAL POWERF[H = is the measured value of inferred from a powerdistribution measurement obtained with the Movable IncoreDetector System (MIDS) or the Power Distribution MonitoringSystem (PDMS). Measurement uncertainty is applied asfollows.When is obtained from MIDS, the measured value ismultiplied by 1 .04.When is obtained from PDMS, the measured value isincreased by an uncertainty factor (UAH), and the factor isdetermined by PDMS, with a lower limit of 4%.Page 9 of 16 W6FCREEKPRTNG CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision I2.8 Reactor Trip~ System Overtemperature AT Setpoint Parameter Values (LCO3.3.1, Table 3.3.1-1, Note 1)ParameterOvertemperature AT reactor trip setpointOvertemperature AT reactor trip setpoint TavgcoefficientOvertemperature AT reactor trip setpoint pressurecoefficientNominal Tavg at RTPNominal RCS operating pressureMeasured RCS AT lead/lag constantMeasured RCS AT lag constantMeasured RCS average temperature lead/lagMeasurdRtaergaeprauela/lg=teValueK1 = 1.10K2 = 0.01 37/°FK3 = 0.000671/psigT' < 586.5°FP' >2235 psig1 = 6 sec1t2 = 3 sec13= 2 sec"14 =16 sec15= 4 secconstantf1(AI) = -0.0227 {23% + (qt-qb)} when (qt-qb) < -23% RTP0% of RTPwhen -23% RTP < (qt-qb,) < 5% RTP0.0184 {(qt-qb) -5%} when (qt-qb,) > 5% RTPWhere, qt and qb are percent RTP in the upper and lower halves of the core,respectively, and qt + qb is the total THERMAL POWER in percent RTP.Page 10 of 16 W0LF CREEKrNUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.9 Reactor Trip System Overpower AT Setpoint Parameter Values (LCO 3.3.1,Table 3.3.1-1, Note 2)ParameterOverpower AT reactor trip setpointOverpower AT reactor trip setpoint Tavgrate/lag coefficientOverpower AT reactor trip setpoint Tavg heatupcoefficientIndicated mavg at RTP (calibration temperaturefor AT instrumentation)Measured RCS AT lead/lag constantMeasured RCS AT lag constantMeasured RCS average temperature lead/lagconstantMeasured RCS average temperature rate/lagconstantValueK4= 1.10I5 =O.02/0F for increasing Tavg= 0/°F for decreasing Tavg0.001 28/°F for T > T"= 0/OF forT<T,,T" _< 586.5°F11= 6 sec1;2 = 3 sec13= 2 sec16= 0 sec;7= 10 secf2(AI) = 0% RTP for all AlPage 11 of 16

'W LF CREEK7NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 12.10 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling ('DNB)Limits (LCO 3.4.1)ParameterPressurizer pressureRCS average temperatureRCS total flow rateIndicated ValuePressure _2220 psigTavg<590.5 °FFlow > 371,000 gpm2.11 Boron Concentration (LCO 3.9.1)The refueling boron concentration shall be greater than or equal to 2300 PPM.2.12 SHUTDOWN MARGIN (LCO 3.1.1, 3.1.4, 3.1.5, 3.1.6, & 3.1.8)The SHUTDOWN MARGIN shall be greater than or equal to 1300 pcm(1.3% Ak/k).2.13 Departure from Nucleate Boilinq Ratio (DNBR) Limits (B 3.4.1, ASA)Safety Analysis DNBR Limit 1.76WRB-2 Design Limit DNBR 1.23Page 12 of 16 W@LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 1APPENDIX AA. Input relating to LCO 3.2.1:W() Q( Z )steadystateW(z) 1EQ (Z)stcadystate 0.5'wher ~ = THERMAL POWERRATED THERMAL POWERfor P > 0.5for P <_ 0.5FQ(Z)maxtransient =Maximum (FQ('Z) xp) calculated over the entire range of power shapesanalyzed for Condition I operations (p = power at which maximumoccurs).FQ(Z)stea'stae = (FQ(Z) xp) calculated at full power (p = 1.0) equilibrium conditions.The W(z) values are generated at full power equilibrium conditions (P = 1.0). W(z)values specific to part-power conditions may also be generated; these can beused for part-power surveillance measurements, rather than the full-power W(z)values. For these part-power W(z) values, the FQ(Z)steady state (denominator inabove equations) is generated at the specific anticipated surveillance conditions.W(Z) values are issued in controlled reports which will be provided on request.Page 13 of 16

• W#LF CREEK'NUCLEAR OPERATING CORPORATIONWolf Creek Generating StationCycle 21Core Operating Limits ReportRevision 1Input relating to SR 3.2.1.2Cycle Burnup(MWD/MTU)> 0 to <150348546743 to 84568654885290499247944596439840100381023610434> 10631Cycle Burnup(MWD/MTU)F0o(Z) Penalty Factor(%)3.082.762.382.002.062.262.442.402.382.362.312.232.152.062.00F o(7) Exclusion Zone(% [INCORE meshpoints])Top I:Bottom10 [7] I: 10 [7]<8 8000> 8,000Page 14 of 16 CREEK ~ ~Wolf Creek yleGenerating 21StationWUCLFA CPR EEK CRORTO Core Operating Limits Report'NUCEAROPERTIN CORORAIONRevision 1B. App~roved Analytical Methods for Determining Core Op~eratinaq LimitsThe analytical methods used to determine the core operating limits shall be thosepreviously reviewed and approved by the NRC, specifically those described in thefollowing documents.1. WCNOC Topical Report TR 90-0025 WO1, "Core Thermal Hydraulic AnalysisMethodology for the Wolf Creek Generating Station." (ET 90-0140, ET 92-0103)NRC Safety Evaluation Report dated October 29, 1992, for the "Core ThermalHydraulic Analysis Methodology for the Wolf Creek Generating Station."2. WCAP-1 1397-P-A, "Revised Thermal Design Procedure," April 1989.NRC Safety Evaluation Report dated January 17, 1989, for the "Acceptancefor Referencing of Licensing Topical Report WCAP-1 1397, Revised ThermalDesign Procedure."3. WCNOC Topical Report NSAG-006, "Transient Analysis Methodology for the WolfCreek Generating Station" (ET-91-0026, ET 92-0142, WM 93-001 0, WM~ 93-0028).NRC Safety Evaluation Report dated September 30, 1993, for the "TransientAnalysis Methodology for the Wolf Creek Generating ,Station."EPRI Topical Report NP-7450(A), "RETRAN-3D -A Program for TransientThermal-Hydraulic Analysis of Complex Fluid Flow Systems," including NRCSafety Evaluation Report dated January 25, 2001, "Safety Evaluation Reporton EPRI Topical Report NP-7450(P), Revision 4, "RETRAN-3D -A Programfor Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems,"(TAC No. MA431 1)." RETRAN-3D code is only utilized in the RETRAN-02mode.4. WCAP-1 0216-P-A, Revision 1A, "Relaxation of Constant Axial Offset Control -FQSurveillance Technical Specification," February 1994.NRC Safety Evaluation Report dated November 26, 1993, "Acceptance forReferencing of Revised Version of Licensing Topical Report WCAP-1 021 6-P,Rev. 1, Relaxation of Constant Axial Offset Control -F0 SurveillanceTechnical ,Specification" (TAC No. M88206).5. WCNOC Topical Report NSAG-007, "Reload Safety Evaluation Methodology for theWolf Creek Generating Station" (ET 92-0032, ET 93-0017).NRC ,Safety Evaluation Report dated March 10, 1993, for the "Reload SafetyEvaluation Methodology for the Wolf Creek Generating Station."6. NRC Safety Evaluation Report dated March 30, 1993, for the "Revision to TechnicalSpecification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054).Page 15 of 16

..... C R I KWolf Creek Generating Station'NUCLEAR OPERATING CORPORATION RoeO evisiong LimisRpr7. WCAP-1 6009-P-A, "Realistic Large Break LOCA Evaluation MethodologyUsing Automated Statistical Treatment of Uncertainty Method (ASTRUM),"Revision 0, January 2005.NRC letter dated November 5, 2004,"Final Safety Evaluation for WCAP-16009-P, Revision 0, "Realistic Large Break LOCA Evaluation MethodologyUsing Automated Statistical Treatment of Uncertainty Method (ASTRUM)"(TAC NO. MB9483)."8. WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport CodePARAGON," August 2004.NRC Safety Evaluation dated March 18, 2004, "Final Safety Evaluation forWestinghouse Topical Report WCAP-1 6045-P, Revision 0, "Qualification ofthe Two-Dimensional Transport Code PARAGON."9. WCAP-1 6045-P-A, Addendum I-A, "Qualification of the NEXUS Nuclear DataMethodology," August 2007.NRC Safety Evaluation dated February 23, 2007, "Final Safety Evaluation forWestinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16045-P-A, Addendum 1, "Qualification of the NEXUS Nuclear DataMethodology" (TAC NO. MC9606)."10. WCAP 1 0965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code,"September 1986.NRC letter dated June 23, 1986, "Acceptance for Referencing of TopicalReport WCAP 10965-P and WCAP 10966-NP."11. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," April 1995.NRC Safety Evaluation Reports dated July 1, 1991, "Acceptance forReferencing of Topical Report WCAP-1 2610, 'VANTAGE+ Fuel AssemblyReference Core Report' (TAC NO. 77258)."NRC Safety Evaluation Report dated September 15, 1994, "Acceptance forReferencing of Topical Report WCAP-12610, Appendix B, Addendum 1,'Extended Burnup Fuel Design Methodology and ZIRLO Fuel PerformanceModels' (TAC NO. M8641 6)."12. WCAP-8745-P-A, "Design Bases for the Thermal Overpower AT and ThermalOvertemperature AT Trip Function." September 1986.NRC Safety Evaluation Report dated April 17, 1986, "Acceptance forReferencing of Licensing Topical Report WCAP-8745(P)/8746(NP), 'DesignBases for the Thermal Overpower AT and Thermal Overtemperature AT TripFunctions."'"Page 16 of 16