MNS-15-041, Cycle 23, Revision 1, Core Operating Limits Report (COLR)

Cycle 23, Revision 1, Core Operating Limits Report (COLR)

ML15177A013
Person / Time
Site:	Mcguire
Issue date:	06/11/2015
From:	Capps S Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
MNS-15-041 MCC-1553.05-00-0595, Rev. 1
Download: ML15177A013 (32)

Text

Steven D. Capps Vice President ENERGY. McGuire Nuclear Station Duke Energy MG01VP 1 12700 Hagers Ferry Road Huntersville, NC 28078 o: 980.875.4805 f: 980.875.4809 Steven. Capps@duke-energy.com June 11,2015 MNS-1 5-041 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC McGuire Nuclear Station Docket No. 50-370 Unit 2, Cycle 23, Revision 1 Core Operating Limits Report Pursuant to McGuire Technical Specification (TS) 5.6.5.d, please find enclosed the McGuire Unit 2 Cycle 23, Revision 1, Core Operating Limits Report (COLR).Questions regarding this submittal should be directed to Kay Crane, Regulatory Affairs at (980) 875-4306.Steven D. Capps Attachment www.duke-energy.com U. S. Nuclear Regulatory Commission June 11,2015 Page 2 cc: Mr. G.E. Miller, Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Victor M. McCree Regional Administrator U. S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 Mr. John Zeiler Senior Resident Inspector McGuire Nuclear Station MCEI-0400-295 Page 1 Revision 1 McGuire Unit 2 Cycle 23 Core Operating Limits Report Revision 1 May 2015 Calculation Number: MCC-1553.05-00-0595, Revision 1 Duke Energy Carolinas, LLC Date Prepared By: Checked By: Checked By: Approved By: C. L. Klein G / fJ(,5 N. D. Stehle r. a. r-e.4-(Sections 2.1 and 2.9 -2.17)M.A.Blom 56 1ýI 1QA Condition 1 The information presented in this report has been prepared and issued in accordance with McGuire Technical Specification 5.6.5.

MCEI-0400-295 Page 2 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Implementation Instructions For Revision 1 Revision Description and PIP Tracking Revision 1 of the McGuire Unit 2 Cycle 23 COLR contains limits specific to the reload core and is revised to include a conditional exemption for the MTC measurement described in SR 3.1.3.2. The power distribution monitoring factors from Appendix A of Revision 0 remain valid and are not transmitted as part of Revision 1.This revision supports CA#4 of PIP M-15-02948.

Implementation Schedule Revision 1 may become effective upon receipt. The McGuire Unit 2 Cycle 23 COLR will cease to be effective during No MODE between cycles 23 and 24.Data files to be Implemented No data files are transmitted as part of this document.Engineering Instruction Inspection Waiver Per EDM-130 "Engineering Drawings", the Engineering Instruction (El) has been waived per Reference "MC- 143 8.88".

MCEI-0400-295 Page 3 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report REVISION LOG Revision 0 Effective Date February 2014 Pages Affected 1-32, Appendix A*COLR M2C23 COLR, Rev. 0 M2C23 COLR, Rev. I 1 May 2015 1-31 Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance.

Appendix A is included only in the electronic COLR copy sent to the NRC.

MCEI-0400-295 Page 4 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 1.0 Core Operating Limits Report This Core Operating Limits Report (COLR) has been prepared in accordance with the requirements of Technical Specification 5.6.5. The Technical Specifications that reference this report are listed below along with the NRC approved analytical methods used to develop and/or determine COLR parameters in Technical Specifications.

NRC Approved TS COLR Methodology (Section Number Technical Specifications COLR Parameter Section 1.1 Number)2.1.1 Reactor Core Safety Limits RCS Temperature and 2.1 6,7,8,9,10,12,15,16 Pressure Safety Limits 3.1.1 Shutdown Margin Shutdown Margin 2.2 6,7,8,12,14,15,16 3.1.3 Moderator Temperature Coefficient MTC 2.3 6,7,8,12,14,16, 17 3.1.4 Rod Group Alignment Limits Shutdown Margin 2.2 6,7,8,12,1.4,15,16 3.1.5 Shutdown Bank Insertion Limits Shutdown Margin 2.2 6,7,8,12,14,15,16 3.1.5 Shutdown Bank Insertion Limits Shutdown Bank Insertion 2.4 2,4,6,7,8,9,10,12,14,15, Limit 16 3.1.6 Control Bank Insertion Limits Shutdown Margin 2.2 6,7,8,12,14,15,16 3.1.6 Control Bank Insertion Limits Control Bank Insertion 2.5 2,4,6,7,8,9,10,12,14,15, Limit 16 3.1.8 Physics Tests Exceptions Shutdown Margin 2.2 6,7,8,12,14,15,16 3.2.1 Heat Flux Hot Channel Factor Fq, AFD, OTAT and 2.6 2,4,6,7,8,9,10,12,15,16 Penalty Factors 3.2.2 Nuclear Enthalpy Rise Hot Channel FAH, AFD and 2.7 2,4,6,7,8,9.10,12,15,16 Factor Penalty Factors 3.2.3 Axial Flux Difference AFD 2.8 2,4,6,7,8,15,16 3.3.1 Reactor Trip System Instrumentation OTAT and OPAT 2.9 6,7,8,9,10,12,15,16 Setpoints Constants 3.4.1 RCS Pressure, Temperature, and Flow RCS Pressure, 2.10 6,7,8,9,10,12 DNB limits Temperature and Flow 3.5.1 Accumulators Max and Min Boron Conc. 2.11 6,7,8,12,14,16 3.5.4 Refueling Water Storage Tank Max and Min Boron Conc. 2.12 6,7,8,12,14,16 3.7.14 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.13 6,7,8,12,14,16 3.9.1 Refueling Operations

-Boron Min Boron Concentration 2.14 6,7,8,12,14,16 Concentration 5.6.5 Core Operating Limits Report (COLR) Analytical Methods 1.1 None The Selected Licensee Commitments that reference this report are listed below: SLC COLR NRC Approved Number Selected Licensing Commitment COLR Parameter Section Methodology Numbr CLR Pramter(Section 1.1 Number)16.9.14 Borated Water Source -Shutdown Borated Water Volume and 2.15 6,7,8,1214,16 Conc. for BAT/RWST 16.9.11 Borated Water Source- Operating Borated Water Volume and 2.16 6,7,8,12,14,16 Conc. for BAT/RWST 16.9.7 Standby Shutdown System Standby Makeup Pump Water Supply 2.17 6 , 1,,2,14,16 MCEI-0400-295 Page 5 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 1.1 Analytical Methods The analytical methods used to determine core operating limits for parameters identified in Technical Specifications and previously reviewed and approved by the NRC as specified in Technical Specification 5.6.5 are as follows.1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," (IM Proprietary).

Revision 0 Report Date: July 1985 Not Used 2. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," (W Proprietary).

Revision 0 Report Date: August 1985 Addendum 2, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," (W Proprietary). (Referenced in Duke Letter DPC-06-101)

Revision I July 1997 3. WCAP-10266-P-A, "The 1981 Version Of Westinghouse Evaluation Model Using BASH Code", (IM Proprietary).

Revision 2 Report Date: March 1987 Not Used 4. WCAP-12945-P-A, Volume I and Volumes 2-5, "Code Qualification Document for Best-Estimate Loss of Coolant Analysis," (W_ Proprietary).

Revision:

Volume 1 (Revision

2) and Volumes 2-5 (Revision 1)Report Date: March 1998 5. BAW-10168P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B&W Proprietary).

Revision 1 SER Date: January 22, 1991 Revision 2 SER Dates: August 22, 1996 and November 26, 1996 Revision 3 SER Date: June 15, 1994 Not Used MCEI-0400-295 Page 6 Revision I McGuire 2 Cycle 23 Core Operating Limits Report 1.1 Analytical Methods (continued)

6. DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," (DPC Proprietary).

Revision 5a Report Date: October 2012 7. DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary).

Revision 0a Report Date: May 2009 8. DPC-NE-3002-A, "UFSAR Chapter 15 System Transient Analysis Methodology".

Revision 4b Report Date: September 2010 9. DPC-NE-2004P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," (DPC Proprietary).

Revision 2a Report Date: December 2008 10. DPC-NE-2005P-A, "Thermal Hydraulic Statistical Core Design Methodology," (DPC Proprietary).

Revision 4a Report Date: December 2008 11. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3 and GDTACO," (DPC Proprietary).

Revision 2 Report Date: August 2012 Not Used 12. DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report," (DPC Proprietary).

Revision 3a Report Date: September 2011 13. DPC-NE-1004A, "Nuclear Design Methodology Using CASMO-3/SIMULATE-3P." Revision la Report Date: January 2009 Not Used MCEI-0400-295 Page 7 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 1.1 Analytical Methods (continued)

14. DPC-NF-2010-A, "Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design." Revision 2a Report Date: December 2009 15. DPC-NE-2011-PA, "Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," (DPC Proprietary).

Revision la Report Date: June 2009 16. DPC-NE-1005-PA, "Nuclear Design Methodology Using CASMO-4 / SIMULATE-3 MOX," (DPC Proprietary).

Revision 1 Report Date: November 12, 2008 17. DPC-NE-1007-PA, "Conditional Exemption of the EOC MTC Measurement Methodology," (DPC and W Proprietary)

Revision 0 Report Date: April 2015 MCEI-0400-295 Page 8 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.0 Operating Limits Cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections.

These limits have been developed using the NRC approved methodologies specified in Section 1.1.2.1 Reactor Core Safety Limits (TS 2.1.1)2.1.1 The Reactor Core Safety Limits are shown in Figure 1.2.2 Shutdown Margin -SDM (TS 3.1.1, TS 3.1.4, TS 3.1.5, TS 3.1.6 and TS 3.1.8)2.2.1 For TS 3.1.1, SDM shall be > 1.3% AK/K in MODE 2 with k-eff < 1.0 and in MODES 3 and 4.2.2.2 For TS 3.1.1, SDM shall be > 1.0% AK/K in MODE 5.2.2.3 For TS 3.1.4, SDM shall be > 1.3% AK/K in MODES 1 and 2.2.2.4 For TS 3.1.5, SDM shall be > 1.3% AK/K in MODE 1 and MODE 2 with any control bank not fully inserted.2.2.5 For TS 3.1.6, SDM shall be > 1.3% AK/K in MODE 1 and MODE 2 with K-eff > 1.0.2.2.6 For TS 3.1.8, SDM shall be > 1.3% AK/K in MODE 2 during PHYSICS TESTS.

MCEI-0400-295 Page 9 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation 3469 MWth Uprated Thermal Power 670 660 DO NOT OPERATE IN THIS AREA 650 640.1I "2400. 80pipsia 0 U)610 600 1945 psia 590 ACCEPTABLE OPERATION 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power MCEI-0400-295 Page 10 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.3 Moderator Temperature Coefficient

-MTC (TS 3.1.3)2.3.1 The Moderator Temperature Coefficient (MTC) Limits are: MTC shall be less positive than the upper limits shown in Figure 2. BOC, ARO, HZP MTC shall be less positive than 0.7E-04 AK/K/°F.EOC, ARO, RTP MTC shall be less negative than the -4.3E-04 AK/K/°F lower MTC limit.2.3.2 300 PPM MTC Surveillance Limit is: Measured 300 PPM ARO, equilibrium RTP MTC shall be less negative than or equal to -3.65E-04 AK/K/°F.2.3.3 The Revised Predicted near-EOC 300 ppm ARO RTP MTC shall be calculated using the procedure contained in DPC-NE-1007-PA If the Revised Predicted MTC is less negative than or equal to the 300 ppm SR 3.1.3.2 Surveillance Limit, and all benchmark data contained in the surveillance procedure is satisfied, then an MTC measurement in accordance with SR 3.1.3.2 is not required to be performed.

2.3.4 60 PPM MTC Surveillance Limit is: 60 PPM ARO, equilibrium RTP MTC shall be less negative than or equal to-4.125E-04 AK/K/0 F.Where: BOC = Beginning of Cycle (burnup corresponding to the most positive MTC)EOC = End of Cycle ARO = All Rods Out HZP = Hot Zero Power RTP = Rated Thermal Power PPM Parts per million (Boron)2.4 Shutdown Bank Insertion Limit (TS 3.1.5)2.4.1 Each shutdown bank shall be withdrawn to at least 222 steps. Shutdown banks are withdrawn in sequence and with no overlap.2.5 Control Bank Insertion Limits (TS 3.1.6)2.5.1 Control banks shall be within the insertion, sequence, and overlap limits shown in Figure 3. Specific control bank withdrawal and overlap limits as a function of the fully withdrawn position are shown in Table 1.

MCEI-0400-295 Page 11 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 0.9 Unacceptable Operation 0t 4 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 Acceptable Operation 0.0 0 10 20 30' 40 50 60 70 80 Percent of Rated Thermal Power 90 100 NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.Refer to OP/2/A/6100/22 Unit 2 Data Book for details.

MCEI-0400-295 Page 12 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power Fully Withdrawn (Maximum = 231)231-S160 .-....140 100 2 lFully Inserted 0 10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power The Rod Insertion Limits (RIL) for Control Bank D (CD), Control Bank C (CC), and Control Bank B (CB) can be calculated by: Bank CD RIL = 2.3(P)- 69 {30<P< 100}Bank CC RIL = 2.3 (P) + 47 {0 _P _76.1]} for CC RIL = 222 [76.1i < P <_ 10}Bank CR RIL = 2.3(P) +163 (0 P <25.7) for GB RIL = 222 {25.7 < P _100]where P = %Ra ted Thermnal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.Refer to 0P/2/A/6100/22 Unit 2 Data Book for details.

MCEI-0400-295 Page 13 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Table 1 RCCA Withdrawal Steps and Sequence Fully Withdrawn at 222 Steps Control Control Control Control BankA BankB BankC BankD Fully Withdrawn at 223 Steps Control Control Control Control BankA BankB BankC BankD 0 Start 0 0 0 116 0 Start 0 0 222 Stop 106 0 0 222 116 0 Start 0 222 222 Stop 106 0 222 222 116 0 Start 222 222 222 Stop 106 Fully Withdrawn at 224 Steps Control Control Control Control BankA BankB BankC Bank!)0 Start 0 0 0 116 0 Start 0 0 223 Stop 117 0 0 223 116 0 Start (0 223 223 Stop 107 0 223 223 116 0 Start 223 223 223 Stop 107 Fully Withdrawn at 225 Steps Control Control Control Control BaokA Bank B BankC Bank!D 0 Start 0 0 0 116 0 Start 0 0 225 Stop 109 0 0 225 116 01 Start 0 225 225 Stop 109 0 225 225 116 0 Start 225 225 225 Stop 109 0 Start 0 0 0 116 0 Start 0 0 224 Stop 111 0 0 224 116 0 Start 0 224 224 Stop 108 0 224 224 116 ) Start 224 224 224 Stop 108 Folly Withdrawn at 226 Steps Control Control Control Control BankA Bank B BankC BankD 0 Start 0 0 0 116 0 Start t 0 226 Stop 110 0 0 226 116 0 Start 0 226 2 2 6 Stop 110 0 226 226 116 0 Start 226 226 226 Stop IIl Fully Withdrawn at 228 Steps Control Control Control Control Bank A Bank!B Bank C Bank!D 0 Start 0 0 0 116 0 Start 0 0 228 Stop 112 0 0 228 116 0 Start 0 228 228 Stop 112 0 228 228 116 0 Start 228 228 228 Stop 112 Fully Withdrawn at 227 Steps Control Control Control Control BankA Bank B Bank C Bank D 0 Start 0 0 0 116 0 Start 0 0 227 Stop I11 0 0 227 116 11 Start 0 227 227 Stop 111 0 227 227 116 0 Start 227 227 227 Stop IIl Fully Withdrawn at 229 Steps Control Control Control Control BankA Bank B BankC Bank D 0 Start 0 0 0 116 0 Start 0 0 229 Stop 113 0 0 229 116 0 Start 0 229 229 Stop 113 0 229 229 116 0 Start 229 229 229 Stop 113 Fully Withdrawn at 230 Steps Control Control Control Control BankA Bank!B BankC Bank!D 0 Start 0 0 0 116 0 Start 0 0 230 Stop 114 0 0 230 116 0 Start 0 230 230 Stop 114 0 230 230 116 0 Start 230 230 230 Stop 114 Fully Withdrawn at 231 Steps Control Control Control Control BankA Bank B Bank C Bank D 0 Start 0 0 116 0 Start 0 0 231 Stop 115 0 0 231 116 0 Start (1 231 231 Stop 115 0 231 231 116 0 Start 231 231 231 Stop 115 MCEI-0400-295 Page 14 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.6 Heat Flux Hot Channel Factor -FQ(X,Y,Z) (TS 3.2.1)2.6.1 Fo(X,Y,Z) steady-state limits are defined by the following relationships:

F/p *K(Z)/P for P > 0.5 F17" r*K(Z)/0.5 for P < 0.5 where, P = (Thermal Power)/(Rated Power)Note: The measured Fo(X,YZ) shall be increased by 3% to account for manufacturing tolerances and 5% to account for measurement uncertainty when comparing against the LCO limits. The manufacturing tolerance and measurement uncertainty are implicitly included in the F 0 surveillance limits as defined in Sections 2.6.5 and 2.6.6.2.6.2 F""" = 2.70 x K(BU)2.6.3 K(Z) is the normalized FQ(X.Y,Z) as a function of core height. The K(Z) function for Westinghouse RFA fuel is provided in Figure 4.2.6.4 K(BU) is the normalized Fo(X,Y,Z) as a function of burnup. F ..... with the K(BU)0 penalty for Westinghouse RFA fuel is analytically confirmed in cycle-specific reload calculations.

K(BU) is set to 1.0 at all burnups.The following parameters are required for core monitoring per the Surveillance Requirements of Technical Specification 3.2.1: D L( FQ(XYZ)

• MQ(XyZ)2.6.5 F (XYZ)OP = UMT

• MT

• TILT MCEI-0400-295 Page 15 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report where: F' (X,Y,Z)OP

=F') (X,Y,Z) =Mo(X,Y,Z)

=Cycle dependent maximum allowable design peaking factor that ensures FQ(X,Y,Z)

LOCA limit will be preserved for operation within the LCO limits. F'j (X,Y,Z)OP includes allowances for calculation and measurement uncertainties.

Design power distribution for F 0.(X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions, and in Appendix Table A-4 for power escalation testing during initial startup operation.

Margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution.

MQ(X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial startup operation.

UMT = Total Peak Measurement Uncertainty. (UMT = 1.05)MT = Engineering Hot Channel Factor. (MT = 1.03)TILT = Peaking penalty to account for allowable quadrant power tilt ratio of 1.02. (TILT = 1.035)L RPS 2.6.6 FQ(X,Y,Z)Fg(X,Y,Z)

• Mc(X,Y,Z)UMT

• MT

• TILT where: FL(XY,Z)RPS Cycle dependent maximum allowable design peaking factor that ensures the FQ(X,Y,Z)

Centerline Fuel Melt (CFM) limit will be preserved for operation within the LCO limits.FQ(X,Y,Z)RPS includes allowances for calculation and measurement uncertainties.

Design power distributions for Fo. F (X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial startup operation.

FQ(X,Y,Z)=

MCEI-0400-295 Page 16 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report MC(X,Y,Z)=

Margin remaining to the CFM limit in core location X,Y,Z from the transient power distribution.

Mc(X,Y,Z) is provided in Appendix Table A-2 for normal operating conditions and in Appendix Table A-5 for power escalation testing during initial startup operation.

UMT = Total Peak Measurement Uncertainty (UMT = 1.05)MT = Engineering Hot Channel Factor (MT = 1.03)TILT = Peaking penalty to account for allowable quadrant power tilt ratio of 1.02. (TILT = 1.035)2.6.7 KSLOPE = 0.0725 where: KSLOPE is the adjustment to K 1 value from the OTAT trip setpoint required to compensate for each 1% that (X,Y,Z) exceeds F, (XY,Z) .2.6.8 Fo(X,YZ) penalty factors for Technical Specification Surveillances 3.2.1.2 and 3.2.1.3 are provided in Table 2.

MCEI-0400-295 Page 17 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for Westinghouse RFA Fuel 1.200 (0.0, 1.00) (4.0, 1.00)(!12.03 0.9259)(4.0, 0.9259)0.800 .N 0.600 0.400 Core Height ft K(Z.0.0 1.0 0.200 < 4.0 1.0> 4.0 0.9259 12.0 0.9259 0.000 !-.....--

._- -0.0 2.0 4.0 6.0 8.0 10.0 12.0 Core Height (ft)

MCEI-0400-295 Page 18 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Table 2 FQ(X,Y,Z) and FAH(X,Y) Penalty Factors For Technical Specification Surveillance's 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burnup (EFPD)0 4 12 25 50 75 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 480 495 505 515 FQ(X,Y,Z)Penalty Factor (%)2.00 2.00 2.00 2.00 2.73 2.06 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.0.0 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 FAHI(X,Y)Penalty Factor (%)2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups. All cycle burnups outside of the range of the table shall use a 2% penalty factor for both FQ(X,Y,Z) and FAH(X,Y) for compliance with the Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.

MCEI-0400-295 Page 19 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.7 Nuclear Enthalpy Rise Hot Channel Factor -FAH(X,Y) (TS 3.2.2)FAH steady-state limits referred to in Technical Specification 3.2.2 is defined by the following relationship.

2.7.1 F6Lj (X, Y)Lc" = MARP (X,Y)* 1.0 + RRH (1.0 -P)]where: FL (X,Y)LcO is the steady-state.

maximum allowed radial peak and includes allowances for calculation/measurement uncertainty.

MARP(X,Y)

= Cycle-specific operating limit Maximum Allowable Radial Peaks. MARP(X,Y) radial peaking limits are provided in Table 3.P Thermal Power Rated Thermal Power RRH = Thermal Power reduction required to compensate for each 1% that the measured radial peak, FM 1 (X,Y), exceeds its limit. RRH also is used to scale the MARP limits as a function of power per the (X, Y) LCO equation. (RRH = 3.34 (0.0 < P < 1.0))The following parameters are required for core monitoring per the surveillance requirements of Technical Specification 3.2.2.2.7.2 F)1 (X,Y)SURV

= -D (Xy)XMAH(XY)

UMR x TILT where: F (X,Y) SURv Cycle dependent maximum allowable design peaking factor AH (X-Y that ensures the Far I(X,Y) limit will be preserved for operation within the LCO limits. FALH (X,Y)sURv includes allowances for calculation/measurement uncertainty.

MCEI-0400-295 Page 20 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report FD FD FAH (X,Y) = Design radial power distribution for FnH. Fam (X,Y) is provided in Appendix Table A-3 for normal operation and in Appendix Table A-6 for power escalation testing during initial startup operation.

MAH(X,Y) = The margin remaining in core location X,Y relative to the Operational DNB limits in the transient power distribution.

MAH(X,Y) is provided in Appendix Table A-3 for normal operation and in Appendix Table A-6 for power escalation testing during initial startup operation.

UMR = Uncertainty value for measured radial peaks (UMR = 1.0).UMR is 1.0 since a factor of 1.04 is implicitly included in the variable MAII(X,Y).

TILT Peaking penalty to account for allowable quadrant power tilt ratio of 1.02 (TILT = 1.035).2.7.3 RRH = 3.34 where: RRH =Thermal power reduction required to compensate for each 1% that the measured radial peak, F2H (X,Y) exceeds its limit. (0 < P < 1.0)2.7.4 TRH = 0.04 where: TRH = Reduction in the OTXT K 1 setpoint required to compensate for each 1%that the measured radial peak, FA", (X,Y) exceeds its limit.2.7.5 FAH (X,Y) penalty factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2.2.8 Axial Flux Difference

-AFD (TS 3.2.3)2.8.1 The Axial Flux Difference (AFD) Limits are provided in Figure 5.

MCEI-0400-295 Page 21 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPS)RFA MARPS Core Axial Peak Ht (ft.) 1.05 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 2.1 3.0 3.25 0.12 1.809 1.855 1.949 1.995 1.974 2.107 2.050 2.009 1.933 1.863 1.778 1.315 1.246 1.2 1.810 1.854 1.940 1.995 1.974 2.107 2.019 1.978 1.901 1.831 1.785 1.301 1.224 2.4 1.809 1.853 1.931 1.978 1.974 2.074 1.995 1.952 1.876 1.805 1.732 1.463 1,462 3.6 1.810 1.851 1.920 1.964 1.974 2.050 1.966 1.926 1.852 1.786 1.700 1.468 1.387 4.8 1.810 1.851 1.906 1.945 1.974 2.006 1.944 1.923 1.854 1.784 1.671 1,299 1,258 6.0 1.810 1.851 1,892 1.921 1.946 1.934 1.880 1.863 1.802 1.747 1.671 1.329 1.260 7.2 1.807 1.844 1.872 1.893 1.887 1.872 1.809 1.787 1.733 1.681 1.598 1.287 1.220 8.4 1.807 1.832 1.845 1.857 1.816 1.795 1.736 1.709 1.654 1.601 1.513 1.218 1.158 9.6 1.807 1.810 1.809 1.791 1.738 1.718 1.657 1.635 1.581 1.530 1.444 1.143 1.091 10.8 1.798 1.787 1.761 1.716 1.654 1.632 1.574 1.557 1.509 1.462 1.383 1.101 1.047 11.4 1.789 1.765 1.725 1.665 1.606 1.583 1.529 1.510 1.464 1.422 1.346 1.067 1.014 MCEI-0400-295 Page 22 Revision I McGuire 2 Cycle 23 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits Unacceptable Operation Al.E G.(-36, 50)-50 30 10 0 10 20 30 40 50 Axial Flux Difference

(% Delta I)NOTE: Compliance with Technical Specification 3.2.1 may require more restrictive AFD limits. Refer to OP/2/A16100/22 Unit 2 Data Book for more details.

MCEI-0400-295 Page 23 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.9 Reactor Trip System Instrumentation Setpoints (TS 3.3.1) Table 3.3.1-1 2.9.1 Overtemperature AT Setpoint Parameter Values Parameter Value Nominal Tavg at RTP Nominal RCS Operating Pressure Overtemperature AT reactor trip setpoint Overtemperature AT reactor trip heatup setpoint penalty coefficient Overternperature AT reactor trip depressurization setpoint penalty coefficient Time constants utilized in the lead-lag compensator for AT Time constant utilized in the lag compensator for AT Time constants utilized in the lead-lag compensator for Tavg Time constant utilized in the measured T,,g lag compensator fl(AI) "positive" breakpoint fl(Al) "negative" breakpoint fl(AI) "positive" slope T' < 585.1 0 F P" = 2235 psig K1 < 1.1978 K2 = 0.0334/°F K3 = 0.001601/psi rI > 8 sec.T2 < 3 sec.'r3 < 2 sec.c4 > 28 sec.T 5 < 4 sec.'6 < 2 sec.= 19.0 %A1= N/A*= 1.769 %AT 0/ %AI fl(AI) "negative" slope= N/A*The fl(Al) "negative" breakpoints and the fI(AI) "negative" slope are less restrictive than the OPAT f 2 (AI) negative breakpoint and slope. Therefore, during a transient which challenges the negative imbalance limits, the OPAT f 2 (AI) limits will result in a reactor trip before the OTAT fl(AI) limits are reached. This makes implementation of the OTAT fl(Al) negative breakpoint and slope unnecessary.

MCEI-0400-295 Page 24 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.9.2 Overpower AT Setpoint Parameter Values Parameter Nominal Tavg at RTP Overpower AT reactor trip setpoint Overpower AT reactor trip Penalty Overpower AT reactor trip heatup setpoint penalty coefficient Time constants utilized in the lead-lag compensator for AT Time constant utilized in the lag compensator for AT Time constant utilized in the measured Tavg lag compensator Time constant utilized in the rate-lag controller for Tavg f 2 (AI) "positive" breakpoint f 2 (AI) "negative" breakpoint f 2 (AI) "positive" slope f 2 (AI) "negative" slope Value T" < 585.1'F K 4 < 1.0864 K 5 = 0.02/°F for increasing Tavg K 5 = 0.0 for decreasing Tavg K6 = 0.001179/°F for T > T-K6 = 0.0 for T < T".r1 > 8 sec.-2 < 3 sec._E3 < 2 sec.L 6 < 2 sec."C7 > 5 sec.= 35.0 %AI= -35.0 %AI= 7.0 %ATC/ %AI= 7.0 %ATo/ %AI MCEI-0400-295 Page 25 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.10 RCS Pressure, Temperature and Flow Limits for DNB (TS 3.4.1)2.10.1 RCS pressure, temperature and flow limits for DNB are shown in Table 4.2.11 Accumulators (TS 3.5.1)2.11.1 Boron concentration limits during MODES 1 and 2, and pressure >1000 psi: Parameter Applicable Burnup Accumulator minimum boron 0 -200 EFPD concentration.

Accumulator minimum boron 200.1 -250 EFPD concentration.

Accumulator minimum boron 250.1 -300 EFPD concentration.

Accumulator minimum boron 300.1 -350 EFPD concentration.

Accumulator minimum boron 350.1 -400 EFPD concentration.

Accumulator minimum boron 400.1 -450 EFPD concentration.

Accumulator minimum boron 450.1 -505 EFPD concentration.

Accumulator minimum boron 505.1 -515 EFPD concentration.

Accumulator maximum boron 0 -515 EFPD concentration.

2.12 Refueling Water Storage Tank -RWST (TS 3.5.4)MODE 3 with RCS Limit 2,475 ppm 2,475 ppm 2,389 ppm 2,282 ppm 2,208 ppm 2,140 ppm 2,072 ppm 1,994 ppm 2,875 ppm 2.12.1 Boron concentration limits during MODES 1, 2, 3, and 4: Parameter RWST minimum boron concentration.

RWST maximum boron concentration.

Limit 2,675 ppm 2,875 ppm MCEI-0400-295 Page 26 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters No. Operable Parameter Indication Channels Limits 1. Indicated RCS Average Temperature meter 4 < 587.2 'F meter 3 < 586.9 'F computer 4 < 587.7 'F computer 3 < 587.5 'F 2. Indicated Pressurizer Pressure meter 4 > 2219.8 psig meter 3 > 2222.1 psig computer 4 > 2215.8 psig computer 3 > 2217.5 psig 3. RCS Total Flow Rate > 388,000 gpm MCEI-0400-295 Page 27 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.13 Spent Fuel Pool Boron Concentration (TS 3.7.14)2.13.1 Minimum boron concentration limit for the spent fuel pool. Applicable when fuel assemblies are stored in the spent fuel pool.Parameter Limit Spent fuel pool minimum boron concentration.

2,675 ppm 2.14 Refueling Operations

-Boron Concentration (TS 3.9.1)2.14.1 Minimum boron concentration limit for the filled portions of the Reactor Coolant System, refueling canal, and refueling cavity for MODE 6 conditions.

The minimum boron concentration limit and plant refueling procedures ensure that core Keff remains within MODE 6 reactivity requirement of Keff < 0.95.Parameter Limit Minimum boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity.2,675 ppm MCEI-0400-295 Page 28 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report 2.15 Borated Water Source -Shutdown (SLC 16.9.14)2.15.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODE 4 with any RCS cold leg temperature

< 300 'F and MODES 5 and 6.Parameter Limit BAT minimum contained borated water volume 10,599 gallons 13.6% Level Note: When cycle burnup is > 460 EFPD, Figure 6 may be used to determine required BAT minimum level.BAT minimum boron concentration BAT minimum water volume required to maintain SDM at 7,000 ppm RWST minimum contained borated water volume RWST minimum boron concentration RWST minimum water volume required to maintain SDM at 2,675 ppm 7,000 ppm 2,300 gallons 47,700 gallons 41 inches 2,675 ppm 8,200 gallons MCEI-0400-295 Page 30 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus RCS Boron Concentration (Valid When Cycle Burnup is > 460 EFPD)This figure includes additional volumes listed in SLC 16.9.14 and 16.9.11 40.0 RCS Boron 35.0 Concentration BAT Level (ppm) (%level)0 < 300 37.0 300 < 500 33.0 30.0 500 < 700 28.0 700 < 1000 23.0 1000 < 1300 13.6 25.0 -> 1300 8.7 20.0>Acceptable 10.0 Unacceptable Operation 5.0 0.0 2600 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2300 RCS Boron Concentration (ppmb)

MCEI-0400-295 Page 31 Revision 1 McGuire 2 Cycle 23 Core Operating Limits Report NOTE: Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance.

This data was generated in the McGuire 2 Cycle 23 Maneuvering Analysis calculation file, MCC-1553.05-00-0590.

Due to the size of the monitoring factor data, Appendix A is controlled electronically within Duke and is not included in the Duke internal copies of the COLR. The Plant Nuclear Engineering Section will control this information via computer file(s) and should be contacted if there is a need to access this infomaation.

Appendix A is included in the COLR copy transmitted to the NRC.