ML13114A292

McGuire, Unit 1, Cycle 23, Revision 0, Core Operating Limits Report

ML13114A292
Person / Time
Site:	McGuire
Issue date:	04/10/2013
From:	Capps S D Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
MCC-1553.05-00-0575, Rev 0, MCEI-0400-280, Rev 0
Download: ML13114A292 (34)

Text

DUKE ENERGY.Steven D. Capps Vice President McGuire Nuclear Station Duke Energy MGO1VP 1 12700 Hagers Ferry Road Huntersville, NC 28078 o: 980.875.4805 f: 980.875.4809 Steven. Capps@duke-energy.com April 10, 2013 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC McGuire Nuclear Station Docket No. 50-369 Unit 1, Cycle 23, Revision 0 Core Operating Limits Report Pursuant to McGuire Technical Specification (TS) 5.6.5.d, please find enclosed the McGuire Unit 1 Cycle 23, Revision 0, 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 Acc)(www.duke-energy.com U. S. Nuclear Regulatory Commission April 10, 2013 Page 2 cc: Mr. John Boska, 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-0400-280 Page 1 Revision 0 McGuire Unit 1 Cycle 23 Core Operating Limits Report Revision 0 February 2013 Calculation Number: MCC-1553.05-00-0575, Revision 0 Duke Energy Date Prepared By: Checked By: Checked By: Approved By: (a (Sections 2.1 and 2.9 -2.17)1/1LI3~(311".3 1 / 3 QA Condition 1 The information presented in this report has been prepared and issued in accordance with McGuire Technical Specification 5.6.5.

MCEI-0400-280 Page 2 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report INSPECTION OF ENGINEERING INSTRUCTIONS Inspection Waived By:spo _ _____Date: 211s_CATAWBA Inspection Waived MCE (Mechanical

& Civil) El Inspected By/Date: RES (Electrical Only) E] Inspected By/Date: RES (Reactor)

[] Inspected By/Date: MOD El Inspected By/Date: Other ( E) l Inspected By/Date: OCONEE Inspection Waived MCE (Mechanical

& Civil) El Inspected By/Date: RES (Electrical Only) El Inspected By/Date: RES (Reactor)

[] Inspected By/Date: MOD EL Inspected By/Date: Other ( ) [] Inspected By/Date: MCGUIRE Inspection Waived MCE (Mechanical

& Civil) Inspected By/Date: RES (Electrical Only) _ Inspected By/Date: RES (Reactor) i Inspected By/Date: MOD Inspected By/Date: Other ( ) [] Inspected By/Date:

MCEI-0400-280 Page 3 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report Implementation Instructions For Revision 0 Revision Description and PIP Tracking Revision 0 of the McGuire Unit 1 Cycle 23 COLR contains limits specific to the reload core.The revision is applicable for Unit 1 operation with or without implementation of the measurement uncertainty recapture (MUR) power uprate.There is no PIP associated with this revision.Implementation Schedule The McGuire Unit I Cycle 23 COLR requires the reload 50.59 be approved prior to implementation and fuel loading.Revision 0 may become effective any time during No MODE between cycles 22 and 23, but must become effective prior to entering MODE 6 which starts cycle 23. The McGuire Unit 1 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.

MCEI-0400-280 Page 4 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected 1-32, Appendix A*COLR M1C23 COLR, Rev. 0 0 February 2013* 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-0400-280 Page 5 Revision 0 McGuire 1 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 the COLR are summarized below.TS Number COLR El Section Pae e Technical Specifications COLR Parameter 2.1.1 Reactor Core Safety Limits 3.1.1 3.1.3 3.1.4 3.1.5 3.1.5 Shutdown Margin Moderator Temperature Coefficient Rod Group Alignment Limits Shutdown Bank Insertion Limits Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.8 Physics Tests Exceptions 3.2.1 Heat Flux Hot Channel Factor 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor 3.2.3 Axial Flux Difference

3.3.1 Reactor

Trip System Instrumentation Setpoints 3.4.1 RCS Pressure, Temperature, and Flow DNB limits 3.5.1 Accumulators

3.5.4 Refueling

Water Storage Tank 3.7.14 Spent Fuel Pool Boron Concentration

3.9.1 Refueling

Operations

-Boron Concentration 5.6.5 Core Operating Limits Report (COLR)RCS Temperature and Pressure Safety Limits Shutdown Margin MTC Shutdown Margin Shutdown Margin Shutdown Bank Insertion Limit Shutdown Margin Control Bank Insertion Limit Shutdown Margin Fq, AFD, OTAT and Penalty Factors FAH, AFD and Penalty Factors AFD OTAT and OPAT Constants RCS Pressure, Temperature and Flow Max and Min Boron Conc.Max and Min Boron Conc.Min Boron Concentration Min Boron Concentration Analytical Methods 2.2 2.3 2.2 2.2 2.4 2.2 2.5 2.2 2.6 9 11 9 9 11 9 15 9 15 2.1 9 2.7 20 2.8 2.9 2.10 2.11 2.12 2.13 2.14 21 24 26 26 26 28 28 1.1 6 The Selected Licensee Commitments that reference this report are listed below: COLR El Section Page SLC Number Selected Licensing Commitment COLR Parameter 16.9.14 16.9.11 16.9.7 Borated Water Source -Shutdown Borated Water Volume and Conc. for BAT/RWST Borated Water Source -Operating Borated Water Volume and Conc. for BAT/RWST Standby Shutdown System Standby Makeup Pump Water Supply 2.15 2.16 2.17 29 30 30 MCEI-0400-280 Page 6 Revision 0 McGuire 1 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," (W Proprietary).

Revision 0 Report Date: July 1985 Not Used for M1C23 2. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," (W Proprietary).

Revision 0 Report Date: August 1985 3. WCAP-10266-P-A, "The 1981 Version Of Westinghouse Evaluation Model Using BASH Code", (W Proprietary).

Revision 2 Report Date: March 1987 Not Used for M1C23 4. WCAP-12945-P-A, Volume 1 and Volumes 2-5, "Code Qualification Document for Best-Estimate Loss of Coolant Analysis," & 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 for M1C23 6. DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," (DPC Proprietary).

Revision 4a Report Date: July 2009 MCEI-0400-280 Page 7 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report 1.1 Analytical Methods (continued)

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," (DPC Proprietary).

Revision la Report Date: December 2008 Not Used for M1C23 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 for M1C23 MCEI-0400-280 Page 8 Revision 0 McGuire 1 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 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 MCEI-0400-280 Page 9 Revision 0 McGuire 1 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-0400-280 Page 10 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation 670 DO NOT OPERATE IN THIS AREA 660 650 640 2400 ps~ia 630 2280 psia~)620)r 610210 a 1945 psia 600 590 ACCEPTABLE 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power MCEI-0400-280 Page 11 Revision 0 McGuire 1 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/0 F.2.3.3 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.

MCEI-0400-280 Page 12 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level'U CSe 1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 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/1/A/6100/22 Unit 1 Data Book for details.

MCEI-0400-280 Page 13 Revision 0 McGuire I Cycle 23 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power 231 220 200:180 160 u 140' 120 a 100 80 60 o 40 20 0 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 CCR[L = 2.3(P) +47 {0<P < 76.1) for CCPJL = 222 {76.1 <P < 100)Bank CB RIL = 2.3((P) +163 {0 < P < 25.7} for CBR JL = 222 {25.7 < P < 100}where P = %Rated Thermal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.Refer to OP/1/A/6100/22 Unit 1 Data Book for details.

MCEI-0400-280 Page 14 Revision 0 McGuire 1 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 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 BankD Fully Withdrawvn at 223 Steps Control Control Control Control BankA Bank B Bank C BankD 0 Start 0 0 0 116 0 Start 0 0 223 Stop 107 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 BankA Bank B Bank C BankD 0 Start 0 0 0 116 0 Start 0 0 224 Stop 108 0 0 224 116 0 Start 0 224 224 Stop 108 0 224 224 116 0 Start 224 224 224 Stop 108 0 Start 0 0 0 116 0 Start 0 0 225 Stop 109 0 0 225 116 0 Start 0 225 225 Stop 109 0 225 225 116 0 Start 225 225 225 Stop 109 Fully Withdrawn at 226 Steps Control Control Control Control BankA BankB BankC BankD 0 Start 0 0 0 116 0 Start 0 0 226 Stop 110 0 0 226 116 0 Start 0 226 226 Stop 110 0 226 226 116 0 Start 226 226 226 Stop 110 Fully Withdrawn at 228 Steps Control Control Control Control BankA BankB BankC BankD Fully Withdrawn at 227 Steps Control Control Control Control BankA BankB BankC BankD 0 Start 0 0 0 116 0 Start 0 0 227 Stop 111 0 0 227 116 0 Start 0 227 227 Stop 111 0 227 227 116 0 Start 227 227 227 Stop 111 Fully Withdrawn at 229 Steps Control Control Control Control BankA BankB BankC BankD 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 230 Steps Control Control Control Control BankA BankB BankC BankD 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 231 Steps Control Control Control Control BankA BankB BnnkC BankD 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 Slop 114 0Start 0 0 0 116 0 Start 0 0 231 Stop 115 0 0 231 116 0 Start 0 231 231 Stop 115 0 231 231 116 0 Start 231 231 231 Stop 115 MCEI-0400-280 Page 15 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report 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 1.2.6 Heat Flux Hot Channel Factor -FQ(XY,Z) (TS 3.2.1)2.6.1 FQ(X,Y,Z) steady-state limits are defined by the following relationships:

F R T , *K(Z)/P for P > 0.5 F T P *K(Z)/0.5 for P < 0.5 where, P = (Thermal Power)/(Rated Power)Note: The measured Fo(X,Y,Z) 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 COLR Sections 2.6.5 and 2.6.6.2.6.2 =2.70 x K(BU)2.6.3 K(Z) is the normalized F 0 (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 F 0 (X,Y,Z) as a function of burnup. F'TP with the K(BU)penalty for Westinghouse RFA fuel is analytically confirmed in cycle-specific reload calculation.

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: 6. FF,(X,Y,Z)

• MQ(X'YZ)2.6.5 Fo(XYZ)OP

= UMT

• MT

• TILT MCEI-0400-280 Page 16 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report where: F (XY,Z)OP = Cycle dependent maximum allowable design peaking factor that ensures F 0 (X,Y,Z) LOCA limit is not exceeded for operation within the AFD, RIL, and QPTR limits.FL OP FQ (X,Y,Z) includes allowances for calculation and measurement uncertainties.

MQ(X,Y,Z)

=Design power distribution for Fo. F D (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.

Mo(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 Fo(X,Y,Z)

=Fo(X,Y,Z)

• Mc(X,Y,Z)UMT

• MT

• TILT where: FLQ(X,y,Z)RPS

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

Centerline Fuel Melt (CFM) limit is not exceeded for operation within the AFD, RIL, and QPTR limits.FLQ(X,Y,Z)RPS includes allowances for calculation and measurement uncertainties.

Design power distributions for Fo. Fgo(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.

F'.(X,Y,Z)

=

MCEI-0400-280 Page 17 Revision 0 McGuire 1 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 RPS compensate for each 1% that Fm' (X,Y,Z) exceeds FL (X,Y,Z) .2.6.8 FQ(X,Y,Z) penalty factors for Technical Specification Surveillances 3.2.1.2 and 3.2.1.3 are provided in Table 2.

MCEI-0400-280 Page 18 Revision 0 McGuire 1 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 1.000 0.800 S0.600 0.400 (0.0, 1.00) (4.0,1.00)

((12.0,0.9259)

(4.0, 0.9259)Core Heg,-ht Uft~ RU 0.0 1.0< &.0 1.0> 4.0 0.9259 12.0 0.9259.0.200 0.000 0.0 2.0 4.0 6.0 Core Height (ft)8.0 10.0 12.0 MCEI-0400-280 Page 19 Revision 0 McGuire I 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 75 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 486 491 496 506 516 FQ(X,Y,Z)Penalty Factor (%)2.00 3.02 2.66 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 FAH(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 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.

MCEJ-0400-280 Page 20 Revision 0 McGuire I 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 FAH(X,Y)Wo=

MARP (XY)* 1.0 + RRH * (1.0- P)]where: F, (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.Thermal Power Rated Thermal Power RRH = Thermal Power reduction required to compensate for each 1% that the measured radial peak, FH (X,Y), exceeds its limit.(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 FALH (X,Y)-suRv F6H (X,Y) x MaH(X,Y)UMR x TILT where: FA (X,Y)su = Cycle dependent maximum allowable design peaking factor that ensures FaH(X,Y) limit is not exceeded for operation SURV within the AFD, RIL, and QPTR limits. FkH (XY)includes allowances for calculation and measurement uncertainty.

MCEI-0400-280 Page 21 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report D FAH (X,Y) = Design radial power distribution for F.AH. Fn (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 MAH(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, F~m (X,Y) exceeds its limit. (0 < P < 1.0)2.7.4 TRH = 0.04 where: TRH = Reduction in the OTAT K 1 setpoint required to compensate for each 1%that the measured radial peak, FAH (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.

MCE1-0400-280 Page 22 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPS)RFA Fuel 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-0400-280 Page 23 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits 0.0'0 C.-0 a-50 30 10 0 10 20 30 40 50 Axial Flux Difference

(% Delta 1)NOTE: Compliance with Technical Specification 3.2.1 may require more restrictive AFD limits. Refer to OP/1/A/6100/22 Unit I Data Book for more details.

MCEI-0400-280 Page 24 Revision 0 McGuire 1 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 4÷Overtemperature AT reactor trip heatup setpoint penalty coefficient Overtemperature 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 Tavg lag compensator fl(AI) "positive" breakpoint fl(AI) "negative" breakpoint fl(AI) "positive" slope* fl(AI) "negative" slope T' < 585.1°F F = 2235 psig Ki < 1.1978 K2 = 0.0334/°F K3 = 0.001601/psi t 1 > 8 sec.-2 < 3 sec.'r 3 < 2 sec.'t 4 > 28 sec.-t 5 < 4 sec.'T 6 < 2 sec.= 19.0 %AI= N/A*= 1.769 %AT 0/ %AI= N/A*The fl(AI) negative breakpoints and slopes for OTAT 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(AI) negative breakpoint and slope unnecessary.

++ AT 0 is assumed to be renormalized to 100% RTP following the MUR power uprate.

MCEI-0400-280 Page 25 Revision 0 McGuire 1 Cycle 23 Core Oper 2.9.2 Overpower AT Setpoint Parameter V: 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 ating Limits Report alues Value T" < 585.1°F K 4 < 1.0864 K 5 = 0.02/°F for increasing Tavg K 5 = 0.0 for decreasing Tavg K 6 = 0.001179/0 F for T > T" K 6 = 0.0 for T < T--c > 8 sec."t 2 < 3 sec.C 3 < 2 sec."c 6 < 2 sec."t 7 > 5 sec.= 35.0 %AI= -35.0 %AI= 7.0 %AT/ %AI= 7.0 %ATc/ %AI++ ATo is assumed to be renormalized to 100% RTP following the MUR power uprate.

MCEI-0400-280 Page 26 Revisiun 0 McGuire 1 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 MODE 3 with RCS pressure >1000 psi: Parameter Accumulator minimum boron concentration.

Accumulator minimum boron concentration.

Accumulator minimum boron concentration.

Accumulator minimum boron concentration.

Accumulator minimum boron concentration.

Accumulator minimum boron concentration.

Accumulator minimum boron concentration.

Accumulator minimum boron concentration.

Accumulator maximum boron concentration.

Applicable Burnup 0 -200 EFPD 200.1 -250 EFPD 250.1 -300 EFPD 300.1 -350 EFPD 350.1 -400 EFPD 400.1 -450 EFPD 450.1 -506 EFPD 506.1 -516 EFPD Limit 2,475 ppm 2,414 ppm 2,347 ppm 2,272 ppm 2,206 ppm 2,142 ppm 2,080 ppm 2,007 ppm 2,875 ppm 0 -516 EFPD 2.12 Refueling Water Storage Tank -RWST (TS 3.5.4)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-0400-280 Page 27 Revision 0 McGuire 1 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 OF meter 3 < 586.9 OF computer 4 < 587.7 OF computer 3 < 587.5 OF 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 > 390,000 gpm**Note: The RCS minimum coolant flow rate assumed in the licensing analyses for the M1C23 core is 388,000 gpm. However, the flow is set at 390,000 gpm, which is conservative.

MCEI-0400-280 Page 28 Revision 0 McGuire 1 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 Minimum boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity.Limit 2,675 ppm MCEI-0400-280 Page 29 Revision 0 McGuire 1 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 bumup is > 455 EFPD, Figure determine required BAT minimum level.6 may be used toI 1 I I I 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-0400-280 Page 30 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report 2.16 Borated Water Source -Operating (SLC 16.9.11)2.16.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODES 1, 2, 3, and MODE 4 with all RCS cold leg temperature

> 300 'F.Parameter BAT minimum contained borated water volume Limit 22,049 gallons 38.0% Level Note: When cycle burnup is > 455 EFPD, Figure 6 may be used to determine required BAT minimum level. I 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 maximum boron concentration (TS 3.5.4)RWST minimum water volume required to maintain SDM at 2,675 ppm 7,000 ppm 13,750 gallons 96,607 gallons 103.6 inches 2,675 ppm 2,875 ppm 57,107 gallons 2.17 Standby Shutdown System -(SLC-16.9.7) 2.17.1 Minimum boron concentration limit for the spent fuel pool required for Standby Makeup Pump Water Supply. Applicable for MODES 1, 2, and 3.Parameter Limit Spent fuel pool minimum boron concentration for TR 2,675 ppm 16.9.7.2.

MCEI-0400-280 Page 31 Revision 0 McGuire 1 Cycle 23 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus RCS Boron Concentration (Valid When Cycle Burnup is > 455 EFPD)This figure includes additional volumes listed in SLC 16.9.14 and 16.9.11 40.0 -.~ii RCS Boron 35.0 -Concentration BAT Level I (ppm) (%level)0<300 37.0 300 < 500 33.0 30.0 ..--- __ _500 < 700 28.0 700 < 1000 23.0 1000 < 1300 13.6> 1300 8.7 2 0 .0 ..... ... ... .- >Acceptable 15.0 10.0 -i m Unacceptable Operation 5.0. 0 200 400 600 800 1000 1200 1400 1600 o180 2000 2200 2400 2600 2800 RCS Boron Concentration (ppmb)

MCEI-0400-280 Page 32 Revision 0 McGuire 1 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 1 Cycle 23 Maneuvering Analysis calculation file, MCC-1553.05-00-0571.

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 information.

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