ML11110A010

Cycle 21, Core Operating Limits Report

ML11110A010
Person / Time
Site:	Mcguire
Issue date:	04/12/2011
From:	Repko R Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
MCEI-0400-249, Rev 0
Download: ML11110A010 (34)

Text

.P Duke Vice President REGIS T. REPKO or Energy@ McGuire Nuclear Station Duke Energy MGOIVP / 12700 Hagers Ferry Rd.

Huntersville, NC 28078 980-875-4111 980-875-4809 fax regis.repko@duke-energy.com April 12, 2011 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC (Duke)

McGuire Nuclear Station Docket Nos. 50-370 Unit 2, Cycle 21 Core Operating Limits Report Pursuant to McGuire Technical Specification (TS) 5.6.5.d, please find enclosed the McGuire Unit 2 Cycle 21 Core Operating Limits Report (COLR).

Questions regarding this submittal should be directed to Kay Crane, McGuire Regulatory Compliance at (980) 875-4306.

Regis T. Repko Attachment www. duke-eoergy. com

U. S. Nuclear Regulatory Commission April 12, 2011 Page 2 cc: Mr. Jon H. Thompson, 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. Joe Brady Senior Resident Inspector McGuire Nuclear Station

MCEI-0400-249 Page I Revision 0 McGuire Unit 2 Cycle 21 Core Operating Limits Report Revision 0 January 2011 Calculation Number: MCC- 1553.05-00-0533, Revision 0 Duke Energy Date Prepared By: 4e,,

Checked By:

Checked By:

(Sections 2.2 and 2. 6 2 8)

Approved By: 2I t QA Condition 1 The information presented in this report has been prepared and issued in accordance with McGuire Technical Specification 5.6.5.

MCEI-0400-249 Page 2 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report INSPECTION OF ENGINEERING INSTRUCTIONS Inspection Waived By:_ RC WA4,1ý Date**/1 1 I-I (Sponsor) U CATAWBA Inspection Waived MCE (Mechanical & Civil) [E Inspected By/Date:

RES (Electrical Only) El Inspected By/Date:

RES (Reactor) [E 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) EL Inspected By/Date:

MOD E] Inspected By/Date:

Other ( El) Inspected By/Date:

MCGUIRE Inspection Waived MCE (Mechanical & Civil) 13 Inspected By/Date:

RES (Electrical Only) [* Inspected By/Date:

RES (Reactor) I Inspected By/Date:

MOD[E" Inspected By/Date:

Other ( ) [] Inspected By/Date:

MCEI-0400-249 Page 3 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report Implementation Instructions For Revision 0 Revision Description and PIP Tracking Revision 0 of the McGuire Unit 2 Cycle 21 COLR contains limits specific to the reload core.

There is no PIP associated with this revision.

Implementation Schedule Revision 0 may become effective any time during No MODE between cycles 20 and 21 but must become effective prior to entering MODE 6 which starts cycle 21. The McGuire Unit 2 Cycle 21 COLR will cease to be effective during No MODE between cycle 21 and 22.

Data rides to be Implemented No data files are transmitted as part of this document.

MCEI-0400-249 Page 4 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected COLR 0 January 2011 1-32, Appendix A* M2C21 COLR, Rev. 0

• 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-0400-249 Page 5 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 1.0 Core Operating Limits Report This Core Operating Limits Report (CO LR) 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 COLR El Number Technical Specifications COLR Parameter Section Paze 1.1 Requirements for Operational MODE 6 MODE 6 Definition 2.1 9 2.1.1 Reactor Core Safety Limits RCS Temperature and 2.2 9 Pressure Safety Limits 3.1.1 Shutdown Margin Shutdown Margin 2.3 9 3.1.3 Moderator Temperature Coefficient MTC 2.4 11 3.1.4 Rod Group Alignment Limits Shutdown Margin 2.3 9 3.1.5 Shutdown Bank Insertion Limits Shutdown Margin 2.3 9 3.1.5 Shutdown Bank Insertion Limits Shutdown Bank Insertion 2.5 11 Limit 3.1.6 Control Bank Insertion Limits Shutdown Margin 2.3 9 3.1.6 Control Bank Insertion Limits Control Bank Insertion 2.6 15 Limit 3.1.8 Physics Tests Exceptions Shutdown Margin 2.3 9 3.2.1 Heat Flux Hot Channel Factor Fq, AFD, OTAT and 2.7 15 Penalty Factors 3.2.2 Nuclear Enthalpy Rise Hot Channel FAH, AFD and 2.8 20 Factor Penalty Factors 3.2.3 Axial Flux Difference AFD 2.9 21 3.3.1 Reactor Trip System Instrumentation OTAT and OPAT 2.10 24 Constants 3.4.1 RCS Pressure, Temperature, and Flow RCS Pressure, 2.11 26 DNB limits Temperature and Flow 3.5.1 Accumulators Max and Min Boron Conc. 2.12 26 3.5.4 Refueling Water Storage Tank Max and Min Boron Conc. 2.13 26 3.7.14 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.14 28 3.9.1 Refueling Operations - Boron Min Boron Concentration 2.15 28 Concentration 5.6.5 Core Operating Limits Report (COLR) Analytical Methods 1.1 6 The Selected Licensee Commitments that reference this report are listed below:

COLR El SLC Number Selected Licensing Commitment COLR Parameter Section Page 16.9.14 Borated Water Source - Shutdown Borated Water Volume and 2.16 29 Conc. for BAT/RWST 16.9.11 Borated Water Source - Operating Borated Water Volume and 2.17 30 Conc. for BAT/RWST 16.9.7 Standby Shutdown System Standby Makeup Pump 2.18 30 Water Supply

MCEI-0400-249 Page 6 Revision 0 McGuire 2 Cycle 21 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 M2C21

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 M2C21

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 I (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 M2C21

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

Revision 4a Report Date: July 2009

MCEI-0400-249 Page 7 Revision 0 McGuire 2 Cycle 21 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 Oa Report Date: May 2009

8. DPC-NE-3002-A, "UFSAR Chapter 15 System Transient Analysis Methodology".

Revision 4a Report Date: April 2009

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-NIE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3," (DPC Proprietary).

Revision la Report Date: December 2008 Not Used for M2C21

12. DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report," (DPC Proprietary).

Revision 2a Report Date: July 2009

13. DPC-NE-1 004A, "Nuclear Design Methodology Using CASMO-3/SIMULATE-3P."

Revision la Report Date: January 2009 Not Used for M2C21

MCEI-0400-249 Page 8 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 1.1 Analytical Methods (continued)

14. DPC-NF-201 0-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-201 -PA, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," (DPC Proprietary).

Revision I a Report Date: June 2009

16. DPC-NE-1005-PA, "Nuclear Design Methodology Using CASMO-4 / SIMIULATE-3 MOX,"

(DPC Proprietary).

Revision 1 Report Date: November 12, 2008

MCEI-0400-249 Page 9 Revision 0 McGuire 2 Cycle 21 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 Requirements for Operational MODE 6 The following condition is required for operational MODE 6.

2.1.1 Reactivity condition requirement for operational MODE 6 is that keff must be less than, or equal to 0.95.

2.2 Reactor Core Safety Limits (TS 2.1.1) 2.2.1 The Reactor Core Safety Limits are shown in Figure 1.

2.3 Shutdown Margin - SDM (TS 3.1.1, TS 3.1.4, TS 3.1.5, TS 3.1.6 and TS 3.1.8) 2.3.1 ForTS 3.1.1, SDM shall be> 1.3% AK/K in MODE 2 with k-eff < 1.0 and in MODES 3 and 4.

2.3.2 For TS 3.1.1, SDM shall be > 1.0% AK/K in MODE 5.

2.3.3 For TS 3.1.4, SDM shall be > 1.3% AK/K in MODES 1 and 2.

2.3.4 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.

2.3.5 For TS 3.1.6, SDM shall be > 1.3% AK/K in MODE I and MODE 2 with K-eff> 1.0.

2.3.6 ForTS 3.1.8, SDM shall be > 1.3% AK/K in MODE 2 during PHYSICS TESTS.

MCEI-0400-249 Page 10 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation 670 DO NOT OPERATE IN THIS AREA 660 650 640 240psa 0 630 2280 psia U 620 610 1 1.945 psla*,*

600 590 ACCEPTABLE 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power

MCEI-0400-249 Page 11 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.4 Moderator Temperature Coefficient - MTC (TS 3.1.3) 2.4.1 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 AKJK/°F.

EOC, ARO, RTP MTC shall be less negative than the -4.3E-04 AK/K/°F lower MTC limit.

2.4.2 300 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.

2.4.3 60 PPM MTC Surveillance Limit is:

60 PPM ARO, equilibrium RTP MTC shall be less negative than or equal to

-4.125E-04 AKIK/°F.

Where: BOC Beginning of Cycle (bumup 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) 2.5 Shutdown Bank Insertion Limit (TS 3.1.5) 2.5.1 Each shutdown bank shall be withdrawn to at least 222 steps except under the conditions listed in Section 2.5.2. Shutdown banks are withdrawn in sequence and with no overlap.

2.5.2 Shutdown banks may be inserted to 219 steps withdrawn individually for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided the plant was operated in steady state conditions near 100% FP prior to and during this exception.

MCEI-0400-249 Page 12 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 0.9 0.8 0.7 E 0.6 0.

= 0.2

~0.0 0 10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.

Refer to OP/2/A16100/22 Unit 2 Data Book for details.

MCEI-0400-249 Page 13 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power 231 220 200 r.180

• 160

• 140

• - 120

-" 100 C 80 60 N 60 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 CCRIL = 2.3(P) +47 (O <P <76.1] for CCRIL =222(76.1 <P< 100)

Bank CB RIL = 2.3(P) +163 (0 < P < 25.7) for CB RIL = 222 (25.7 < P < 100}

where P = %Rated Thermal Power NOTES: (1) 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.

(2) Anytime any shutdown bank or control banks A, B, or C are inserted below 222 steps withdrawn, control bank D insertion is limited to > 200 steps withdrawn (see Sections 2.5.2 and 2.6.2)

MCEI-0400-249 Page 14 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report Table 1 RCCA Withdrawal Steps and Sequence Fully Withdrawn at 222 Steps Fully Withdrawn at 223 Steps Control Control Control Control Control Control Control Control BankA Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start. 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 222 Stop 106 0 0 223 Stop 107 0 0 222 116 0 Start 0 223 116 0 Start 0 222 222 Stop 106 0 223 223 Stop 107 0 222 222 116 0 Start 223 223 116 0 Start 222 222 222 Stop 106 223 223 223 Stop 107 Fully Withdrawn at 224 Steps Fully Withdrawn at 225 Steps Control Control Control Control Control Control Control Control Bank A Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start .0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 224 Stop 109 0 0 225 Stop 109 0 0 224 116 0 Start 0 225 116 0 Start 0 224 224 Stop t08 0 225 225 Stop 109 0 224 224 116 0 Start 225 225 116 0 Start 224 224 224 Stop 108 225 225 225 Stop 109 Fully Withdrawn at 226 Steps Fully Withdrawn at 227 Steps Control Control Control Control Control Control Control Control BankA Bank B Bank C BankD Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 22 6 Stop ItO 0 0 227 Stop III 0 0 226 116 0 Start 0 227 116 0 Start 0 22 226 226 Stop I10 0 227 7 Stop 111 0 226 226 116 0 Start 227 227 116 0 Start 226 226 226 Stop 110 227 227 227 Stop Ill Fully Withdrawn at 228 Steps Fully Withdrawn at 229 Steps Control Control Control Control Control Control Control Control BankA BankB -BankC BankD Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 228 Stop 112 0 0 229 Stop 113 0 0 228 116 0 Start 0 229 116 0 Start 0 22 2 228 8 Stop 112 0 229 29 Stop 113 0 228 228 116 0 Start 229 229 116 0 Start 22 8 228 228 Stop 112 229 229 229 Stop 113 Fully Withdrawn at 230 Steps Fully Withdrawn at 231 Steps Control Control Control Control Control Control Control Control Bank A Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 OSlart 0 0 116 0 Start 0 0 230 Stop 114 0 0 231 Stop 115 0 0 230 116 0 Start 0 231 116 0 Start 0 230 230 Stop 114 0 231 231 Stop 115 0 230 230 116 0 Start 231 231 116 0 Slart 23 230 230 0 Stop 114 231 231 231 Stop 115

MCEI-0400-249 Page 15 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.6 Control Bank Insertion Limits (TS 3.1.6) 2.6.1 Control banks shall be within the insertion, sequence, and overlap limits shown in Figure 3 except under the conditions listed in Section 2.6.2. Specific control bank withdrawal and overlap limits as a function of the fully withdrawn position are shown in Table 1.

2.6.2 Control banks A, B, or C may be inserted to 219 steps withdrawn individually for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided the plant was operated in steady state conditions near 100% FP prior to and during this exception.

2.7 Heat Flux Hot Channel Factor - FQ(X,Y,Z) (TS 3.2.1) 2.7.1 FQ(X,Y,Z) steady-state limits are defined by the following relationships:

FQRU' *K(Z)/P for P > 0.5 F REP

• K(Z)/0.5 for P < 0.5 where, P = (Thermal Power)/(Rated Power)

Note: Measured FQ(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 FQ surveillance limits as defined in COLR Sections 2.7.5 and 2.7.6.

2.7.2 F QRTP= 2.70 x K(BU) 2.7.3 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.

2.7.4 K(BU) is the normalized FQ(X,Y,Z) as a function of burnup. K(BU) for Westinghouse RFA fuel is 1.0 for all burnups.

The following parameters are required for core monitoring per the Surveillance Requirements of Technical Specification 3.2.1:

F(L (XYZYZ)°P -FQ(X,Y,Z)

• MQ(X,Y,Z) 2.7.5 UMT

• NIT

• TILT

MCEI-0400-249 Page 16 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report where:

FL (X YZ)OP Cycle dependent maximum allowable design peaking factor that ensures FQ(X,Y,Z) LOCA limit will be preserved for operation within the LCO limits. F0' (XY,Z)OP includes allowances for calculation and measurement uncertainties.

Fr (X,Y,Z)= Design power distribution for FQ. FL (X,Y,Z) is provided in Appendix Table A-I for normal operating conditions, and in Appendix Table A-4 for power escalation testing during initial startup operation.

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

D L RPS FQ(X,Y,Z)

• Mc(X,Y,Z) 2.7.6 FQ(X,Y,Z) -

UMT

• MT

• TILT where:

L(X,Y,Z)R.PS Cycle dependent maximum allowable design peaking factor that ensures FQ(XY,Z) Centerline Fuel Melt (CFM) limit will be preserved for operation within the LCO limits.

FQ(X,Y,Z)RPas includes allowances for calculation and measurement uncertainties.

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

MCEI-0400-249 Page 17 Revision 0 McGuire 2 Cycle 21 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) 2.7.7 KSLOPE = 0.0725 where:

KSLOPE is the adjustment to K1 value from the OTAT trip setpoint required to compensate for each 1% that Fo (X,Y,Z) exceeds F. (X,Y,Z)R 2.7.8 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-0400-249 Page 18 Revision 0 McGuire 2 Cycle 21 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) 1.000

((12.0,0.9259)

(4.0, 0.9259) 0.800 +

iE,0.600 0.400 +

Core Height (ft) IK(Z) 0.0 1.000 0.200 + <4 1.000

>4 0.9259 12.0 0.O259 0.000 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Core Height (ft)

MCEI-0400-249 Page 19 Revision 0 McGuire 2 Cycle 21 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 FQ(X,Y,Z) FAH(X,Y)

(EFPD) Penalty Factor (%) Penalty Factor (%)

0 2.00 2.00 4 2.00 2.00 12 2.00 2.00 25 2.00 2.00 50 2.79 2.00 75 2.00 2.00 100 2.00 2.00 125 2.00 2.00 150 2.00 2.00 175 2.00 2.00 200 2.00 2.00 225 2.00 2.00 250 2.00 2.00 275 2.00 2.00 300 2.00 2.00 325 2.00 2.00 350 2.00 2.00 375 2.00 2.00 400 2.00 2.00 425 2.00 2.00 450 2.00 2.00 475 2.00 2.00 500 2.00 2.00 510 2.00 2.00 523 2.00 2.00 531 2.00 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups. All cycle bumups 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-0400-249 Page 20 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.8 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.

2.8.1 FaH(X,Y)Lco= MARP (X,Y)* [.+ * (1.0- P)+

where:

FaH (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, Fm. (X,Y), exceeds its limit. RRH also is used to scale the MARY limits as a function of power per the [F (X, Y)]wCo 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.

SURV (X,Y)FAl D (XY)Xx M (X,Y) 2.8.2 FE (XY AH _

UMRMx xTILT where:

S URV F,AL (X,Y) = Cycle dependent maximum allowable design peaking factor that ensures the FAH(XY) limit will be preserved for operation within the LCO limits. F.H (X,Y)sURv includes allowances for calculation/measurement uncertainty.

MCEI-0400-249 Page 21 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report FD (X,Y) = Design radial power distribution for FAH" F* (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.

MAIH(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 MA(X,Y).

TILT = Peaking penalty to account for allowable quadrant power tilt ratio of 1.02 (TILT = 1.035).

2.8.3 RRH = 3.34 where:

RRH = Thermal power reduction required to compensate for each 1% that the measured radial peak, F, (X,Y) exceeds its limit. (0 < P S 1.0) 2.8.4 TRH = 0.04 where:

TRH = Reduction in the OTAT KI setpoint required to compensate for each 1%

that the measured radial peak, FA (X,Y) exceeds its limit.

2.8.5 FAH (X,Y) penalty factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2.

2.9 Axial Flux Difference - AFD (TS 3.2.3) 2.9.1 The Axial Flux Difference (AFD) Limits are provided in Figure 5.

MCEI-0400-249 Page 22 Revision 0 McGuire 2 Cycle 21 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-0400-249 Page 23 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits I-0 Cu 5-I-

Cu 0

Cu C) 5-.

a-.

-50 -40 -30 -20 -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/2/A/6100/22 Unit 2 Data Book for more details.

MCEI-0400-249 Page 24 Revision 0 MeGuire 2 Cycle 21 Core Operating Limits Report 2.10 Reactor Trip System Instrumentation Setpoints (TS 3.3.1) Table 3.3.1-1 2.10.1 Overtemperature AT Setpoint Parameter Values Parameter Value Nominal Tavg at RTP T' < 585.1OF Nominal RCS Operating Pressure P= 2235 psig Overtemperature AT reactor trip setpoint KI <1.1978 Overtemperature AT reactor trip heatup setpoint K2 0.0334/°F penalty coefficient Overtemperature AT reactor trip depressurization K3 0.001601/psi setpoint penalty coefficient Time constants utilized in the lead-lag compensator 'E1 > 8 sec.

for AT 't2 < 3 sec.

Time constant utilized in the lag compensator for AT T3 < 2 sec.

Time constants utilized in the lead-lag compensator T4 > 28 sec.

for T,,g "15< 4 sec.

Time constant utilized in the measured Tavg lag 't6 < 2 sec.

compensator f l ( Al) "positive" breakpoint = 19.0 %AI fl(AI) "negative" breakpoint = N/A*

fl(AI) "positive" slope = 1.769 %AT/ %AI fl(Al) "negative" slope = N/A*

The fl(Al) "negative" breakpoints and the fl(Al) "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(A1) limits are reached. This makes implementation of the OTAT fl(AI) negative breakpoint and slope unnecessary.

MCEI-0400-249 Page 25 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.10.2 Overpower AT Setpoint Parameter Values Parameter Value Nominal Tavg at RTP T" < 585.1°F Overpower AT reactor trip setpoint K4 < 1.0864 Overpower AT reactor trip Penalty K5 = 0.02/'F for increasing Tavg K5 = 0.0 for decreasing Tavg Overpower AT reactor trip heatup K6 = 0.001 179/1F forT > T" setpoint penalty coefficient K6 =0.0 for T < T" Time constants utilized in the lead- -I > 8 sec.

lag compensator for AT T2 < 3 sec.

Time constant utilized in the lag T3 < 2 sec.

compensator for AT Time constant utilized in the < 2 sec.

<6 measured Tavg lag compensator Time constant utilized in the rate-lag _C7> 5 sec.

controller for Tavg f 2(AI) "positive" breakpoint = 35.0 %AI f 2(AI) "negative" breakpoint = -35.0 %AI f 2(AI) "positive" slope = 7.0 %ATo/ %AI f2 (AI) "negative" slope = 7.0 %AT 0 / %AI

MCEI-0400-249 Page 26 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.11 RCS Pressure, Temperature and Flow Limits for DNB (TS 3.4.1) 2.11.1 RCS pressure, temperature and flow limits for DNB are shown in Table 4.

2.12 Accumulators (TS 3.5.1) 2.12.1 Boron concentration limits during MODES I and 2, and MODE 3 with RCS pressure >1000 psi:

Parameter Applicable Bumup Limit Accumulator minimum boron 0 - 200 EFPD 2,475 ppm concentration.

Accumulator minimum boron 200.1 - 250 EFPD 2,475 ppm concentration.

Accumulator minimum boron 250.1 - 300 EFPD 2,418 ppm concentration.

Accumulator minimum boron 300.1 - 350 EFPD 2,327 ppm concentration.

Accumulator minimum boron 350.1 - 400 EFPD 2,253 ppm concentration.

Accumulator minimum boron 400.1 - 450 EFPD 2,194 ppm concentration.

Accumulator minimum boron 450.1 - 500 EFPD 2,136 ppm concentration.

Accumulator minimum boron 500.1 - 531 EFPD 2,076 ppm concentration.

Accumulator maximum boron 0 - 531 EFPD 2,875 ppm concentration.

2.13 Refueling Water Storage Tank - RWST (TS 3.5.4) 2.13.1 Boron concentration limits during MODES 1, 2, 3, and 4:

Parameter Limit RWST minimum boron concentration. 2,675 ppm RWST maximum boron concentration. 2,875 ppm

MCEI-0400-249 Page 27 Revision 0 McGuire 2 Cycle 21 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 > 388,000 gpm

MCEI-0400-249 Page 28 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.14 Spent Fuel Pool Boron Concentration (TS 3.7.14) 2.14.1 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 2.15 Refueling Operations - Boron Concentration (TS 3.9.1) 2.15.1 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 2,675 ppm System, the refueling canal, and the refueling cavity.

MCEI-0400-249 Page 29 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.16 Borated Water Source - Shutdown (SLC 16.9.14) 2.16.1 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 > 460 EFPD, Figure 6 may be used to determine required BAT minimum level. I BAT minimum boron concentration 7,000 ppm BAT minimum water volume required to 2,300 gallons maintain SDM at 7,000 ppm 47,700 gallons RWST minimum contained borated water volume 41 inches RWST minimum boron concentration 2,675 ppm RWST minimum water volume required to 8,200 gallons maintain SDM at 2,675 ppm

MCEI-0400-249 Page 30 Revision 0 McGuire 2 Cycle 21 Core Operating Limits Report 2.17 Borated Water Source- Operating (SLC 16.9.11) 2.17.1 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 Limit BAT minimum contained borated water volume 22,049 gallons 38.0% Level Note: When cycle bumup is > 460 EFPD, Figure 6 may be used to determine required BAT minimum level.

BAT minimum boron concentration 7,000 ppm BAT minimum water volume required to 13,750 gallons maintain SDM at 7,000 ppm 96,607 gallons RWST minimum contained borated water volume 103.6 inches RWST minimum boron concentration 2,675 ppm RWST maximum boron concentration (TS 3.5.4) 2,875 ppm RWST minimum water volume required to 57,107 gallons maintain SDM at 2,675 ppm 2.18 Standby Shutdown System - (SLC-16.9.7) 2.18.1 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-0400-249 Page 31 Revision 0 McGuire 2 Cycle 21 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 < 700K. 28.0 700 < 1000 1 23.0 1000 < 1300 13.6

> 1300 8.7 25.0

.. 20.0 Acceptable 15.0 10.0 Unacceptable Operation]

5.0 0.0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 RCS Boron Concentration (ppmb)

,S °

• MCEI-0400-249 Page 32 Revision 0 McGuire 2 Cycle 21 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 21 Maneuvering Analysis calculation file, MCC-I1553.05-00-0528. 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.