ML18283A299

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McGuire, Unit 2, Cycle 26, Core Operating Limits Report, Revision 0
ML18283A299
Person / Time
Site: Mcguire
Issue date: 08/31/2018
From:
Duke Energy Carolinas
To:
Office of Nuclear Reactor Regulation
References
MCEI-0400-368 MCC-1553.05-00-0659, Rev 0
Download: ML18283A299 (31)
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McGuire Unit 2 Cycle 26 Core Operating Limits Report Revision 0 August 2018 Calculation Number: MCC-1553.05-00-0659 , Revision 0 Duk e Energy Carolinas , LLC QA Condition 1 MCEI-0400-368 Page I R evision 0 The jnformation presented in this report has been prepared and issued in acco rdan ce with McGuire Technical Specification

5.6.5. McGuire

2 Cycle 26 Core Operating Limits Report Implementation Instructions for Revision 0 Revision Description and CR Tracking MCEI-0400-368 Page2 Revision 0 Revision O of the McGuire Unit 2 Cycle 26 COLR contains limits specific to the reload core. There is no CR associated with this revision.

The 50.59 AR is 02224883.

Implementation Schedule The McGuire Unit 2 Cycle 26 COLR requires the reload 50.59 be approved prior to implementation and fuel loading. Revision O may become effective any time during No MODE between cycles 25 and 26, but must become effective prior to entering MODE 6 which starts cycle 26. The McGuire Unit 2 Cycle 25 COLR will cease to be effective during No MODE between cycles 25 and 26. Data Files to be Implemented No data files are transmitted as part of this document.

Revision 0 MCEI-0400-368 Page3 Revision 0 McGuire 2 Cycle 26 Core Operating Limits Report REVISION LOG Effective Date August 2018 Pages Affected 1-31, Appendix A* COLR M2C26 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.

McGuire 2 Cycle 26 Core Operating Limits Report MCEI-0400-368 Page4 Revision 0 1.0 Core Operating Limits Report TS Number 2.1.1 3.1.l 3.1.3 3.1.4 3.1.5 3.1.5 3.1.6 3.1.6 3.1.8 3.2.1 3.2.2 3.2.3 3.3.1 3.4.1 3.5.1 3.5.4 3.7.14 3.9.l

  • 5.6.5 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 COLR Methodology (Section Technical Specifications COLR Parameter Section 1.1 Number) Reactor Core Safety Limits RCS Temperature and 2.1 6,7,8,9, 10, 12,15,16, 18, Pressure Safety Limits 19 Shutdown Margin Shutdown Margin 2.2 6,7,8,12, 14,15,16,18, 19 Moderator Temperature Coefficient MTC 2.3 6,7,8, 14,16, 17 Rod Group Alignment Limits Shutdown Margin 2.2 6,7,8,12, 14,15,16,18, 19 Shutdown Bank Insertion Limits Shutdown Margin 2.2 6,7,8, 12, 14,15, 16, 18, 19 Shutdown Bank Insertion Limits Shutdown Bank Insertion 2.4 2,4,6,7,8,9, 10, 12, 14, 15, Limit 16,18,19 Control Bank Insertion Limits Shutdown Margin 2.2 6, 7,8,12, 14, 15,16, 18, 19 Control Bank Insertion Limits Control Bank Insertion 2.5 2,4,6,7,8,9, 10, 12, 14, 15, Limit 16,18,19 Physics Tests Exceptions Shutdown Margin 2.2 6, 7,8,12, 14,15, 16, 18, 19 Heat Flux Hot Channel Factor Fq, AFD, OT~T and 2.6 2,4,6,7,8,9, 10, 12, 15, 16, Penalty Factors 18,19 Nuclear Enthalpy Rise Hot Channel Fm,AFDand 2.7 2,4,6,7,8,9, 10, 12, 15,16, Factor Penalty Factors 18,19 Axial Flux Difference AFD 2.8 2,4,6,7,8,15,l6 Reactor Trip System Instrumentation OT~T and OP~T 2.9 6,7,8,9,10,12,15,16, 18, Setpoints Constants 19 RCS Pressure, Temperature, and Flow RCS Pressure, 2.10 6,7 ,8,9, 10, 12, 18, 19 DNB limits Temperature and Flow Accumulators Max and Min Boron Cone. 2.11 6,7,8,14,16 Refueling Water Storage Tank Max and Min Boron Cone. 2.12 6,7,8,14,16 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.13 6,7,8,14,16 Refueling Operations

-Boron Min Boron Concentration 2.14* 6,7,8,14,16 Concentration Core Operating Limits Report (COLR) Analytical Methods 1.1 None The Selected Licensee Commitments that reference this report are listed below: COLR NRC Approved SLC Selected Licensing Commitment Section Methodology Number COLR Parameter (Section 1.1 Number) 16.9.14 Borated Water Source-Shutdown Borated Water Volume and 2.15 6,7,8,14,16 Cone. for BAT/RWST 16.9.11 Borated Water Source -Operating Borated Water Volume and 2.16 6,7,8,14,16 Cone. for BAT/RWST 16.9.7 Standby Shutdown System Standby Makeup Pump Water 2.17 6,7,8,14,16 Suooly McGuire 2 Cycle 26 Core Operating Limits Report 1.1 Analytical Methods MCEI-0400-368 Page 5 Revision 0 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 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 1 July 1997 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 4. WCAP-12945-P-A, Volume 1 and Volumes 2-5, "Code Qualification Document for Estimate Loss of Coolant Analysis," (W Proprietary).

Revision:

Volume !.(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 McGuire 2 Cycle 26 Core Operating Limits Report 1.1 Analytical Methods (continued)

MCEI-0400-368 Page 6 Revision 0 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 l Report Date: March 2015 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 5 Report Date: March 2016 11. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TAC03," (DPC Proprietary).

Revision 0 Report Date: April 3, 1995 Not Used 12. DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report," (DPC Proprietary).

Revision 3c Report Date: March 2017 13. DPC-NE-1004A, "Nuclear Design Methodology Using CASM0-3/S1MULATE-3P." Revision la Report Date: January 2009 Not Used McGuire 2 Cycle 26 Core Operating Limits Report 1.1 Analytical Methods (continued)

MCEI-0400-368 Page? Revision 0 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 CASM0-4 / 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 18. WCAP-12610-P-A, "VANTAGE+

Fuel Assembly Reference Core Report," (W Proprietary).

Revision 0 Report Date: April 1995 19. WCAP-12610-P-A

& CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOŽ," CY:J... Proprietary).

Revision 0 Report Date: July 2006 McGuire 2 Cycle 26 Core Operating Limits Report 2.0 Operating Limits MCEI-0400-368 Page 8 Revision 0 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% ~K/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% ~K/K in MODE 5. 2.2.3 For TS 3.1.4, SDM shall be~ 1.3% ~K/K in MODES 1 and 2. 2.2.4 For TS 3.1.5, SDM shall be~ 1.3% ~K/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% in MODE 1 and MODE 2 with K-eff~ 1.0. 2.2.6 For TS 3.1.8, SDM shall be~ 1.3% in MODE 2 during PHYSICS TESTS.

McGuire 2 Cycle 26 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation MCEI-0400-368 Page 9 Revision 0 DO NOT OPERA TE IN THIS AREA 660 1------1------1-------+

~~--< 630 l------+------".....,.-+----------jf-----""""';;;::-----------j-~------1

'-' bJ) E-s r:r.i 620 610 0.0 0.2 ACCEPTABLE OPERATION 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power McGuire 2 Cycle 26 Core Operating Limits Report 2.3 Moderator Temperature Coefficient

-MTC (TS 3.1.3) 2.3.1 The Moderator Temperature Coefficient (MTC) Limits are: MCEI-0400-368 Page 10 Revision 0 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 ~K/K/°F. EOC, ARO, RTP MTC shall be less negative than the -4.3E-04 ~°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 Af(/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 a 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

~°F. Where: BOC = Beginning of Cycle (bum up corresponding to the most positive MTC) EOC = End of Cycle ARO = All Rods Out HZP = Hot Zero Power R TP = 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.

1.0 0.9 ..... 0.8 = .... 0.7 I.=: =-u~ 0.6 0 i~ 0.5 t <l 0.4 Q.~ 89 0.3 .. =-..... 0.2 .. "C Q 0.1 0.0 McGuire 2 Cycle 26 Core Operating Limits Report Figure 2 MCEI-0400-368 Page 11 Revision 0 Moderator Temperature Coefficient Upper Limit Versus Power Level Unacceptable Operation Acceptable Operation 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/A/6100/22 Unit 2 Data Book for details.

231 220 200 ---= 180 = ;.. ] 160 ;t:: 140 Q. .... ~120 = .s 100 ;t:: "' 0 80 = 0 ::: ;.. 60 "' = .... "Q 40 0 20 0 McGuire 2 Cycle 26 Core Operating Limits Report Figure 3 MCEI-0400-368 Page 12 Revision 0 Control Bank Insertion Limits Versus Percent Rated Thermal Power Fully Withdrawn (Maximum=

231-~ ---_,_ .... ---------/ * / ./ / / ,,, / / / / Fully Withdrawn

/ , (Minimum=

222) / / Control Bank B --/ ,,, 000%, 161) F / ./ =1 (0%, 163) I ./ / I/ -./ ,,, / / ,, / ./ Control Bank C ..... / ./ '/ ,,,, / ./ / / / / / / / " ,/ / / / / Control Bank D / / / / ,,, / / R (0%,47) I/ / / / f---Fully Inserted / / -(30%, 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 CC RIL = 2.3(P) +47 {O < P.::::. 76.1} for CC RIL = 222 {76.1 < P < 100} Bank CB RIL = 2.3(P) + 163 {O < P.::::. 25. 7} for CB RIL = 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/2/A/6100/22 Unit 2 Data Book for details.

MCEI-0400-368 Page 13 Revision 0 McGuire 2 Cycle 26 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 Bank A BankB BankC BankD Bank A BankB BankC BankD 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 BankB Banke BankD Bank A BankB BankC BankD 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 224 Stop 108 0 0 225 Stop 109 0 0 224 116 0 Start 0 225 116 0 Start 0 224 224 Stop 108 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 Bank A BankB BankC BankD Bank A BankB BankC BankD 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 226 Stop 110 0 0 227 Stop 111 0 0 226 116 0 Start 0 227 116 0 Start 0 226 226 Stop 110 0 227 227 Stop Ill 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 Bank A BankB BankC BankD Bank A BankB BankC BankD 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 .o 229 116 0 Start 0 228 228 Stop 112 0 229 229 Stop 113 0 228 228 116 0 Start 229 229 116 0 Start 228 228 228 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 BankA BankB BankC BankD Bank A BankB BankC BankD 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 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 Start 230 230 230 Stop 114 231 231 231 Stop 115 McGuire 2 Cycle 26 Core Operating Limits Report 2.6 Heat Flux Hot Channel Factor -FQ(X,Y,Z) (TS 3.2.1) MCEI-0400-3 68 Page 14 Revision 0 2.6.1 FQ(X,Y,Z) steady-state limits are defined by the following relationships:

where, F ~TP *K(Z)IP p~TP *K(Z)/0.5 for P > 0.5 for PS 0.5 P = (Thermal Power)/(Rated Power) Note: The measured FQ(X,Y,Z) shal.l 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 Sections 2.6.5 and 2.6.6. 2.6.2 p~TP = 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 FQ(X,Y,Z) as a function ofburnup.

F~TP with the K(BU) 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: L F~(X,Y,Z)

  • M 0 (X,Y,Z) 2.6.5 FQ(X,Y,Z)OP

= UMT *MT* TILT where: MCEI-0400-368 Page 15 Revision 0 McGuire 2 Cycle 26 Core Operating Limits Report Ft (X,Y,Z)OP

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

~OCA limit will be preserved for operation within the LCO limits. Ft (X,Y,z)OP includes allowances for calculation and measurement uncertainties.

Ft(X,Y,Z)

Design power distribution for FQ. Ft (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.

MQ(X,Y,Z)

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) 2.6.6 F~(X,Y,Z)RPS

= F~(X,Y,Z)

  • Mc(X,Y,Z)

UMT *MT* TILT where: F~(X,Y,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. F~(X,Y,Z)RPS includes allowances for calculation and measurement uncertainties.

D FQ(X,Y,Z)

Defined in Section 2.6.5.

McGuire 2 Cycle 26 Core Operating Limits Report MCEI-0400-368 Page 16 Revision 0 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

  • Defined in Section 2.6.5. MT = Defined in Section 2.6.5. TILT = Defined in Section 2.6.5. 2.6.7 KSLOPE = 0.0725 where: KSLOPE is the adjustment to Kt value from the OT~T trip setpoint required to RPS compensate for each 1 % that Ft (X,Y,Z) exceeds Ft (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.

1.200 1.000 0.800 §: 0.600 0.400 0.200 0.000 MCEI-0400-368 Page 17 .Revision 0 McGuire 2 Cycle 26 Core Operating Limits Report (0.0, 1.00) Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for Westinghouse RFA Fuel (4.0, 1.00) -I (4.0, 0.9259) Core Height (ft) K(Z) 0.0 1.0 :5 4.0 1.0 >4.0 0.9259 12.0 0.9259 0.0 2.0 4.0 6.0 8.0 Core Height (ft) (12.0, 0.9259) 10.0 12.0 McGuire 2 Cycle 26 Core Operating Limits Report Table 2 FQ(X,Y,Z) and Fm(X,Y) Penalty Factors MCEI-0400-368 Page 18 Revision 0 For Technical Specification Surveillance's 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burnup FQ(X,Y,Z)

Fm(X,Y) {EFPD) Penalty Factor{%)

Penalty Factor(%)

4 2.00 2.00 12 2.00 2.00 25 2.00 2.00 50 2.00 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 2.00 460 2.00 2.00 475 2.00 2.00 489 2.00 2.00 494 2.00 2.00 499 2.00 2.00 509 2.00 2.00 519 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 F Af{(X,Y) for compliance with the Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.

McGuire 2 Cycle 26 Core Operating Limits Report MCEI-0400-368 Page 19 Revision 0 2.7 Nuclear Enthalpy Rise Hot Channel Factor -F m(X,Y) (TS 3.2.2) F ~H steady-state limits referred to in Technical Specification 3.2.2 is defined by the following relationship.

where: Fl'tt (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, Ff 8 (X,Y), exceeds its limit. RRH also is used to scale the MARP limits as a function of power per the Fk 8 (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. 2:1.2 pL X y SURV =* F~/X, Y) X M~H(X, Y) ~He ' ) UMRxTILT where: Fk 8 (X,Y)sURv

= Cycle dependent maximum allowable design peaking factor that ensures the F ~ttCX, Y) limit will be preserved for operation within the LCO limits. Fk 8 (X,Y)sURv includes allowances for calculation/measurement uncertainty.

MCEI-0400-368 Page 20 Revision 0 McGuire 2 Cycle 26 Core Operating Limits Report F~H (X, Y) = Design radial power distribution for F t.H* 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.

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

M~(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 Mt.H(X,Y).

TILT = Peaking penalty to account for allowable quadrant power tilt ratio of 1.02 (TILT= 1.035). 2.7.3 RRH is defined in Section 2.7.1. 2.7.4 TRH = 0.04 where: TRH = Reduction in the OT~ T K 1 setpoint required to compensate for each 1 % that the measured radial peak, Fk!i (X, Y) exceeds its limit. 2.7.5 F,.,H (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.

Core Ht (ft.) 1.05 1.1 1.2 0.12 1.8092 1.8553 1.9248 1.2 1.8102 1.8540 1.9248 2.4 1.8093 1.8525 1.9312 3.6 1.8098 1.8514 1.9204 4.8 1.8097 1.8514 1.9058 6.0 1.8097 1.8514 1.8921 7.2 1.8070 1.8438 1.8716 8.4 1.8073 1.8319 1.8452 9.6 1.8072 1.8102 1.8093 10.8 1.7980 1.7868 1.7611 11.4 1.7892 1.7652 1.7250 McGuire 2 Cycle 26 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPS) RFAMARPS Axial Peak 1.3 1.4 1.5 1.6 1.7 1.8 1.9146 1.9179 2.0621 2.0498 2.0090 1.9333 1.9146 1.9179 2.1073 2.0191 1.9775 1.9009 1.9146 1.9179 2.0735 1.9953 1.9519 1.8760 1.9146 1.9179 2.0495 1.9656 1.9258 1.8524 1.9146 1.9179 2.0059 1.9441 1.9233 1.8538 1.9212 1.9179 1.9336 1.8798 1.8625 1.8024 1.8930 1.8872 1.8723 1.8094 1.7866 1.7332 1.8571 1.8156 1.7950 1.7359 1.7089 1.6544 1.7913 1.7375 1.7182 1.6572 1.6347 1.5808 1.7163 1.6538 1.6315 1.5743 1.5573 1.5088 1.6645 1.6057 1.5826 1.5289 1.5098 1.4637 1.9 MCEI-0400-368 Page 21 Revision 0 2.1 3.0 1.8625. 1.7780 1.3151 1.8306 1.7852 1.3007 1.8054 1.7320 1.4633 1.7855 1.6996 1.4675 1.7836 1.6714 1.2987 1.7472 1.6705 1.3293 1.6812 1.5982 1.2871 1.6010 1.5127 1.2182 1.5301 1.4444 1.1431 1.4624 1.3832 1.1009 1.4218 1.3458 1.0670 3.25 1.2461 1.2235 1.4616 1.3874 1.2579 1.2602 1.2195 1.1578 1.0914 1.0470 1.0142

-C .., ... ... .., i=.. -50 McGuire 2 Cycle 26 Core Operating Limits Report Figure 5 MCEI-0400-368 Page 22 Revision 0 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits (-18, 100) (+10, 100) Unacceptable Operation 90 Unacceptable Operation 80 Acceptable Operation 70 60 50 (-36, 50) (+21, 50) 40 30 20 10 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/A/6100/22 Unit 2 Data Book for more details.

McGuire 2 Cycle 26 Core Operating Limits Report MCEl-0400-368 Page 23 Revision 0 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 Nominal Tavg at RTP Nominal RCS Operating Pressure Overtemperature Li T reactor trip setpoint Overtemperature Li T reactor trip heatup setpoint penalty coefficient Overtemperature LiT reactor trip depressurization setpoint penalty coefficient Time constants utilized in the lead-lag compensator for LiT Time constant utilized in the lag compensator for Li T Time constants utilized in the lead-lag compensator for Tavg Time constant utilized in the measured Tavg lag compensator f1 (Lil) "positive" breakpoint f1 (Lil) "negative" breakpoint f1 (Lil) "positive" slope f1 (Lil) "negative" slope T' :'.S 585.l °F P' = 2235 psig K 1 ::S 1.1978 K2 = 0.03341°F K3 = 0.001601/psi 1:1:::8sec.

1:2 ::S 3 sec. 1:3 ::S 2 sec. 1:4::: 28 sec. 1:5 ::S 4 sec. 't6 ::S 2 sec. = 19.0 %Lil =NIA* = 1.769 %ti Toi %Lil =NIA* The f1 (Lil) "negative" breakpoint and the f1 (Lil) "negative" slope are less restrictive than the OPLiT f2(Lil) negative breakpoint and slope. Therefore, during a transient which challenges the negative imbalance limits, the OP ti T f2(Lil) limits will result in a reactor trip before the OT Li T f 1 (Lil) limits are reached. This makes implementation of the OTLiT f1 (Lil) negative breakpoint and slope unnecessary.

MCEI-0400-368 Page 24 Revision 0 McGuire 2 Cycle 26 Core Operating Limits Report 2.9.2 Overpower AT Setpoint Parameter Values Parameter Nominal Tavg at RTP Overpower T reactor trip setpoint Overpower T reactor trip Penalty Overpower T reactor trip heatup setpoint penalty coefficient Time constants utilized in the lag compensator for T Time constant utilized in the lag compensator for T Time constant utilized in the measured T avg lag compensator Time constant utilized in the rate-lag controller for T avg fi(~I) "positive" breakpoint fi(Af) "negative" breakpoint fi(~I) "positive" slope fi(~I) "negative" slope T" S 585.1 °F K4 S 1.0864 Ks= 0.02/°F for increasing Tavg Ks= 0.0 for decreasing Tavg K6 = 0.001179/°F for T > T" K6 = 0.0 for TS T"

  • 1 2:: 8 sec. *2 S 3 sec. *3 S 2 sec. *6 S 2 sec. *7 2:: 5 sec. = 35.0 %Af = -35.0 %~I = 7.0 %~Toi %~I = 7.0 %~Toi %~I McGuire 2 Cycle 26 Core Operating Limits Report 2.10 RCS Pressure, Temperature and Flow Limits for DNB (TS 3.4.1) MCEI-0400-368 Page 25 Revision 0 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 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 -494 EFPD 494.1 -509 EFPD 509.l -519 EFPD 0-519 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.

R WST maximum boron concentration.

2,475 ppm 2,475 ppm 2,475 ppm 2,406 ppm 2,290 ppm 2,215 ppm 2,144 ppm 2,074 ppm 2,047 ppm 2,875 ppm 2,675 ppm 2,875 ppm McGuire 2 Cycle 26 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters No. Operable Parameter Indication Channels 1. Indicated RCS Average Temperature meter 4 meter 3 computer 4 computer 3 2. Indicated Pressurizer Pressure meter 4 meter 3 computer 4 computer 3 3. RCS Total Flow Rate MCEI-0400-368 Page 26 Revision 0 Limits .S 587.2 °F .S 586.9 °F .S 587.7 °F .S 587.5 °F :::: 2212.3 psig :::: 2215.0 psig :::: 2209 .1 psig :::: 2211.3 psig :::: 388,000 gpm McGuire 2 Cycle 26 Core Operating Limits Report 2.13 Spent Fuel Pool Boron Concentration (TS 3.7.14) MCEl-0400-3 68 Page 27 Revision 0 2.13.1 Minimum boron concentration limit for the spent fuel pool. Applicable when fuel assemblies are stored in the spent fuel pool. Parameter 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 ofKeff _s 0.95. Parameter Minimum boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity. 2,675 ppm MCEI-0400-368 Page 28 Revision 0 McGuire 2 Cycle 26 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 S 300 °F and MODES 5 and 6. Parameter Note: When cycle burnup is > 464 EFPD, Figure 6 may be used to determine required BAT minimum level. BAT minimum contained borated water volume BAT minimum boron concentration BAT minimum water volume required to maintain SDM at 7,150 ppm RWST minimum contained borated water volume RWST minimum boron concentration RWST minimum water volume required to maintain SDM at 2,675 ppm 10,599 gallons 13.6% Level 7,150 ppm 2,300 gallons 47,700 gallons 41 inches 2,675 ppm 8,200 gallons MCEI-0400-3 68 Page 29 Revision 0 McGuire 2 Cycle 26 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. * *Note: The SLC 16.9.11 applicability is down to Mode 4 temperatures of > 300°F. The minimum volumes calculated support cooldown to 200°F to satis UFSAR Cha ter 9 re uirements.

Parameter Note: When cycle burnup is > 464 EFPD, Figure 6 may be used to determine required BAT minimum level. BAT minimum contained borated water volume BAT minimum boron concentration BAT minimum water volume required to maintain SDM at 7,150 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 2.17 Standby Shutdown System -(SLC-16.9.7) 22,049 gallons 38.0% Level 7,150 ppm 13,750 gallons 96,607 gallons 103.6 inches 2,675 ppm 2,875 ppm 57,107 gallons 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 Spent fuel pool minimum boron concentration for TR 16.9.7.2.

2,675 ppm C ;.,. = '-" ;.,. Em--< McGuire 2 Cycle 26 Core Operating Limits Report Figure 6 Boric Ac.id Storage Tank Indicated Level Versus RCS Boron Concentration (Valid When Cycle Burnup is> 464 EFPD) MCEI-0400-368 Page 30 Revision 0 This figure includes additional volumes listed in SLC 16.9.14 and 16.9.11 40.0 ~----,--------.,------,-----------,----,-----,---------.--~-----,--------,------,-----, 35.0 30.0 25.0 i + 20.0 15.0 ! i 10.0 i -*--_ L ____ -I ! Unacceptable Operation 5.0 --i ---! I ------4-----! / I i -----i Acceptable RCS Boro_n I Concentration I BAT Level (ppm) I (%level) -----0 < 300 ____ L_ 37.0 ---300 < 500 I 33.o ----500 <_700 __ J __ 28.0 ----700 < 1 ooo J 23.o 1000 < 1300 r 13.6

  • 1300 _____ J-= :).7-= __ J __ _ -*------*-* ! I i --____ , ---r --.. ---L i -( _____ -------l ----**1-__ .. ____ , ____ _ 0.0 +---,---.---;----,---;---,---.-------+--+---+---.---;------1 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 RCS Boron Concentration (ppmb)

McGuire 2 Cycle 26 Core Operating Limits Report MCEI-0400-368 Page 31 Revision 0 NOTE: Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance.

This data was generated in the McGuire 2 Cycle 26 Maneuvering Analysis calculation file, MCC-1553.05-00-0652.

Due to the size of the . monitoring factor data, Appendix A is controlled electronically within the Duke document management system and is not included in the Duke internal copies of the COLR. The Plant Reactor Engineering and Support Systems section will control this information via computer file(s) and should be contacted ifthere is a need to access this information.

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

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