ML14169A493

Core Operating Limits Report for Cycle 22, Revision 1. Part 1 of 2

ML14169A493
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Site:	Catawba
Issue date:	04/22/2014
From:	Hager N Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation
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ML14169A328	List: CNS-14-073, Core Operating Limits Report for Cycle 22, Revision 1. Part 2 of 2 ML14169A331 ML14169A493
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CNS-14-073 CNEI-0400-269, Rev 1
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Catawba Unit 1 Cycle 22 Core Operating Limits Report Revision 1 April2014 CNEI-0400-269 Page l of 31 Revision l Calculation Number: CNC-1553.05-00-0610, Revision 1 Date Prepared By: ;~~~!2~

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Approved By: J::f.: riM.ff::t

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QA Condition 1 The information presented in this report has been prepared and issued in accordance with Catawba Technical Specification 5.6.5.

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CNEI-0400-269 Page 2 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Implementation Instructions for Revision 1 Revision Description and PIP Tracking Revision 1 of the Catawba Unit 1 Cycle 22 COLR contains limits specific to the reload core and is revised to provide explicit implementation instructions to ensure the C1C22 COLR implementation occurs properly for analyzed core conditions. PIP #C-10-08327 is associated with this revision.

Note: Several PIPs resulted in changes to COLR limits in Revision 0. These changes were to COLR Section 2.10 Boron Dilution Mitigation System as a result of PIP #C-10-08327 and COLR Section 2.18 Boration Systems Borated Water Source as a result of PIP #C-14-02092.

Implementation Schedule The Catawba Unit 1 Cycle 22 COLR requires the reload 50.59 be approved prior to implementation and fuel loading.

Revision 1 may become effective any time during No MODE between Cycles 21 and 22 but must become effective prior to entering MODE 6 which starts Cycle 22 AND the C1C22 COLR requires a BDMS Alarm Ratio less than or equal to 2 prior to entering MODE 5; the BDMS Alarm Ratio is being modified for both trains in Unit 1 in EC 110698 during 1EOC21.

The Catawba Unit 1 Cycle 22 COLR will cease to be effective during No MODE between Cycles 22 and 23.

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

Engineering Instruction Inspection Waiver Per EDM-130 Engineering Drawings, the Engineering Instruction (EI) has been waived per Reference CN -1438.88.

CNEI-0400-269 Page 3 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected COLR 0

April 2014 1-31, Appendix A*

C1C22 COLR, Rev. 0 1

April 2014 1-31 C1C22 COLR, Rev. 1

• 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.

CNEI-0400-269 Page 4 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 1.0 Core Operating Limits Report This Core Operating Limits Report (COLR) has been prepared in accordance with requirements of Technical Specification 5.6.5. Technical Specifications that reference this report are listed below along with the NRC approved analytical methods used to develop and/or determine COLR parameters identified in Technical Specifications.

TS Section Technical Specifications COLR Parameter COLR Section NRC Approved Methodology (Section 1.1 Number) 2.1.1 Reactor Core Safety Limits RCS Temperature and Pressure Safety Limits 2.1 6, 7, 8, 9, 10, 12, 15, 16 3.1.1 Shutdown Margin Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16 3.1.3 Moderator Temperature Coefficient MTC 2.3 6, 7, 8, 12, 14, 16 3.1.4 Rod Group Alignment Limits Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16 3.1.5 Shutdown Bank Insertion Limit Shutdown Margin 2.2 2, 4, 6, 7, 8, 9, Rod Insertion Limits 2.4 10, 12, 14, 15, 16 3.1.6 Control Bank Insertion Limit Shutdown Margin 2.2 2, 4, 6, 7, 8, 9, Rod Insertion Limits 2.5 10, 12, 14, 15, 16 3.1.8 Physics Tests Exceptions Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16 3.2.1 Heat Flux Hot Channel Factor FQ 2.6 2, 4, 6, 7, 8, 9, 10, AFD 2.8 12, 15, 16 OT'T 2.9 Penalty Factors 2.6 3.2.2 Nuclear Enthalpy Rise Hot Channel F'H 2.7 2, 4, 6, 7, 8, 9, Factor Penalty Factors 2.7 10, 12, 15, 16 3.2.3 Axial Flux Difference AFD 2.8 2, 4, 6, 7, 8, 15, 16 3.3.1 Reactor Trip System Instrumentation OT'T 2.9 6, 7, 8, 9, 10, 12 OP'T 2.9 15, 16 3.3.9 Boron Dilution Mitigation System Reactor Makeup Water Flow Rate 2.10 6, 7, 8, 12, 14, 16 3.4.1 RCS Pressure, Temperature and Flow limits for DNB RCS Pressure, Temperature and Flow 2.11 6, 7, 8, 9, 10, 12 3.5.1 Accumulators Max and Min Boron Conc.

2.12 6, 7, 8, 12, 14, 16 3.5.4 Refueling Water Storage Tank Max and Min Boron Conc.

2.13 6, 7, 8, 12, 14, 16 3.7.15 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.14 6, 7, 8, 12, 14, 16 3.9.1 Refueling Operations - Boron Concentration Min Boron Concentration 2.15 6, 7, 8, 12, 14, 16 5.6.5 Core Operating Limits Report (COLR)

Analytical Methods 1.1 None The Selected License Commitments that reference this report are listed below SLC Section Selected Licensing Commitment COLR Parameter COLR Section NRC Approved Methodology (Section 1.1 Number) 16.7-9 Standby Shutdown System Standby Makeup Pump Water Supply 2.16 6, 7, 8, 12, 14, 16 16.9-11 Boration Systems - Borated Water Source - Shutdown Borated Water Volume and Conc. for BAT/RWST 2.17 6, 7, 8, 12, 14, 16 16.9-12 Boration Systems - Borated Water Source - Operating Borated Water Volume and Conc. for BAT/RWST 2.18 6, 7, 8, 12, 14, 16

CNEI-0400-269 Page 5 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 1.1 Analytical Methods 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 Best-Estimate Loss of Coolant Analysis, (W Proprietary).

Revision: Volume 1 (Revision 2) and Volumes 2-5 (Revision 1)

Report Date: March 1998 5.

BAW-10168P-A, B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (B&W Proprietary).

Revision 1 SER Date: January 22, 1991 Revision 2 SER Dates: August 22, 1996 and November 26, 1996.

Revision 3 SER Date: June 15, 1994.

Not Used

CNEI-0400-269 Page 6 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 1.1 Analytical Methods (continued) 6.

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

Revision 5a Report Date: October 2012 7.

DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary).

Revision 0a Report Date: May 2009 8.

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

Revision 4b Report Date: September 2010 9.

DPC-NE-2004P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," (DPC Proprietary).

Revision 2a Report Date: December 2008

10. DPC-NE-2005P-A, "Thermal Hydraulic Statistical Core Design Methodology," (DPC Proprietary).

Revision 4a Report Date: December 2008

11. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3 and GDTACO," (DPC Proprietary).

Revision 2 Report Date: August 2012 Not Used

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

Revision 3a Report Date: September 2011

13. DPC-NE-1004-A, "Nuclear Design Methodology Using CASMO-3/SIMULATE-3P."

Revision 1a Report Date: January 2009 Not Used

CNEI-0400-269 Page 7 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 1.1 Analytical Methods (continued)

14. DPC-NF-2010-A, "Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design."

Revision 2a Report Date: December 2009

15. DPC-NE-2011-PA, "Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," (DPC Proprietary).

Revision 1a Report Date: June 2009

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

(DPC Proprietary).

Revision 1 Report Date: November 12, 2008

17. BAW-10231P-A, "COPERNIC Fuel Rod Design Computer Code" (Framatome ANP Proprietary)

Revision 1 SER Date: January 14, 2004 Not Used

CNEI-0400-269 Page 8 of 31 Revision 1 Catawba 1 Cycle 22 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 NRC approved methodologies specified in Section 1.1.

2.1 Reactor Core Safety Limits (TS 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, TS 3.1.8) 2.2.1 For TS 3.1.1, SDM shall be greater than or equal to 1.3% 'K/K in MODE 2 with Keff < 1.0 and in MODES 3 and 4.

2.2.2 For TS 3.1.1, SDM shall be greater than or equal to 1.0% 'K/K in MODE 5.

2.2.3 For TS 3.1.4, SDM shall be greater than or equal to 1.3% 'K/K in MODE 1 and MODE 2.

2.2.4 For TS 3.1.5, SDM shall be greater than or equal to 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 greater than or equal to 1.3% 'K/K in MODE 1 and MODE 2 with Keff > 1.0.

2.2.6 For TS 3.1.8, SDM shall be greater than or equal to 1.3% 'K/K in MODE 2 during PHYSICS TESTS.

CNEI-0400-269 Page 9 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation 580 590 600 610 620 630 640 650 660 670 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power RCS Tavg (°F)

DO NOT OPERATE IN THIS AREA 2100 psia 2280 psia 2400 psia 1945 psia ACCEPTABLE OPERATION

CNEI-0400-269 Page 10 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.3 Moderator Temperature Coefficient - MTC (TS 3.1.3) 2.3.1 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 'K/K/°F.

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

2.3.3 60 PPM MTC Surveillance Limit is:

Measured 60 PPM ARO, equilibrium RTP MTC shall be less negative than or equal to -4.125E-04 'K/K/°F.

Where:

BOC = Beginning of Cycle (burnup corresponding to most positive MTC)

EOC = End of Cycle ARO = All Rods Out HZP = Hot Zero Thermal Power RTP = Rated Thermal Power PPM = Parts per million (Boron) 2.4 Shutdown Bank Insertion Limit (TS 3.1.5) 2.4.1 Each shutdown bank shall be withdrawn to at least 222 steps. Shutdown banks are withdrawn in sequence and with no overlap.

2.5 Control Bank Insertion Limits (TS 3.1.6) 2.5.1 Control banks shall be within the insertion, sequence, and overlap limits shown in Figure 3. Specific control bank withdrawal and overlap limits as a function of the fully withdrawn position are shown in Table 1.

CNEI-0400-269 Page 11 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.

Refer to the Unit 1 ROD manual for details.

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 0

10 20 30 40 50 60 70 80 90 100 Moderator Temperature Coefficient (1.0E-04 DK/K/oF)

Percent of Rated Thermal Power Unacceptable Operation Acceptable Operation

CNEI-0400-269 Page 12 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power The Rod Insertion Limits (RIL) for Control Bank D (CD), Control Bank C (CC), and Control Bank B (CB) can be calculated by:

Bank CD RIL = 2.3(P) - 69 {30 < P < 100}

Bank CC RIL = 2.3(P) +47

{0 < P < 76.1} for CC RIL = 222 {76.1 < 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: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.

Refer to the Unit 1 ROD manual for details.

0 20 40 60 80 100 120 140 160 180 200 220 0

10 20 30 40 50 60 70 80 90 100 Rod Insertion Position (Steps Withdrawn)

Percent of Rated Thermal Power Fully Withdrawn (Maximum = 231) 231 Control Bank B Control Bank C Control Bank D (0%, 163)

(0%, 47)

(30%, 0)

(100%, 161)

Fully Inserted Fully Withdrawn (Minimum = 222)

CNEI-0400-269 Page 13 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Table 1 Control Bank Withdrawal Steps and Sequence 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

0 222 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 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 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 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

0 226 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 111 0

226 226 116 0 Start 227 227 116 0 Start 226 226 226 Stop 110 227 227 227 Stop 111 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

0 228 Stop 112 0

0 229 Stop 113 0

0 228 116 0 Start 0

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 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

0 230 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 Fully Withdrawn at 222 Steps Fully Withdrawn at 223 Steps Fully Withdrawn at 224 Steps Fully Withdrawn at 225 Steps Fully Withdrawn at 226 Steps Fully Withdrawn at 227 Steps Fully Withdrawn at 230 Steps Fully Withdrawn at 231 Steps Fully Withdrawn at 228 Steps Fully Withdrawn at 229 Steps

CNEI-0400-269 Page 14 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.6 Heat Flux Hot Channel Factor - FQ(X,Y,Z) (TS 3.2.1) 2.6.1 FQ(X,Y,Z) steady-state limits are defined by the following relationships:

F RTP Q

• K(Z)/P for P > 0.5 Error! Bookmark not defined.Error! Bookmark not defined.

F RTP Q

• K(Z)/0.5 for P < 0.5

where, P = (Thermal Power)/(Rated Power)

Note: The 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 limit. The manufacturing tolerance and measurement uncertainty are implicitly included in the FQ surveillance limits as defined for COLR Sections 2.6.5 and 2.6.6.

2.6.2 FRTP Q

= 2.70 x K(BU) 2.6.3 K(Z) is the normalized FQ(X,Y,Z) as a function of core height. K(Z) for Westinghouse RFA fuel is provided in Figure 4.

2.6.4 K(BU) is the normalized FQ(X,Y,Z) as a function of burnup. FRTP Q

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:

2.6.5

[F L

Q(X,Y,Z)]OP =

Y,Z)

• MQ(X,Y,Z)

UMT

• MT

• TILT F

D Q(X, where:

[

L Q

F (X,Y,Z)]OP = Cycle dependent maximum allowable design peaking factor that ensures FQ(X,Y,Z) LOCA limit is not exceeded for operation within the AFD, RIL, and QPTR limits.

L Q

F (X,Y,Z)

OP includes allowances for calculation and measurement uncertainties.

D Q

F (X,Y,Z) =

Design power distribution for FQ.

D Q

F (X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions and in

CNEI-0400-269 Page 15 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 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 that accounts for allowable quadrant power tilt ratio of 1.02. (TILT = 1.035) 2.6.6

[F L

Q(X,Y,Z)]

RPS =

F D

Q(X,Y,Z)

• MC(X,Y,Z)

UMT

• MT

• TILT where:

[F L

Q(X,Y,Z)]RPS = Cycle dependent maximum allowable design peaking factor that ensures FQ(X,Y,Z) Centerline Fuel Melt (CFM) limit is not exceeded for operation within the AFD, RIL, and QPTR limits.

[F L

Q(X,Y,Z)]RPS includes allowances for calculation and measurement uncertainties.

F D

Q(X,Y,Z) =

Design power distributions for FQ. F D

Q(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 operations.

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 operations.

UMT =

Total Peak Measurement Uncertainty. (UMT = 1.05)

MT = Engineering Hot Channel Factor. (MT = 1.03).

CNEI-0400-269 Page 16 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report TILT =

Peaking penalty that accounts for allowable quadrant power tilt ratio of 1.02. (TILT = 1.035) 2.6.7 KSLOPE = 0.0725 where:

KSLOPE = Adjustment to K1 value from OT'T trip setpoint required to compensate for each 1%

M Q

F (X,Y,Z) exceeds L

Q F (X,Y,Z)

RPS.

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.

CNEI-0400-269 Page 17 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for Westinghouse RFA Fuel 0.000 0.200 0.400 0.600 0.800 1.000 1.200 0.0 2.0 4.0 6.0 8.0 10.0 12.0 K(Z)

Core Height (ft)

(0.0, 1.00)

(4.0, 1.00)

(12.0, 0.9259)

(4.0, 0.9259)

Core Height (ft)

K(Z) 0.0 1.000

<4 1.000

>4 0.9259 12.0 0.9259

CNEI-0400-269 Page 18 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Table 2 FQ(X,Y,Z) and F'H(X,Y) Penalty Factors For Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burnup FQ(X,Y,Z)

F'H(X,Y)

(EFPD)

Penalty Factor(%)

Penalty Factor (%)

4 2.00 2.00 12 2.00 2.00 25 2.00 2.00 50 2.41 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 2.00 492 2.00 2.00 500 2.00 2.00 502 2.00 2.00 507 2.00 2.00 517 2.00 2.00 527 2.00 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups.

All cycle burnups outside the range of the table shall use a 2%

penalty factor for both FQ(X,Y,Z) and F'H(X,Y) for compliance with the Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.

CNEI-0400-269 Page 19 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.7 Nuclear Enthalpy Rise Hot Channel Factor - FH(X,Y) (TS 3.2.2)

F'H steady-state limits referred to in Technical Specification 3.2.2 are defined by the following relationship.

2.7.1 LCO L

H Y)]

(X,

[F'

= MARP (X,Y) * ¬<<

1/4>>º 1.0 + 1 RRH * (1.0 - P) where:

LCO L

H Y)]

(X,

[F' 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% measured radial peak, M

H F' (X,Y), exceeds the 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

[

L H

F' (X,Y)]

SURV =

TILT UMR Y)

(X, M

Y)

(X, F

H D

H where:

[

L H

F' (X,Y)]

SURV =

Cycle dependent maximum allowable design peaking factor that ensures F'H(X,Y) limit is not exceeded for operation within the AFD, RIL, and QPTR limits.

L H

F' (X,Y)

SURV includes allowances for calculation and measurement uncertainty.

F D

'H (X,Y) = Design radial power distribution for F'H. F D

'H (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.

CNEI-0400-269 Page 20 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report M'H(X,Y) = Margin remaining in core location X,Y relative to Operational DNB limits in the transient power distribution.

M'H(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 set to 1.0 since a factor of 1.04 is implicitly included in the variable M'H(X,Y).

TILT = Peaking penalty that accounts 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% measured radial peak, M

H F' (X,Y) exceeds its limit. (0 < P < 1.0) 2.7.4 TRH = 0.04 where:

TRH =

Reduction in OT'T K1 setpoint required to compensate for each 1%

measured radial peak, M

H F' (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 Axial Flux Difference (AFD) Limits are provided in Figure 5.

CNEI-0400-269 Page 21 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPS)

RFA Fuel MARPs 100% Full Power Core Height (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 1.8092 1.8553 1.9489 1.9953 1.9741 2.1073 2.0498 2.0090 1.9333 1.8625 1.7780 1.3151 1.2461 1.20 1.8102 1.8540 1.9401 1.9953 1.9741 2.1073 2.0191 1.9775 1.9009 1.8306 1.7852 1.3007 1.2235 2.40 1.8093 1.8525 1.9312 1.9779 1.9741 2.0735 1.9953 1.9519 1.8760 1.8054 1.7320 1.4633 1.4616 3.60 1.8098 1.8514 1.9204 1.9641 1.9741 2.0495 1.9656 1.9258 1.8524 1.7855 1.6996 1.4675 1.3874 4.80 1.8097 1.8514 1.9058 1.9449 1.9741 2.0059 1.9441 1.9233 1.8538 1.7836 1.6714 1.2987 1.2579 6.00 1.8097 1.8514 1.8921 1.9212 1.9455 1.9336 1.8798 1.8625 1.8024 1.7472 1.6705 1.3293 1.2602 7.20 1.8070 1.8438 1.8716 1.8930 1.8872 1.8723 1.8094 1.7866 1.7332 1.6812 1.5982 1.2871 1.2195 8.40 1.8073 1.8319 1.8452 1.8571 1.8156 1.7950 1.7359 1.7089 1.6544 1.6010 1.5127 1.2182 1.1578 9.60 1.8072 1.8102 1.8093 1.7913 1.7375 1.7182 1.6572 1.6347 1.5808 1.5301 1.4444 1.1431 1.0914 10.80 1.7980 1.7868 1.7611 1.7163 1.6538 1.6315 1.5743 1.5573 1.5088 1.4624 1.3832 1.1009 1.0470 11.40 1.7892 1.7652 1.7250 1.6645 1.6057 1.5826 1.5289 1.5098 1.4637 1.4218 1.3458 1.0670 1.0142 Axial Peak

CNEI-0400-269 Page 22 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits NOTE: Compliance with Technical Specification 3.2.1 may require more restrictive AFD limits. Refer to the Unit 1 ROD manual for operational AFD limits.

0 10 20 30 40 50 60 70 80 90 100

-50

-40

-30

-20

-10 0

10 20 30 40 50 Percent of Rated Thermal Power Axial Flux Difference (% Delta I)

(+10, 100)

(-20, 100)

(-36, 50)

(+21, 50)

Acceptable Operation Unacceptable Operation Unacceptable Operation

CNEI-0400-269 Page 23 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.9 Reactor Trip System Instrumentation Setpoints (TS 3.3.1) Table 3.3.1-1 2.9.1 Overtemperature 'T Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T' < 585.1 °F Nominal RCS Operating Pressure P' = 2235 psig Overtemperature 'T reactor trip setpoint ++

K1 = 1.1978 Overtemperature 'T reactor trip heatup setpoint penalty coefficient K2 = 0.03340/oF Overtemperature 'T reactor trip depressurization setpoint penalty coefficient K3 = 0.001601/psi Time constants utilized in the lead-lag compensator for 'T W1 = 8 sec.

W2 = 3 sec.

Time constant utilized in the lag compensator for 'T W3 = 0 sec.

Time constants utilized in the lead-lag compensator for Tavg W4 = 22 sec.

W5 = 4 sec.

Time constant utilized in the measured Tavg lag compensator W6 = 0 sec.

f1('I) "positive" breakpoint

= 19.0 %'I f1('I) "negative" breakpoint

= N/A*

f1('I) "positive" slope

= 1.769 %'T0/ %'I f1('I) "negative" slope

= N/A*

f1('I) negative breakpoints and slopes for OT'T are less restrictive than the OP'T f2('I) negative breakpoint and slope. Therefore, during a transient which challenges the negative imbalance limits, OP'T f2('I) limits will result in a reactor trip before OT'T f1('I) limits are reached. This makes implementation of an OT'T f1('I) negative breakpoint and slope unnecessary.

++ 'T0 is assumed to be renormalized to 100% RTP following the MUR power uprate.

CNEI-0400-269 Page 24 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.9.2 Overpower 'T Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T" < 585.1 ºF Overpower 'T reactor trip setpoint ++

K4 = 1.0864 Overpower 'T reactor trip penalty K5 = 0.02 / °F for increasing Tavg K5 = 0.00 / °F for decreasing Tavg Overpower 'T reactor trip heatup setpoint penalty coefficient K6 = 0.001179/°F for T > T

K6 = 0.0 /°F for T < T

Time constants utilized in the lead-lag compensator for 'T W1 = 8 sec.

W2 = 3 sec.

Time constant utilized in the lag compensator for 'T W3 = 0 sec.

Time constant utilized in the measured Tavg lag compensator W6 = 0 sec.

Time constant utilized in the rate-lag controller for Tavg W7 = 10 sec.

f2('I) "positive" breakpoint

= 35.0 %'I f2('I) "negative" breakpoint

= -35.0 %'I f2('I) "positive" slope

= 7.0 %'T0/ %'I f2('I) "negative" slope

= 7.0 %'T0/ %'I

++ 'T0 is assumed to be renormalized to 100% RTP following the MUR power uprate.

CNEI-0400-269 Page 25 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.10 Boron Dilution Mitigation System - BDMS (TS 3.3.9) 2.10.1 Reactor Makeup Water Pump combined flow rate limits:

Applicable MODE Limit MODE 3 80 gpm MODE 4 or 5 70 gpm 2.11 RCS Pressure, Temperature and Flow DNB Limits (TS 3.4.1)

The 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 1 and 2, and MODE 3 with RCS pressure >1000 psi:

Parameter Applicable Burnup Limit Accumulator minimum boron concentration.

0 - 200 EFPD 2,500 ppm Accumulator minimum boron concentration.

200.1 - 250 EFPD 2,500 ppm Accumulator minimum boron concentration.

250.1 - 300 EFPD 2,413 ppm Accumulator minimum boron concentration.

300.1 - 350 EFPD 2,307 ppm Accumulator minimum boron concentration.

350.1 - 400 EFPD 2,226 ppm Accumulator minimum boron concentration.

400.1 - 450 EFPD 2,149 ppm Accumulator minimum boron concentration.

450.1 - 500 EFPD 2,079 ppm Accumulator minimum boron concentration.

500.1 - 527 EFPD 2,004 ppm Accumulator maximum boron concentration.

0 - 527 EFPD 3,075 ppm

CNEI-0400-269 Page 26 of 31 Revision 1 Catawba 1 Cycle 22 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

CNEI-0400-269 Page 27 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 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.

2,700 ppm RWST maximum boron concentration.

3,075 ppm 2.14 Spent Fuel Pool Boron Concentration (TS 3.7.15) 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.

2,700 ppm 2.15 Refueling Operations - Boron Concentration (TS 3.9.1) 2.15.1 Minimum boron concentration limit for 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 the MODE 6 reactivity requirement of Keff 0.95.

Parameter Limit Minimum boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity.

2,700 ppm

CNEI-0400-269 Page 28 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.16 Standby Shutdown System - (SLC-16.7-9) 2.16.1 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 16.7-9-3.

2,700 ppm 2.17 Boration Systems Borated Water Source - Shutdown (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 MODE 4 with any RCS cold leg temperature < 210°F, and MODES 5 and 6.

Parameter Limit BAT minimum boron concentration 7,000 ppm Volume of 7,000 ppm boric acid solution required to maintain SDM at 68oF 2,000 gallons BAT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-11) 13,086 gallons (14.9%)

NOTE: When cycle burnup is 471 EFPD, Figure 6 may be used to determine the required BAT Minimum Level.

RWST minimum boron concentration 2,700 ppm Volume of 2,700 ppm boric acid solution required to maintain SDM at 68 oF 7,000 gallons RWST Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-11) 48,500 gallons (8.7%)

CNEI-0400-269 Page 29 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report 2.18 Boration Systems Borated Water Source - Operating (SLC 16.9-12) 2.18.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODES 1, 2, and 3 and MODE 4 with all RCS cold leg temperatures > 210°F*.

• NOTE: The SLC 16.9-12 applicability is down to MODE 4 temperatures of

> 210oF. The minimum volumes calculated support cooldown to 200oF to satisfy UFSAR Chapter 9 requirements.

Parameter Limit BAT minimum boron concentration 7,000 ppm Volume of 7,000 ppm boric acid solution required to maintain SDM at 210oF 13,500 gallons BAT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-12) 25,200 gallons (45.8%)

NOTE: When cycle burnup is 471 EFPD, Figure 6 may be used to determine the required BAT Minimum Level.

RWST minimum boron concentration 2,700 ppm Volume of 2,700 ppm boric acid solution required to maintain SDM at 210 oF 57,107 gallons RWST Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-12) 98,607 gallons (22.0%)

CNEI-0400-269 Page 30 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is 471 EFPD)

This figure includes additional volumes listed in SLC 16.9-11 and 16.9-12 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 0

200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 BAT Level (% Level)

Primary Coolant Boron Concentration (ppmb)

Acceptable Operation Unacceptable Operation RCS Boron Concentration BAT Level (ppm)

(%level) 0 < 300 43.0 300 < 500 40.0 500 < 700 37.0 700 < 1000 30.0 1000 < 1300 14.9 1300 < 2700 9.8

> 2700 9.8

CNEI-0400-269 Page 31 of 31 Revision 1 Catawba 1 Cycle 22 Core Operating Limits Report Appendix A Power Distribution Monitoring Factors Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance. This data was generated in the Catawba 1 Cycle 22 Maneuvering Analysis calculation file, CNC-1553.05-00-0608. 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 Catawba Electrical and Reactor Systems Engineering Section controls this information via computer files 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.