ML19009A296

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Cycle 25 Core Operating Limits Report
ML19009A296
Person / Time
Site: Catawba Duke Energy icon.png
Issue date: 10/31/2018
From:
Duke Energy Carolinas
To:
Office of Nuclear Reactor Regulation
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ML19009A293 List:
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RA-18-0269 CNEI-0400-341, Rev 0
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CNEI-0400-3 41 Page 1 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Revision 0 October 201 8

Reference:

CNC

-1553.05-00-06 73, Rev. 0 Duke Energy Carolinas, LLC QA CONDITION 1 The information presented in this report has been prepared and issued in accordance with Catawba Technical Specification 5.6.5.

CNEI-0400-341 Page 2 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Implementa tion Instructions for Revision 0 Revision Description and CR Tracking Revision 0 of the Catawba Unit 1 Cycle 2 5 Core Operating Limits Report (COLR) contains limits specific to the reload core.

There is no CR associated with this revision.

The 50.59 AR is 02236225. Implementation Schedule The Catawba Unit 1 Cycle 2 5 COLR requires the reload 50.59 be approved prior to implementation and fuel loading.

Revision 0 may become effective any time during No MODE between cycles 2 4 and 2 5 but must become effective prior to entering MODE 6 which starts cycle 2

5. The Catawba Unit 1 Cycle 2 5 COLR will cease to be effective during No MODE between cycle 2 5 and 2 6. Data files to be Implemented No data files are transmitted as part of this document.

CNEI-0400-341 Page 3 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected COLR 0 October 20 18 1-3 1, Appendix A

0

  • Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance and is not uploaded as part of the EI body.

However, Appendix A is uploaded into the document management system, for ease of transmittal to the NRC.

CNEI-0400-341 Page 4 Revision 0 Catawba 1 Cycle 2 5 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, 19, 20 3.1.1 Shutdown Margin Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19, 20 3.1.3 Moderator Temperature Coefficient MTC 2.3 6, 7, 8, 14, 16, 18 3.1.4 Rod Group Alignment Limits Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19, 20 3.1.5 Shutdown Bank Insertion Limit Shutdown Margin 2.2 2, 4, 6, 7, 8, 9, 10 Rod Insertion Limits 2.4 1 2, 14, 15, 16, 19, 20 3.1.6 Control Bank Insertion Limit Shutdown Margin 2.2 2, 4, 6, 7, 8, 9, 10 Rod Insertion Limits 2.5 1 2, 14, 15, 16, 19, 20 3.1.8 Physics Tests Exceptions Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19, 20 3.2.1 Heat Flux Hot Channel Factor F Q 2.6 2, 4, 6, 7, 8, 9, 10, AFD 2.8 12, 15, 16, 19, 20 OTT 2.9 Penalty Factors 2.6 3.2.2 Nuclear Enthalpy Rise Hot Channel FH 2.7 2, 4, 6, 7, 8, 9, 10 Factor Penalty Factors 2.7 12, 15, 16, 19, 20 3.2.3 Axial Flux Difference AFD 2.8 2, 4, 6, 7, 8, 15, 16

3.3.1 Reactor

Trip System Instrumentation OTT 2.9 6, 7, 8, 9, 10, 12 OPT 2.9 15, 16, 19, 20 3.3.9 Boron Dilution Mitigation System Reactor Makeup Water Flow Rate 2.10 6, 7, 8, 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 , 19, 20 3.5.1 Accumulators Max and Min Boron Conc.

2.12 6, 7, 8, 14, 16

3.5.4 Refueling

Water Storage Tank Max and Min Boron Conc.

2.13 6, 7, 8, 14, 16 3.7.15 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.14 6, 7, 8, 14, 16

3.9.1 Refueling

Operations

- Boron Concentration Min Boron Concentration 2.15 6, 7, 8, 14, 16 5.6.5 Core Operating Limits Report (COLR)

Analytical Methods 1.1 None The Selected Licens ee Commitments that reference this report are listed below SLC Section Selected Licens ee 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, 14, 16 16.9-11 Boration Systems

- Borated Water Source - Shutdown Borated Water Volume and Conc. for BAT/RWST 2.17 6, 7, 8, 14, 16 16.9-12 Boration Systems

- Borated Water Source - Operating Borated Water Volume and Conc. for BAT/RWST 2.18 6, 7, 8, 14, 16

CNEI-0400-341 Page 5 Revision 0 Catawba 1 Cycle 2 5 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 101) Revision 1 Report Date: July 1997 3. WCAP-1026 6-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-0400-341 Page 6 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report

1.1 Analytical

Methods (continued)

6. DPC-NE-3000-PA, "Thermal

-Hydraulic Transient Analysis Methodology," (DPC Proprietary).

Revision 5 a Report Date: October 2012

7. DPC-NE-300 1-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary).

Revision 1 Report Date:

March 20 15 8. DPC-NE-3002-A, " UFSAR Chapter 15 System Transient Analysis Methodology".

Revision 4 b 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 20 16 11. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3

," (DPC Proprietary).

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

Revision 3 c Report Date: March 201 7 13. DPC-NE-1004-A, "Nuclear Design Methodology Using CASMO

-3/SIMULATE

-3P." Revision 1 a Report Date: January 2009 Not Used CNEI-0400-341 Page 7 Revision 0 Catawba 1 Cycle 2 5 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 2 a 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 1 a 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, 200 8 17. BAW-10231P-A, "COPERNIC Fuel Rod Design Computer Code" (Framatome ANP Proprietary)

Revision 1 SER Date: January 14, 2004 Not Used 18. DPC-NE-1007-PA, "Conditional Exemption of the EOC MTC Measurement Methodology", (DPC and W Proprietary)

Revision 0 Report Date: April 2015

19. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," (W Proprietary).

Revision 0 Report Date: April 1995

20. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1

-A, "Optimized ZIRLO

Ž," (W Proprietary).

Revision 0 Report Date: July 2006

CNEI-0400-341 Page 8 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report

2.0 Operating

Limits Cycl e-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-0400-341 Page 9 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation

CNEI-0400-341 Page 10 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report

2.3 Moderator

Temperature Coefficient

- MTC (TS 3.1.3)

2.3.1 Moderator

Temperature Coefficient (MTC) Limits ar e: 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

-4.3 E-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 -3.65 E-04 K/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:

Measured 60 PPM ARO, equilibrium RTP MTC shall be less negative than or equal to -4.125 E-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 s (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-0400-341 Page 1 1 Revision 0 Catawba 1 Cycle 2 5 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 100Moderator Temperature Coefficient(1.0E-04 K/K/oF)Percent of Rated Thermal PowerUnacceptable OperationAcceptable Operation CNEI-0400-341 Page 12 Revision 0 Catawba 1 Cycle 2 5 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 C C RIL = 2.3(P) +47 {0 P 76.1} for CC RIL = 222 {76.1 < P 100} Bank C B RIL = 2.3(P) +163 {0 P 25.7} for CB RIL = 222 {25.7 < P 100} where P = %Rated Thermal Power NOTE S: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits. Refer to the Unit 1 ROD manual for details.

020406080100120140160180200220 0102030405060708090100Rod Insertion Position (Steps Withdrawn)Percent of Rated Thermal PowerFully Withdrawn(Maximum = 231)231Control Bank BControl Bank CControl Bank D(0%, 163)(0%, 47)(30%, 0)(100%, 161)Fully InsertedFully Withdrawn(Minimum = 222)

CNEI-0400-341 Page 13 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Table 1 Control Bank Withdrawal Steps and Sequence ControlControlControlControlControlControlControlControlBank ABank BBank CBank DBank ABank BBank CBank D0 Start 0 0 00 Start 0 0 0 1160 Start 0 0 1160 Start 0 0222 Stop 106 0 0223 Stop 107 0 0 222 1160 Start 0 223 1160 Start 0 222222 Stop 106 0 223223 Stop 107 0 222 222 1160 Start 223 223 1160 Start 222 222222 Stop 106 223 223223 Stop 107ControlControlControlControlControlControlControlControlBank ABank BBank CBank DBank ABank BBank CBank D0 Start 0 0 00 Start 0 0 0 1160 Start 0 0 1160 Start 0 0224 Stop 108 0 0225 Stop 109 0 0 224 1160 Start 0 225 1160 Start 0 224224 Stop 108 0 225225 Stop 109 0 224 224 1160 Start 225 225 1160 Start 224 224224 Stop 108 225 225225 Stop 109ControlControlControlControlControlControlControlControlBank ABank BBank CBank DBank ABank BBank CBank D0 Start 0 0 00 Start 0 0 0 1160 Start 0 0 1160 Start 0 0226 Stop 110 0 0227 Stop 111 0 0 226 1160 Start 0 227 1160 Start 0 226226 Stop 110 0 227227 Stop 111 0 226 226 1160 Start 227 227 1160 Start 226 226226 Stop 110 227 227227 Stop 111ControlControlControlControlControlControlControlControlBank ABank BBank CBank DBank ABank BBank CBank D0 Start 0 0 00 Start 0 0 0 1160 Start 0 0 1160 Start 0 0228 Stop 112 0 0229 Stop 113 0 0 228 1160 Start 0 229 1160 Start 0 228228 Stop 112 0 229229 Stop 113 0 228 228 1160 Start 229 229 1160 Start 228 228228 Stop 112 229 229229 Stop 113ControlControlControlControlControlControlControlControlBank ABank BBank CBank DBank ABank BBank CBank D0 Start 0 0 00 Start 0 0 0 1160 Start 0 0 1160 Start 0 0230 Stop 114 0 0231 Stop 115 0 0 230 1160 Start 0 231 1160 Start 0 230230 Stop 114 0 231231 Stop 115 0 230 230 1160 Start 231 231 1160 Start 230 230230 Stop 114 231 231231 Stop 115Fully Withdrawn at 230 StepsFully Withdrawn at 231 StepsFully Withdrawn at 228 StepsFully Withdrawn at 229 StepsFully Withdrawn at 222 StepsFully Withdrawn at 223 StepsFully Withdrawn at 224 StepsFully Withdrawn at 225 StepsFully Withdrawn at 226 StepsFully Withdrawn at 227 Steps

CNEI-0400-341 Page 14 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report 2.6 Heat Flux Hot Channel Factor

- F Q(X,Y,Z) (TS 3.2.1) 2.6.1 F Q(X,Y,Z) steady-state limits are defined by the following relationships:

FRTP Q*K(Z)/P for P > 0.5 FRTP Q*K(Z)/0.5 for P 0.5 where, P = (Thermal Power)/(Rated Power)

Note: The measured F Q(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 F Q 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 F Q(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 F Q(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 t o 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)

  • M Q (X , Y , Z)UMT
  • TILT F D Q (X , where: [L Q F(X,Y,Z)]OP = Cycle dependent maximum allowable design peaking factor that ensures F Q(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 F Q. D Q F(X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial startup operation.

CNEI-0400-341 Page 15 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report M Q(X,Y,Z) = Margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution. M 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 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)

  • M C (X , Y , Z)UMT
  • TILT where: [F L Q(X,Y,Z)]RPS = Cycle dependent maximum allowable design peaking factor that ensures F Q(X,Y,Z) Centerline Fuel Melt (CFM) limit is not exceeded for operation within the AFD, RIL, and QP TR limits. [F L Q(X,Y,Z)]RPS includes allowances for calculation and measurement uncertainties.

F D Q(X,Y,Z) = Defined in Section 2.6.5.

M C(X,Y,Z) = Margin remaining to the CFM limit in core location X,Y,Z from the transient power distribution. M C(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 = Defined in Section 2.6.5.

MT = Defined in Section 2.6.5.

TILT = Defined in Section 2.6.5.

CNEI-0400-341 Page 16 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report

2.6.7 KSLOPE

= 0.0725 where: KSLOPE = Adjustment to K 1 value from OTT 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 F Q(X,Y,Z) Penalty Factors for Technical Specification Surveillances 3.2.1.2 and 3.2.1.3 are provided in Table 2.

CNEI-0400-341 Page 17 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Figure 4 K(Z), Normalized F Q(X,Y,Z) as a Function of Core Height for Westinghouse RFA Fuel 0.0000.2000.4000.6000.8001.0001.2000.02.04.06.08.010.012.0K(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)0.01.000<41.000>40.925912.00.9259 CNEI-0400-341 Page 18 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Table 2 F Q(X,Y,Z) and FH(X,Y) Penalty Factors For Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burnup F Q(X,Y,Z) FH(X,Y) (EFPD) Penalty Factor(%)

Penalty Factor (%)

4 2.00 2.00 12 2.0 0 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 45 0 2.00 2.00 4 55 2.00 2.00 475 2.00 2.00 485 2.00 2.00 495 2.00 2.00 505 2.00 2.00 515 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 F Q(X,Y,Z) and FH(X,Y) for compliance with the Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.

CNEI-0400-341 Page 19 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report

2.7 Nuclear

Enthalpy Rise Hot Channel Factor

- FH(X,Y) (TS 3.2.2)

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

2.7.1 LCO L HY)](X,[F= MARP (X,Y)

  • 1.0 + 1 RRH * (1.0 - P) where: LCO L HY)](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 PowerRated Thermal Power RRH = Thermal Power reduction required to compensate for each 1%

the measured radial peak, M H F(X,Y), exceeds the limit.

RRH also is used to scale the MARP limits as a function of power per the LCO L H Y)(X, F equation.

(RRH = 3.34 (0.0 < P

< 1.0) The following parameters are required for core monitoring per the surveillance requirements of Technical Specification 3.2.2.

2.7.2 [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 the FH(X,Y) limit is not exceeded for operation within the LCO limits. L H F(X,Y)SURV includes allowances for calculation and measurement uncertainty.

F DH (X,Y) = Design radial power distribution for FH. F DH (X,Y) is provided in Appendix Table A-3 for normal operation and in

CNEI-0400-341 Page 20 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Appendix Table A-6 for power escalation testing during initial startup operation.

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

MH(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 MH(X,Y). TILT = Defined in Section 2.6.5.

2.7.3 RRH is defined in Section 2.7.1.

2.7.4 TRH = 0.04 where: TRH = Reduction in OTT K 1 setpoint required to compensate for each 1% that the measured radial peak, M H F(X,Y) exceeds its limit.

2.7.5 FH(X,Y) Penalty Factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2. 2.8 Axial Flux Differenc e - AFD (TS 3.2.3)

2.8.1 Axial

Flux Difference (AFD) Limits are provided in Figure 5.

CNEI-0400-341 Page 21 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARP s) RFA MARPs Core Height (ft) Axial Peak 1.05 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 2.1 3 3.25 0.12 1.8092 1.8553 1.9248 1.9146 1.9179 2.0621 2.0498 2.0090 1.9333 1.8625 1.7780 1.3151 1.2461 1.20 1.8102 1.8540 1.9248 1.9146 1.9179 2.1073 2.0191 1.9775 1.9009 1.8306 1.7852 1.3007 1.2235 2.40 1.8093 1.8525 1.931 2 1.9146 1.9179 2.0735 1.9953 1.9519 1.8760 1.8054 1.7320 1.4633 1.4616 3.60 1.8098 1.8514 1.9204 1.9146 1.9179 2.0495 1.9656 1.9258 1.8524 1.7855 1.6996 1.4675 1.3874 4.80 1.8097 1.8514 1.9058 1.9146 1.9179 2.0059 1.9441 1.9233 1.8538 1.7836 1.6714 1.29 87 1.2579 6.00 1.8097 1.8514 1.8921 1.9212 1.9179 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 1.73 59 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 CNEI-0400-341 Page 22 Revision 0 Catawba 1 Cycle 2 5 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 90100-50-40-30-20-10 0 10 20 30 40 50Percent of Rated Thermal PowerAxial Flux Difference (% Delta I)(+10, 100)

(-20, 100)(-36, 50)(+21, 50)Acceptable OperationUnacceptable OperationUnacceptable Operation CNEI-0400-341 Page 23 Revision 0 Catawba 1 Cycle 2 5 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 K 1 = 1.1978 Overtemperature T reactor trip heatup setpoint penalty coefficient K 2 = 0.03340/

°F Overtemperature T reactor trip depressurization setpoint penalty coefficient K 3 = 0.001601/psi Time constants utilized in the lead

-lag compensator for T 1 = 8 sec. 2 = 3 sec. Time constant utilized in the lag compensator for T 3 = 0 sec. Time constants utilized in the lead

-lag compensator for Tavg 4 = 22 sec. 5 = 4 sec. Time constant utilized in the measured Tavg lag compensator 6 = 0 sec. f 1 (I) "positive" breakpoint

= 19.0 %I f 1 (I) "negative" breakpoint

= N/A* f 1 (I) "positive" slope

= 1.769 %T 0/ %I f 1 (I) "negative" slope = N/A*

  • f 1 (I) negative breakpoints and slopes for OTT are less restrictive than the OPT f 2 (I) negative breakpoint and slope. Therefore, during a transient which challenges the negative imbalance limits

, OPT f 2 (I) limits will result in a reactor trip before OTT f 1 (I) limits are reached. This makes implementation of an OTT f 1 (I) negative breakpoint and slope unnecessary.

CNEI-0400-341 Page 24 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report

2.9.2 Overpower

T Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T" 585.1°F Overpower T reactor trip setpoint K 4 = 1.0864 Overpower T reactor trip penalty K 5 = 0.02 / °F for increasing Tavg K 5 = 0.00 / °F for decreasing Tavg Overpower T reactor trip heatup setpoint penalty coefficient K 6 = 0.001179/

°F for T > T

K 6 = 0.0 /°F for T T Time constants utilized in the lead

-lag compensator for T 1 = 8 sec. 2 = 3 sec. Time constant utilized in the lag compensator for T 3 = 0 sec. Time constant utilized in the measured Tavg lag compensator 6 = 0 sec. Time constant utilized in the rate

-lag controller for Tavg 7 = 10 sec. f 2 (I) "positive" breakpoint

= 35.0 %I f 2 (I) "negative" breakpoint

= -35.0 %I f 2 (I) "positive" slope

= 7.0 %T 0/ %I f 2 (I) "negative" slope

= 7.0 %T 0/ %I CNEI-0400-341 Page 25 Revision 0 Catawba 1 Cycle 2 5 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. 2.1 2 Accumulators (TS 3.5.1) 2.1 2.1 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, 392 ppm Accumulator minimum boron concentration.

300.1 - 350 EFPD 2, 28 9 ppm Accumulator minimum boron concentration.

350.1 - 400 EFPD 2, 203 ppm Accumulator minimum boron concentration.

400.1 - 450 EFPD 2, 128 ppm Accumulator minimum boron concentration.

450.1 - 4 75 EFPD 2, 052 ppm Accumulator minimum boron concentration

. 4 75.1 - 5 05 EFPD 2,0 14 ppm Accumulator minimum boron concentration.

5 05.1 - 5 15 EFPD 1,9 64 ppm Accumulator maximum boron concentration.

0 - 515 EFPD 3,075 ppm CNEI-0400-341 Page 26 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters PARAME TER INDICATION No. Operable CHANNELS LIMITS 1. Indicated RCS Average Temperature meter 4 meter 3 computer 4 computer 3 2. Indicated Pressurizer Pressure meter 4 09.8 psi g meter 3 12.1 p si g computer 4 05.8 psi g computer 3 07.5 psi g 3. RCS Total Flow Rate 8,000 gpm CNEI-0400-341 Page 27 Revision 0 Catawba 1 Cycle 2 5 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 2.1 4 Spent Fuel Pool Boron Concentration (TS 3.7.15) 2.1 4.1 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 2.1 5 Refueling Operations

- Boron Concentration (TS 3.9.1) 2.1 5.1 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 remain s 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-0400-341 Page 28 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report 2.1 6 Standby Shutdown System

- (SLC-16.7-9) 2.1 6.1 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 2.1 7 Boration System s Borated Water Source

- Shutdown (SLC 16.9

-11) 2.1 7.1 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 2 1 0°F, and MODES 5 and 6. Parameter Limit NOTE: When cycle burnup is 455 EFPD, Figure 6 may be used to determine the required BAT Minimum Level

. BAT minimum boron concentration 7,000 ppm Volume of 7,000 ppm boric acid solution required to maintain SDM at 68 o F 2 ,000 gallons B AT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9

-11) 13,086 gallons (14.9%) RWST minimum boron concentration 2,700 ppm Volume of 2,700 ppm boric acid solution required to maintain SDM at 68 o F 7,000 gallons RWST Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9

-11) 48,500 gallons (8.7%)

CNEI-0400-341 Page 29 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report 2.1 8 Boration Systems Borated Water Source

- Operating (SLC 16.9

-12) 2.1 8.1 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 > 2 1 0°F*.

  • NOTE: The SLC 16.9

-12 applicability is down to MODE 4 temperatures of > 210 oF. The minimum volumes calculated support cooldown to 200 oF to satisfy UFSAR Chapter 9 requirements.

Parameter Limit NOTE: When cycle burnup is 455 EFPD, Figure 6 may be used to determine the required BAT Minimum Level

. B AT minimum boron concentration 7,000 ppm Volume of 7,000 ppm boric acid solution required to maintain SDM at 2 1 0 o F 13,500 gallons B AT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9

-12) 25,200 gallons (45.8%) RWST minimum boron concentration 2,700 ppm Volume of 2,700 ppm boric acid solution required to maintain SDM at 2 1 0 o F 57,107 gallons RWST Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9

-12) 98,607 gallons (22.0%)

CNEI-0400-341 Page 30 Revision 0 Catawba 1 Cycle 2 5 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is 455 EFPD) This figure includes additional volumes listed in SLC 16.9

-11 and 16.9

-12 0.05.010.015.020.025.030.035.040.045.050.0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600BAT Level (% Level)Primary Coolant Boron Concentration (ppmb)Acceptable OperationUnacceptable OperationRCS BoronConcentrationBAT Level(ppm)(%level)0 < 30043.0300 < 50040.0500 < 70037.0700 < 1000 30.01000 < 130014.91300 < 27009.8> 2700 9.8 CNEI-0400-341 Page 31 Revision 0 Catawba 1 Cycle 2 5 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 25 Maneuvering Analysis calculation file, CNC

-1553.05-00-0669. 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 if there is a need to access this information.

Appendix A is available to be transmitted to the NRC.

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