Text

Core Operating Limits Report (COLR) for Cycle 21 Reload Core - Revision 2
	ML16043A050
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Site:	Catawba
Issue date:	02/09/2016
From:	Henderson K; Duke Energy Carolinas
To:	; Office of Nuclear Reactor Regulation
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ML16043A028	List: CNS-16-012, Transmittal of Core Operating Limits Report (COLR) for Cycle 21 Reload Core - Revision 2; ML16043A050;
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	CNS-16-012 CNC-1553.05-00-0621, Rev. 2
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Attachment Catawba Unit 2 Cycle 21 COLR - Revision 2 (paper COLR version)

CNEI-0400-28 1 Page 1 Revision 2 Catawba Unit 2 Cycle 21 Core Operating Limits Report Revision 2 January 2016 Calculation Number: CNC- 1553.05-00-0621 Duke Energy Date J. S. Young via fusion Prepared By:

Checked By: C. L. Klein via fusion Checked By: J. A. Reed via fusion (Sections 1.1, 2.1i and 2.9 -2.18)

Approved By: M. A. Biota via fusion Signatures and dates captured electronically via Fusion QA Condition 1 The information presented in this report has been prepared and issued in accordance with Catawba Technical Specification 5.6.5.

CNEI-0400-28 1 Page 2 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Implementation Instructions for Revision 2 Revision Description and CR Tracking Revision 2 of the Catawba Unit 2 Cycle 21 COLR contains limits specific to the reload core and is revised to include updated pressurizer pressure limits. The power distribution monitoring factors from Appendix A of Revision 0 remain valid and are not transmitted as part of Revision 2.

There is no CR associated with this revision.

Implementation Schedule Revision 2 may become effective upon receipt. The Catawba Unit 2 Cycle 21 COLR will cease to be effective during No MODE between cycle 21 and 22.

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

CNEI-0400-28 1 Page 3 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected COLR 0 February 2015 1-31, Appendix A* C2C21 COLR, Rev. 0 1 July 2015 1-31 C2C21 COLR, Rev. 1 2 January 2016 1-31 C2C21 COLR, Rev. 2

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-0400-28 1 Page 4 Revision 2 Catawba 2 Cycle 21 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 TCOLR NRC Approved Section Technical Specifications COLR Parameter Section Methodology (Section 2.1.1 Reactor Core Safety Limits 1

________________________________ ________________________1.1 RCS Temperature and Pressure 1 2.1 Number) 6, 7, 8, 9, 10, 12, 15,

___Safety Limits 416 3.1.1 Shutdown Margin ___Shutdown Margin 12.2 6, 7, 8, 12, 14, 15, 16 3.1.3 Moderator Temperature Coefficient MTC ___ 2.3 6, 7, 8,12, 14, 16, 18 3.1.4 Rod Grou~p* 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 J 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 OTAT 2.9 Penalty__ Factors 2.6 3.2.2 Nuclear Enthalpy Rise Hot Channel FAHl 2.7 2, 4, 6, 7, 8, 9, Factor Penalty Factors j 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 OTAT 2.9 6, 7, 8, 9, 10, 12,

___ _______________ OPAT 2.9 15, 16 3.3.9 Boron Dilution Mitigation S~ystem Reactor Makeup Water Flow Rate 2.10 6, 7, 8, 12, 14, 16 3.4.1 RCS Pressure, Temperature and Flow RCS Pressure, Temperature and Flow I2.11 6, 7, 8, 9, 10, 12 limits for DNB_____ ______

3.5.1 Accumulators Max and Min Boron Cone. 12.12 6, 7, 8, 12, 14, 16 3.5.4 Refueling_ Water Storage Tank Max and Min Boron Cone. j2.13 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 Min Boron Concentration 2.15 6, 7, 8, 12, 14, 16 Concentration _______

5.6.5 Core Operating Limits Report -Analytical Methods 1.1 None

___-(COLR) _______________ _______

The Selected License Commitments that reference this report are listed below SLC COLR J NRC Approved Section Selected Licensing Commitment 16.7-9 Stndy Shutdown System I

COLR Parameter Standb~y MkupPm Water SuI~py-Section 2.16 Methodology (Section I1.____Nu1.bNuber 6, 7, 8, 12, 14, 16 16.9-11 Boration Systems - Borated Water Borated Water Volume and Cone. for 2.17 6, 7, 8, 12, 14, 16

_____Source - Shutdown BATiRWST_______________

16.9-12 Boration Systems - Borated Water Borated Water Volume and Cone. for 2.18 -6, 7, 8, 12, 14, 16

_____Source - Operating BAT/RWST _____ __________

CNEI-0400-28 1 Page 5 Revision 2 Catawba 2 Cycle 21 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 for C2C21

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 for C2C21

4. WCAP-12945-P-A, Volume I and Volumes 2-5, "Code Qualification Document for Best-Estimate Loss of Coolant Analysis," (W Proprietary).

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

Report Date: March 1998

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

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

Revision 3 SER Date: June 15, 1994.

Not Used for C2C21

CNEI-0400-28 1 Page 6 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 1.1 Analytical Methods (continued)

6. DPC-NE-3 000-PA, "Thermal-Hydraulic Transient Analysis Methodology," (DPC Proprietary).

Revision 5a Report Date: October 2012

7. DPC-NE-3 001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary).

Revision 0a Report Date: May 2009 Note: The WLOP Correlation is used for the HZP Steam Line Break DNBR Analysis as approved by the following SER:

Letter from G. Edward Miller (NRC) to Mr. K. Henderson (Duke Energy), "Catawba Nuclear Station, Units 1 and 2 and McGuire Nuclear Station Units 1 and 2 - Issuance of Amendments RE: DPC-NE-3001-P, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology (TAC Nos. MF31 19, MF3 120, MF3 121, and MF3 122), March 25, 2015.

8. DPC-NE-3 002-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-0 1," (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-2008-PA, "Fuel Mechanical Reload Analysis Methodology Using TACO3 and GDTACO," (DPC Proprietary).

Revision 2 Report Date: August 2012 Not Used for C2C21

CNEI-0400-28 1 Page 7 Revision 2 Catawha 2 Cycle 21 Core Operating Limits Report 1.1 Analytical Methods (continued)

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

Revision 3a Report Date: September 2011

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

Revision 1la Report Date: January 2009 Not Used for C2C21

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-20 11-PA, "Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," (DPC Proprietary).

Revision la Report Date: June 2009

16. DPC-NE-1005-PA, "Duke Power Nuclear Design Methodology Using CASMO-4 /

SIMULATE-3 MOX", (DPC Proprietary).

Revision 1 Report Date: November 12, 2008

17. BAW-1023 1P-A, "COPERNIC Fuel Rod Design Computer Code" (Framatome ANP Proprietary)

Revision 1 SER Date: January 14, 2004 Not Used for C2C21

18. DPC-NE- 1007-PA, "Conditional Exemption of the EOC MTC Measurement Methodology",

(DPC and W Proprietary)

Revision 0 Report Date: April 2015

CNEI-0400-28 1 Page 8 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 2.0 Operating Limits Cycle-specific parameter limits for 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)

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% AK/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% AK/K in MODE 5.

2.2.3 For TS 3.1.4, SDM shall be greater than or equal to 1.3% AK/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% AK/K in MODE 1 and MODE 2 with any control bank not fully inserted.

2.2.5 For TS 3.1.6, SDM shall be greater than or equal to 1.3% AK/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% AK/K in MODE 2 during PHYSICS TESTS.

CNEI-0400-281I Page 9 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation 670 660 DO NOT OPERATE IN THIS AREA 650 6400 590 ACCEPTABLE OPERATION 580!!

0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power

CNEI-0400-281I Page 10 Revision 2 Catawba 2 Cycle 21 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 AK/K/°F.

EOC, ARO, RTP MTC shall be less negative than the -4.3E-04 AK/K/0 F lower MTC limit.

2.3.2 300 ppm MTC Surveillance Limit is:

Measured 300 ppm ARO, equilibrium RTP MTC shall be less negative than or equal to -3.65E-04 AK/K/°F.

2.3.3 The Revised Predicted near-EOC 300ppm 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 and 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 -4.125E-04 AK/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-0400-281I Page 11 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 0.9 0.8 0.7 0.6 o

0.5 0.4 0.3 0.2 0.1 0.0 0 10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.

Refer to the Unit 2 ROD manual for details.

CNEI-0400-28 1 Page 12 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power Fully Withdrawn (Maximum =231) 231 - , ,

22 0 - - - - - - - - - - - - - - - - - - - - ---- - - - - - - - - - -

200 / ."/

j ~Fully Withdrawn

,* 180 // *Control Bank B (Mnmu- 22

160 (0%, 163)*

10 /

8 J o BBank Controll 60 j 60 20 0

0 10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power The Rod Insertion Limits (RIL) for Control Bank D (CD), Control Bank C (CC), and Control Bank B (CB) can be calculated by:

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

Bank CCRIL = 2.3(P) +147 {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 =% ofRated Thermal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.

Refer to the Unit 2 ROD manual for details.

CNEI-0400-28 1 Page 13 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Table 1 Control Bank Withdrawal Steps and Sequence Fully Withdrawn at 222 Steps Fully Withdrawn at 223 Steps Control Control Control Control Control Control Control Control BankA Bank B Bank C BankD Bank A Bank B Bank C Bank D 0OStart 0 0 0 0OStart 0 0 0 116 0OStart 0 0 116 0OStart 0 0 222 Stop 106 0 0 223 Stop 107 0 0 222 116 0OStart 0 223 116 0 Start 0 222 222 Stop 106 0 223 223 Stop 107 0 222 222 116 0OStart 223 223 11I6 0OStart 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 BankA BankB BankC BankD Bank A Bank B Bank C Bank D 0 Start 0 0 0 0OStart 0 0 0 116 0OStart 0 0 116 0OStart 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 0OStart 225 225 116 0 Start 224 224 224 Stop 108 225 225 225 Stop 109 Fully Withdrawn at 226 Steps Fully Withdrawn at 227 Steps Control Control Control Control Control Control Control Control BankA BankB BankC BankD Bank A Bank B Blank C Bank D 0OStart 0 0 0 0OStart 0 0 0 116 0OStart 0 0 116 0OStart 0 0 226 Stop 11I0 0 0 227 Stop 111 0 0 226 116 0 Start 0 227 116 0 Start 0 22 6 226 Stop 110 0 227 227 Stop 111 0 226 226 l116 0OStart 227 227 116 0OStart 226 226 226 Stop 11I0 227 227 227 Stop 111 Fully Withdrawn at 228 Steps Fully Withdrawn at 229 Steps Control Control Control Control Control Control Control Control BankA BankB BankC BankD Bank A Bank B Bank C Bank D 0OStart 0 0 0 0OStart 0 0 0 116 0OStart 0 0 116 0OStart 0 0 228 Stop 112 0 0 229 Stop 113 0 0 228 116 0OStart 0 229 116 0OStart 0 228 228 Stop 112 0 229 229 Stop 113 0 228 228 116 0OStart 229 229 116 0OStart 228 228 228 Stop 112 229 229 229 Stop 113 Folly Withdrawn at 230 Steps Folly Withdrawn at 231 Steps Control Control Control Control Control Control Control Control BankA BankB BankC BanklD Bank A BankfB Bank C Bank D 0OStart 0 0 0 0OStart 0 0 0 116 0OStart 0 0 116 0OStart 0 0 230OStop 114 0 0 231 Stop 115 0 0 230 116 0 Start 0 231 116 0OStart 0 230 230 Stop 114 0 231 231 Stop 115 0 230 230 116 0OStart 231 231 116 0OStart 230 230 230OStop 114 231 231 231 Stop 115

CNEI-0400-28 1 Page 14 Revision 2 Catawba 2 Cycle 21 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 RrP *K(Z)/P for P > 0.5 FO~rP *K(Z)/0.5 for P _<0.5 where,

- Thermal Power Rated Thermal Power Note: Measured FQ(X,Y,Z) shall be increased by 3% to account for manufacturing tolerances and 5% to account for measurement uncertainty when comparing against the LCO limits. The manufacturing tolerance and measurement uncertainty are implicitly included in the FQ surveillance limits as defined in COLR Sections 2.6.5 and 2.6.6.

2.6.2 F oRTP= 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. F,RTP with the QO K(BU) penalty for Westinghouse RFA fuel is analytically confirmed in cycle-specific reload calculation. K(BU) is set to 1.0 at all burnups.

CNEI-0400-281I Page 15 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report The following parameters are required for core monitoring per the Surveillance Requirements of Technical Specification 3.2.1:

t xY,'Ip 0 )rL FQ(X,Y,Z)

MQ(X,Y,Z) 2.65

[Q(XYZ] = UMT

MT

TILT where:

[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 AFD, RIL, and QPTR limits.

F* (X,Y,Z)°" includes allowances for calculation and measurement uncertainties.

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

MQ(X,Y,Z) = Margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution. MQ(X,Y,Z) is provided in Appendix Table A-i for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial startup operation.

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)

F*(X,Y,Z)

Mc(X,Y,Z) 2.6.6 [IFQ(X,Y,Z)] R5=

UMT

MT

TILT where:

[F*(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 AFD, RIL, and QPTR limits.

[F*(X,Y,Z)]RPs includes allowances for calculation and measurement uncertainties.

CNEI-0400-28 1 Page 16 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report D~XYZ Design power distributions for FQ. FQ(X,Y,Z) is provided in Appendix Table A-i for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial MQ(X,Y,Z) =

startup operations.

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 =Measurement Uncertainty (UMT =1.05)

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

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

KSLOPE = adjustment to K 1 value from OTAT trip setpoint required to compensate for each 1% measured F* (X,Y,Z) exceeds [ FL RPS (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.

CNEI-0400-28 1 Page 17 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for RFA Fuel 1.200-1.000 (0.0, 1.00) (40.0 (12.0, 0.9259)

(4.0, 0.9259) 0.800 N. 0.600 Core Height (ft) K(Z) 0.400 0.0 1.000

<4 1.000

>4 0.9259 12 0.9259 0.200 0.000 I I 1 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Core Height (ft)

CNEI-0400-28 1 Page 18 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Table 2 FQ(X,Y,Z) and F*I(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*(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 2.00 475 2.00 2.00 495 2.00 2.00 512 2.00 2.00 517 2.00 2.00 527 2.00 2.00 537 2.00 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups.

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

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

CNEI-0400-28 1 Page 19 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 2.7 Nuclear Enthalpy Rise Hot Channel Factor - FAHj(X,Y) (TS 3.2.2)

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

2.7.1 [FkH (X, y)]Lco = MARP (X,Y) * [1.0 + ~ *1. -P) where:

[FkH (X, y)]Lco is the steady-state, maximum allowed radial peak and includes allowances for calculation/measurement uncertainty.

MARP'(X,Y) = Cycle-specific operating limit Maximum Allowable Radial Peaks. MARP(X,Y) radial peaking limits are provided in Table 3.

Thermal Power

=Rated Thermal Power RRH = Thermal Power reduction required to compensate for each 1% measured radial peak, F*H (X,Y), exceeds the limit.

(RRH = 3.34, 0.O< P <1.0)

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

2.7.2 [ Fk (X,Y)]sv - F*H (X, Y) *MAH(X, Y)

UMR

TILT where:

[ F* (X,Y)]sv = Cycle dependent maximum allowable design peaking factor that ensures FAHj(X,Y) limit is not exceeded for operation within AFD, RIL, and QPTR limits. F+/-H (X,Y)sUvicue allowances for calculational and measurement uncertainty.

DD FAHt (X,Y) =Design power distribution for F,. Fj (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-0400-28 1 Page 20 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report MAHt(X,Y) =Margin remaining in core location X,Y relative to the Operational DNB limits in the transient power distribution.

MAH(X,Y) is provided in Appendix Table A-3 for normal operation and in Appendix Table A-6 for power escalation testing during initial startup operation.

UMR = Uncertainty value for measured radial peaks (UMVR = 1.0).

UMR is 1.0 since a factor of 1.04 is implicitly included in the variable M (,)

TILT =Peaking penalty to account for allowable quadrant power tilt ratio of 1.02. (TILT =1.035) 2.7.3 RRH = 3.34 where:

RLRH = Thermal Power reduction required to compensate for each 1% measured radial peak, FjAH (X,Y) exceeds its limit. (0 < P < 1.0) 2.7.4 TRH = 0.04 where:

TRH = Reduction in OTAT K1 setpoint required to compensate for each 1%

measured radial peak, F* (X,Y) exceeds its limit.

2.7.5 F*~(X,Y) Penalty Factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2.

2.8 Axial Flux Difference - AIFD (TS 3.2.3) 2.8.1 Axial Flux Difference (AFD) Limits are provided in Figure 5.

CNEI-0400-28 1 Page 21 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPs)

RFA Fuel MARPs 100% Full Power Core He ight Axial Pe ak (if) 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 1.9312 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 1.2987 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 1.7359 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

CNEI-0400-28 1 Page 22 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits 0

0) 0*

-50 -40 -30 -20 -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 the Unit 2 ROD manual for operational AFD limits.

CNEI-0400-28 1 Page 23 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 2.9 Reactor Trip System Instrumentation Setpoints (TS 3.3.1) Table 3.3.1-1 2.9.1 Overtemperature AT Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T' < 590.8 °F Nominal RCS Operating Pressure P' = 2235 psig Overtemperature AT reactor trip setpoint K1 = 1.1953 Overtemperature AT reactor trip heatup setpoint K2 = 0.03 163/°F penalty coefficient Overtemperature AT reactor trip depressurization K3 = 0.001414/psi setpoint penalty coefficient Time constants utilized in the lead-lag compensator z1= 8 sec.

for AT z2= 3 sec.

Time constant utilized in the lag compensator for AT "3= 0 sec.

Time constants utilized in the lead-lag compensator "4= 22 sec.

for Tavg "5= 4 sec.

Time constant utilized in the measured Tavg lag "6= 0 sec.

compensator fl (AI) "positive' breakpoint = 3.0 %AI fl (AI) "negative" breakpoint fl (Al) "positive" slope = 1.525 %AT 0 / %AI fl(AI) "negative" slope = N/A*

fl (Al) negative breakpoints and slopes for OTAT are less restrictive than OPAT f2 (AI) negative breakpoint and slope. Therefore, during a transient which challenges negative imbalance limits, OPAT f2 (AI) limits will result in a reactor trip before OTAT fl (Al) limits are reached. This makes implementation of an OTAT fl (Al) negative breakpoint and slope unnecessary.

CNEI-0400-28 1 Page 24 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 2.9.2 Overpower AT Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T" < 590.8 °F Overpower AT reactor trip setpoint K4 = 1.0819 Overpower AT reactor trip penalty K5= 0.02 / 0 F for increasing Tavg K5 = 0.00 / 0 F for decreasing Tavg Overpower AT reactor trip heatup setpoint K6 = 0.001291/°F for T > T" penalty coefficient K 6 =0.0 /°F for T < T" Time constants utilized in the lead-lag z1=8 sec.

compensator for AT ;2=3 sec.

Time constant utilized in the lag "3= 0 sec.

compensator for AT Time constant utilized in the measured Tavg "6= 0 sec.

lag compensator Time constant utilized in the rate-lag "7= 10 sec.

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

CNEI-0400-28 1 Page 25 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 2.10 Boron Dilution Mitigation System (TS 3.3.9) 2.10.1 Reactor Makeup Water Pump flow rate limits:

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

RCS pressure, temperature and flow limits for DNB are shown in Table 4.

2.12 Accumulators (TS 3.5.1) 2.12.1 Boron concentration limits during MODES 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 2,425 ppm Accumulator minimum boron concentration. 300.1 - 350 EFPD 2,325 ppm Accumulator minimum boron concentration. 350.1 - 400 EFPD 2,244 ppm Accumulator minimum boron concentration. 400.1 - 450 EFPD 2,170 ppm Accumulator minimum boron concentration. 450.1 - 527 EFPD 2,101 ppm Accumulator minimum boron concentration. 527.1 - 537 EFPD 1,991 ppm Accumulator maximum boron concentration. 057537 EFPD 0- EP ,7 ppm 3,075 p

CNEI-0400-28 1 Page 26 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters No. Operable PARAMETER INDICATION CHANNELS LIMITS

1. Indicated RCS Average Temperature meter 4 < 589.6 0F meter 3 < 589.3 °F computer 4 <*590.1 °F computer 3 <*589.9 *F

2. Indicated Pressurizer Pressure meter 4 > 2209.8 psig meter 3 >Ž2212.1lpsig computer 4 >2205.8 psig computer 3 > 2207.5 psig

3. RCS Total Flow Rate > 390,000 gpm

CNEI-0400-28 1 Page 27 Revision 2 Catawba 2 Cycle 21 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 pooi. 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 MODE 6 reactivity requirement of Keff_< 0.95.

Parameter Limit Minimum boron concentration of the Reactor Coolant 2,700 ppm System, the refueling canal, and the refueling cavity.

CNEI-0400-28 1 Page 28 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 2.16 Standby Shutdown System - (SLC-16.7-9) 2.16.1 Minimum boron concentration limit for the spent fuel pooi required for Standby Makeup Pump Water Supply. Applicable for MODES 1, 2, and 3.

Parameter Limit Spent fuel pooi minimum boron concentration for TR 2,700 ppm 16.7-9.3.

2.17 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 2000 gallons to maintain SDM at 68 °F NOTE:

determine When cycle burnup required is > 460 level.

BAT minimum EFPD, Figure 6 may be used to BAT Minimum Shutdown Volume (Includes the 13,086 gallons additional volumes listed in SLC 16.9-11) (14.9% level)

RWST minimum boron concentration 2,700 ppm Volume of 2,700 ppm boric acid solution required 7,000 gallons to maintain SDM at 68 °F RWST Minimum Shutdown Volume (Includes the 48,500 gallons additional volumes listed in SLC 16.9-1 1) (8.7% level)

CNEI-0400-28 1 Page 29 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report 2.18 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

> 210°F. The minimum volumes calculated support cooldown to 200 0F to satisfy UFSAR Chapter 9 requirements.

Parameter Limit BAT minimum boron concentration 7,000 ppm Volume of 7,000 ppm boric acid solution required 13,500 gallons to maintain SDM at 21 0°F INOTE: When cycle burnup is > 460 EFPD, Figure 6 may be used to determine required BAT minimum level.

BAT Minimum Shutdown Volume (Includes the 25,200 gallons additional volumes listed in SLC 16.9-12) (45.8% level)

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

CNEI-0400-28 1 Page 30 Revision 2 Catawba 2 Cycle 21 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is > 460 EFPD)

This figure includes additional volumes listed in SLC 16.9-11 and 16.9-12 50.0 ________________ _

RCS Boron 45.0 Concentration 'BAT Level

'*,*(ppm) I(%level)

...... ; ;*, 000 <<0500 . ...40.0430......

35.0 f!500.

f < 700 ......... 37.0......

700 <1000 3~0.

-* *:*J ' 100 < 1300 14.9

i13001< <2700 9.8

25. Unacceptable

'-- 20.0 Operation !!Acceptable Operation 15.0 -

10.0 5.0... ': *;

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

CNEI-0400-281I Page 31 Revision 2 Catawba 2 Cycle 21 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 2 Cycle 21 Maneuvering Analysis calculation file, CNC-1553.05-00-0618. Due to the size of monitoring factor data, Appendix A

is controlled electronically within Duke and is not included in Duke internal copies of the COLR.

Catawba Reactor and Electrical Systems Engineering controls monitoring factor via computer files and should be contacted if questions concerning this information arise.

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

Note: Revision 2 of the COLR will not transmit Appendix A because there is no change from Revision 0.