RA-20-0135, Core Operating Limits Report (COLR) for Unit 2 Cycle 24 Reload Core, Rev. 1

Core Operating Limits Report (COLR) for Unit 2 Cycle 24 Reload Core, Rev. 1

ML20140A067
Person / Time
Site:	Catawba
Issue date:	04/27/2020
From:	Simril T Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-20-0135
Download: ML20140A067 (36)

Text

Tom Slmrll

( ~ DUKE Vice President ENERGY. Catawba Nuclear Station Duke Energy CN01VP 14800 Concord Road York, SC 29745 o: 803.701 .3340 f* 803.701 .3221 tom.simril@duke-energy.com RA-20-0135 April 27, 2020 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station, Unit 2 Facility Operating License Number NPF-52 Docket Number 50-414 Core Operating Limits Report (COLR) for Unit 2 Cycle 24 Reload Core, Rev. 1 Pursuant to Catawba Technical Specification 5.6.5.d, please find attached an information copy of the subject COLR. Table 1 of the COLR was revised to present data in an equation format to enhance compliance with Technical Specification 3.1.6 (Control Bank Insertion Limits).

The power distribution monitoring factors from Appendix A of Revision O remain valid and are not transmitted as part of Revision 1.

This letter and enclosed COLR do not contain any regulatory commitments.

Please direct any questions or concerns to Sherry Andrews, Regulatory Affairs, at (803) 701-3424.

Sincerely, Tom Simril Vice President, Catawba Nuclear Station Enclosure (Catawba Unit 2 Cycle 24 COLR, Rev. 1) www.duke-energy.com

U.S. Nuclear Regulatory Commission RA-20-0135 April 27, 2020 Page l 2 xc (with enclosure):

L. Dudes, Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257 J. D. Austin, Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station M. Mahoney, Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mailstop O-8B1A Rockville, MD 20852

U.S. Nuclear Regulatory Commission RA-20-0135 April 27, 2020 Page l 3 bxc (with enclosure):

ELL-EC2ZF RGC Unit 2 COLR Manual bxc (without enclosure):

M.B. Hare W. Murphy NCMPA-1 NCEMC PMPA RGC Date File

Enclosure Catawba Unit 2 Cycle 24 COLR, Rev. 1 (paper COLR excluding Attachment A)

(_~DUKE

<{; ENERGY..

Facility Code : CN Applicable Facilities :

Document Number : CNEI-0400-355 Document Revision Number : 001 Document EC Number :

Change Reason : AR #02320950 (PRR)

Document Title : Catawba 2 Cycle 24 Core Operating Limits Report Shayotovich, Andrew Originator 3/23/2020 Hager, Nicholas R Verifier 3/23/2020 Berklite, Samson V. Safety Analysis Review 3/23/2020 Murphy, William J Site Impact Review 3/23/2020 Andrews, Sherry E Regulatory Affairs Review 3/23/2020 Robinson, Duncan Approver 3/23/2020 Notes :

CNEI-0400-355 Page 1 Revision 1 Catawba 2 Cycle 24 Core Operating Limits Report Revision 1 March 2020

Reference:

CNC-1553.05-00-0689, Rev. 1 Reload 50.59 #: 02286541 QA CONDITION 1 The information presented in this report has been prepared and issued in accordance with Catawba Technical Specification 5.6.5.

CNEI-0400-355 Page 2 Revision 1 Catawba 2 Cycle 24 Core Operating Limits Report Implementation Instructions for Revision 1 Revision Description and CR Tracking Revision 1 of the Catawba Unit 2 Cycle 24 COLR contains limits specific to the reload core.

This revision was initiated by AR #02320950 (PRR) creates a new contingency enclosure to address IRE Rod Control System (IRE) failures during repositioning/realignment of Control Banks (except for Control Bank D), and reconfigures the body of the procedure PT/0/A/4150/030 accordingly. Table 1 of the COLR is revised to present data in an equation format as requested by the PRR.

The power distribution monitoring factors from Appendix A of Revision 0 remain valid and are not changed as part of Revision 1.

Implementation Schedule The Catawba Unit 2 Cycle 24 COLR requires the reload 50.59 (AR #02286541) be approved prior to implementation and fuel loading.

Revision 1 may become effective immediately upon receipt. The Catawba unit 2 Cycle 24 COLR will cease to be effective during No MODE between cycle 24 and 25.

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

Additional Information CDR was performed by Safety Analysis confirming that there is no impact to the CNS 2 Chapter 15 safety analyses as a result of this revision.

CNS Reactor Engineering performed site inspection in accordance with AD-NF-ALL-0807 and AD-NF-NGO-0214.

CNS Regulatory Affairs performed a site review to verify that this change meets regulatory requirements.

CNEI-0400-355 Page 3 Revision 1 Catawba 2 Cycle 24 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected COLR 0 September 2019 1-31, Appendix A* C2C24 COLR, Rev. 0 1 March 2020 1-3, 13 C2C24 COLR, Rev. 1

• 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-0400-355 Page 4 Revision 0 Catawba 2 Cycle 24 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.

NRC Approved TS Technical Specifications COLR COLR Parameter Methodology (Section Section Section 1.1 Number) 2.1.1 Reactor Core Safety Limits RCS Temperature and 2.1 6, 7, 8, 9, 10, 12, 15, Pressure Safety Limits 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 12, 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 12, 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 FQ 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 2.10 6, 7, 8, 14, 16 Rate 3.4.1 RCS Pressure, Temperature and Flow limits RCS Pressure, Temperature 2.11 6, 7, 8, 9, 10, 12, for DNB and Flow 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 Licensee Commitments that reference this report are listed below NRC Approved SLC Selected Licensing COLR COLR Parameter Methodology (Section Section Commitment 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 Borated Water Volume and Conc. for 2.17 6, 7, 8, 14, 16 Source - Shutdown BAT/RWST 16.9-12 Boration Systems - Borated Water Borated Water Volume and Conc. for 2.18 6, 7, 8, 14, 16 Source - Operating BAT/RWST

CNEI-0400-355 Page 5 Revision 0 Catawba 2 Cycle 24 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 Report Date: 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 LOCA, (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-0400-355 Page 6 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report 1.1 Analytical Methods (continued)

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

Revision 5a Report Date: October 2012

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

Revision 1 Report Date: March 2015

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

Revision 4c Report Date: January 2019

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

Revision 2a Report Date: December 2008

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

Revision 5 Report Date: March 2016

11. DPC-NE-2008-PA, "Fuel Mechanical Reload Analysis Methodology Using TACO3," (Duke Energy Proprietary).

Revision 0 Report Date: April 1995 Not Used

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

Revision 3c Report Date: March 2017

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

Revision 1a Report Date: January 2009 Not Used

CNEI-0400-355 Page 7 Revision 0 Catawba 2 Cycle 24 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, "Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," (Duke Energy Proprietary).

Revision 1a Report Date: June 2009

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

(Duke Energy Proprietary).

Revision 1 Report Date: November 2008

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, McGuire Nuclear Station Catawba Nuclear Station Conditional Exemption of the EOC MTC Measurement Methodology, (Duke Energy 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', (W Proprietary).

Revision 0 Report Date: July 2006

CNEI-0400-355 Page 8 Revision 0 Catawba 2 Cycle 24 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% 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-0400-355 Page 9 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation 670 DO NOT OPERATE IN THIS AREA 660 650 640 2400 psia RCS Tavg (°F) 630 2280 psia 620 2100 psia 610 1945 psia 600 590 ACCEPTABLE OPERATION 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power

CNEI-0400-355 Page 10 Revision 0 Catawba 2 Cycle 24 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 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 an 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 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-0400-355 Page 11 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 0.9 Unacceptable Operation Moderator Temperature Coefficient 0.8 0.7 0.6 0.5 (1.0E-04 K/K/F)

Acceptable Operation 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-0400-355 Page 12 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power Fully Withdrawn (Maximum = 231)

~

231 220 - - - --

/

/

/ "

., /

/ - - -

/ /

200

/

/

Fully Withdrawn /

... /

V I I/

(Minimum = 222)

Rod Insertion Position (Steps Withdrawn) 180 / I Control Bank B I/

F

/

/

/

/ I (100%, 161) 160

~ (0%, 163)

/

/

/

., /

/

140 ., / /

/ /

I Control Bank C I V 120 /

/ I I I/

/

/ /

/ /

100 /

/

/

/ /

80 /

V

/

V

/

I/

/

Control Bank D 60

/ /

/

., /

40 ~ (0%, 47) I

/

/

~ ~ Fully Inserted

/

20 ., /

(30%, 0) /

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

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

Bank CC RIL = 2.3(P) +47 {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 = % 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-0400-355 Page 13 Revision 1 Catawba 2 Cycle 24 Core Operating Limits Report Table 1 Control Bank Withdrawal Sequence Equation Control Control Control Control Bank A Bank B Bank C Bank D 0 Start 0 0 0 116 0 Start 0 0 CBA Stop CBA - 116 0 0 CBA 116 0 Start 0 CBA CBB Stop CBB - 116 0 CBA CBB 116 0 Start CBA CBB CBC CBC - 116 Where:

CBA = Fully withdrawn position of Control Bank A CBB = Fully withdrawn position of Control Bank B CBC = Fully withdrawn position of Control Bank C Allowed Control Bank Fully Withdrawn Positions Range from 223 Steps to 231 Steps for frequent RCCA Reposition Required per CNEI-0400-091, Rev. 6, RCCA Axial Repositioning Schedule for Catawba Nuclear Station.

CNEI-0400-355 Page 14 Revision 0 Catawba 2 Cycle 24 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 QRTP *K(Z)/P for P > 0.5 F QRTP *K(Z)/0.5 for P < 0.5 where, Thermal Power P =

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 QRTP = 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 QRTP with the 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.

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

D L FQ(X,Y,Z)

• MQ(X,Y,Z) 2.6.5 [FQ(X,Y,Z)]OP = UMT

• MT

• TILT where:

[ FQL (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.

OP FQL (X,Y,Z) includes allowances for calculation and measurement uncertainties.

FQD (X,Y,Z) = Design power distribution for FQ. FQD (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-0400-355 Page 15 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report MQ(X,Y,Z) = Margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution. MQ(X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial startup operation.

UMT = Total Peak Measurement Uncertainty. (UMT = 1.05)

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

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

D L RPS FQ(X,Y,Z)

• MC(X,Y,Z) 2.6.6 [FQ(X,Y,Z)] = UMT

• MT

• TILT where:

L

[FQ(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.

L

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

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

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 = Defined in Section 2.6.5.

MT = Defined in Section 2.6.5.

TILT = Defined in Section 2.6.5.

CNEI-0400-355 Page 16 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report 2.6.7 KSLOPE = 0.0725 where:

KSLOPE = adjustment to K1 value from OTT trip setpoint required to compensate for each 1% measured FQM (X,Y,Z)

RPS exceeds [ FQL (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-0400-355 Page 17 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for RFA Fuel 1.200 (0.0, 1.00) (4.0, 1.00) 1.000 I (12.0, 0.9259)

(4.0, 0.9259) 0.800 K(Z) 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 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Core Height (ft)

CNEI-0400-355 Page 18 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Table 2 FQ(X,Y,Z) and FH(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) FH(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 2.00 500 2.00 2.00 510 2.00 2.00 520 2.00 2.00 530 2.00 2.00 540 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 FH(X,Y) for compliance with Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.

CNEI-0400-355 Page 19 Revision 0 Catawba 2 Cycle 24 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.

1 2.7.1 [FLH (X, Y)]LCO = MARP (X,Y)

• 1.0 + RRH * (1.0 - P) where:

[FLH (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 P =

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

(RRH = 3.34, 0.0 < P < 1.0)

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

SURV FDH (X, Y)

• M H (X, Y) 2.7.2 [ F (X,Y)]

L H =

UMR

• TILT where:

SURV

[ FLH (X,Y)] = Cycle dependent maximum allowable design peaking factor that ensures FH(X,Y) limit is not exceeded for operation SURV within AFD, RIL, and QPTR limits. F L (X,Y) includes H

allowances for calculation and measurement uncertainty.

D D FH (X,Y) = Design power distribution for FH. FH (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-0400-355 Page 20 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report MH(X,Y) = Margin remaining in core location X,Y relative to the 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 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 K1 setpoint required to compensate for each 1%

measured radial peak, FMH (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 Difference - AFD (TS 3.2.3) 2.8.1 Axial Flux Difference (AFD) Limits are provided in Figure 5.

CNEI-0400-355 Page 21 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPs)

RFA Steady State Limiting Value Between Loss of Flow Accident (LOFA) MARPs and FHLOCA Core Axial Peak 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 3.25 0.12 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.3151 1.2461 1.20 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.3007 1.2235 2.40 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.4633 1.4616 3.60 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.4675 1.3874 4.80 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.2987 1.2579 6.00 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.3293 1.2602 7.20 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.5982 1.2871 1.2195 8.40 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6010 1.5127 1.2182 1.1578 9.60 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.5808 1.5301 1.4444 1.1431 1.0914 10.80 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.5743 1.5573 1.5088 1.4624 1.3832 1.1009 1.0470 11.40 1.6058 1.6058 1.6058 1.6058 1.6057 1.5826 1.5289 1.5098 1.4637 1.4218 1.3458 1.0670 1.0142

CNEI-0400-355 Page 22 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits

(-20, 100) 100 (+10, 100)

Unacceptable Operation Unacceptable Operation 90 80 Percent of Rated Thermal Power 70 60 Acceptable Operation 50

(-36, 50)

(+21, 50) 40 30 20 10 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-0400-355 Page 23 Revision 0 Catawba 2 Cycle 24 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' < 590.8 °F Nominal RCS Operating Pressure P' = 2235 psig Overtemperature T reactor trip setpoint K1 = 1.1953 Overtemperature T reactor trip heatup setpoint K2 = 0.03163/°F penalty coefficient Overtemperature T reactor trip depressurization K3 = 0.001414/psi setpoint penalty coefficient Time constants utilized in the lead-lag compensator 1 = 8 sec.

for T 2 = 3 sec.

Time constant utilized in the lag compensator for T 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 f1(I) "positive" breakpoint = 3.0 %I f1(I) "negative" breakpoint = N/A*

f1(I) "positive" slope = 1.525 %T0/ %I f1(I) "negative" slope = /

• f1 (I) negative breakpoints and slopes for OTT are less restrictive than OPT f2 (I) negative breakpoint and slope. Therefore, during a transient which challenges negative imbalance limits, OPT f2 (I) limits will result in a reactor trip before OTT f1 (I) limits are reached. This makes implementation of an OTT f1 (I) negative breakpoint and slope unnecessary.

CNEI-0400-355 Page 24 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report 2.9.2 Overpower T Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T" < 590.8 °F Overpower T reactor trip setpoint K4 = 1.0819 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 K6 = 0.001291/°F for T > T

penalty coefficient K6 = 0.0 /°F for T < T

Time constants utilized in the lead-lag 1 = 8 sec.

compensator for T 2 = 3 sec.

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

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

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

for Tavg 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

CNEI-0400-355 Page 25 Revision 0 Catawba 2 Cycle 24 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 < 80 gpm MODE 4 or 5 < 70 gpm 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 - 300 EFPD 2,500 ppm Accumulator minimum boron concentration. 300.1 - 400 EFPD 2,377 ppm Accumulator minimum boron concentration. 400.1 - 530 EFPD 2,216 ppm Accumulator minimum boron concentration. 530.1 - 540 EFPD 2,032 ppm Accumulator maximum boron concentration. 0 - 540 EFPD 3,075 ppm

CNEI-0400-355 Page 26 Revision 0 Catawba 2 Cycle 24 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 F 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 2212.1 psig computer 4 2205.8 psig computer 3 2207.5 psig

3. RCS Total Flow Rate 387,000 gpm

CNEI-0400-355 Page 27 Revision 0 Catawba 2 Cycle 24 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 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-0400-355 Page 28 Revision 0 Catawba 2 Cycle 24 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 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 < 210oF, 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 oF NOTE: When cycle burnup is > 469 EFPD, Figure 6 may be used to determine required BAT minimum level.

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 oF RWST Minimum Shutdown Volume (Includes the 48,500 gallons additional volumes listed in SLC 16.9-11) (8.7% level)

CNEI-0400-355 Page 29 Revision 0 Catawba 2 Cycle 24 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 oF *.

• 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 13,500 gallons to maintain SDM at 210oF NOTE: When cycle burnup is > 469 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 oF RWST Minimum Shutdown Volume (Includes the 98,607 gallons additional volumes listed in SLC 16.9-12) (22.0% level)

CNEI-0400-355 Page 30 Revision 0 Catawba 2 Cycle 24 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is > 469 EFPD)

This figure includes additional volumes listed in SLC 16.9-11 and 16.9-12 50.0 I RCS Boron 45.0 +

Concentration BAT Level 40.0 I +

(ppm) 0 < 300

(%level) 43.0 I 300 < 500 40.0 35.0 500 < 700 37.0 f---

700 < 1000 30.0 1000 < 1300 14.9 BAT Level (% Level) 30.0 1300 < 2700 9.8

> 2700 9.8 25.0 Unacceptable I 20.0 Operation Acceptable Operation +- + +-

15.0 t +

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-0400-355 Page 31 Revision 0 Catawba 2 Cycle 24 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 24 Maneuvering Analysis calculation file, CNC-1553.05-00-0678, Rev 1. 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 Systems Support 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.