MNS-17-009, Cycle 24, Revision 3 - Core Operating Limits Report

Cycle 24, Revision 3 - Core Operating Limits Report

ML17073A090
Person / Time
Site:	Mcguire
Issue date:	03/03/2017
From:	Capps S Duke Energy Carolinas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
MNS-17-009
Download: ML17073A090 (33)

Text

(~DUKE

• -X~

Steven D. Capps

~

Vice President ENERGY McGuire Nuclear Station Duke Energy MGOl VP I 12700 Hagers Ferry Road Huntersville, NC 28078 0: 980.875.4805 f: 980.875.4809 Steven.Capps@duke-energy.com Serial No:*.MNS-17-009 March 3, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC McGuire Nuclear Station Docket No. 50-370 Unit 2, Cycle 24, Revision 3 Core Operating Limits Report Pursuant to McGuire Technical Specification 5.6.5.d, please find enclosed the McGuire Unit 2 Cycle 24, Revision 3, Core Operating Limits Report (COLR).

Questions regarding this submittal should be directed to P.T. Vu, Regulatory Affairs at (980) 875-4302.

Steven D. Capps Attachment www.duke-energy.com

U.S. Nuclear Regulatory Commission March 3, 2017 Page2 xc. C. Haney, Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Ave., NE, Suite 1200 Atlanta, GA 30303-1257 M. Mahoney, Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mail Stop 0-8G9A Rockville, MD 20852-2738 A. Hutto

.NRC Senior Resident Inspector McGuire Nuclear Station

MCEI-0400-318 Page 1 Revision 3 McGuire Unit 2 Cycle 24 Core Operating Limits Report Revision 3 January 2017 Calculation Number: MCC-1553.05-00-0618, Revision 3 Duke Energy Carolinas, LLC QA Condition 1 The information presented in this report has been prepared and issued in accordance with McGuire Technical Specification 5.6.5.

.i I

MCEI-0400-318 Page2 Revision 3 McGuire 2 Cycle 24 Core Operating Limits Report Implementation Instructions For Revision 3 Revision Description and CR Tracking Revision 3 of the McGuire Unit 2 Cycle 24 COLR contains limits specific to the reload core, and is revised to reflect increase in analyzed burnup window. The power distribution monitoring factors from Appendix A of Revision 0 remain valid and are not transmitted as part of Revision 3.

There is no CR associated with this revision, however the McGuire 2 Cycle 24 burnup is tracked by NTM #01975574-2.

Implementation Schedule Revision 3 may become effective immediately upon receipt.

The McGuire Unit 2 Cycle 24 COLR will cease to be effective during No MODE between cycles 24 and 25.

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

MCEI-0400-318 Page 3 Revision3 McGuire 2 Cycle 24 Core Operating Limits Report REVISION LOG Revision Effective Date _ Pages Affected COLR 0 August 2015 1-31, Appendix A* M2C24 COLR, Rev. 0 1 September 2015 1-31 M2C24 COLR, Rev. 1 2 May 2016 1-31 M2C24 COLR, Rev. 2 3 January 2017 1-3, 18, 25 M2C24 COLR, Rev. 3

• 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.

MCEI-0400-318 Page4 Revision 2 McGuire 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 the requirements of Technical Specification 5.6.5. The Technical Specifications that reference this report are listed below along with the NRC approved analytical methods used to develop and/or determine COLR parameters in Technical Specifications.

NRC Approved TS COLR Methodology (Section Number Technical Specifications COLR Parameter Section 1.1 Number) 2.1.1 Reactor Core Safety Limits RCS Temperature and 2.1 6,7,8,9,10,12,15,16

'"Pressure Safety Limits 3.1.1 Shutdown Margin Shutdown Margin 2.2 6,7,8,12,14,15,l6 3.1.3 Moderator Temperature Coefficient MTC 2.3 6,7,8,12,14,16, 17 3.1.4 Rod Group Alignment Limits Shutdown Margin 2.2 6,7,8,12,14,15,16 3.1.5 Shutdown Bank Inse1tion Limits Shutdown Margin 2.2 6, 7,8,12,14,15,16 3.1.5 Shutdown Bank Inse1tion Limits Shutdown Bank Inse1tion 2.4 2,4,6, 7,8,9,10, 12,14,15, Limit 16 3.1.6 Control Bank Insertion Limits Shutdown Margin 2.2 6, 7,8,12,14, 15, 16 3.1.6 Control Bank Insertion Limits Control Bank Insertion 2.5 2,4,6, 7,8,9,10,12,14,15, Limit 16 3.1.8 Physics Tests Exceptions Shutdown Margin 2.2 6,7,8, 12,14,15,16 3.2.l Heat Flux Hot Channel Factor Fq, AFD, OTiiT and 2.6 2,4,6,7,8,9, 10, 12,15,16 Penalty Factors 3.2.2 Nuclear Enthalpy Rise Hot Channel F iiH, AFD and 2.7 2,4,6,7,8,9, 10, 12,15,16 Factor Penalty Factors 3.2.3 ' Axial Flux Difference AFD 2.8 2,4,6,7,8,15,16 3.3.l Reactor Trip System Instrumentation OTiiT and OPiiT 2.9 6,7,8,9,10,12, 15,16 Setpoints Constants 3.4.1 RCS Pressure, Temperature, and Flow RCS Pressure, 2.10 6,7,8,9,10,12 DNB limits Temperature and Flow 3.5.1 Accumulators Max and Min Boron Cone. 2.11 6,7,8,12,14,16 3.5.4 Refueling Water Storage Tank Max and Min Boron Cone. 2.12 6,7,8,12,14,16 3.7.14 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.13 6,7,8,12,14,16 3.9.1 Refueling Operations Boron Min Boron Concentration 2.14 6,7,8,12,14,16 Concentration 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 COLR SLC Methodology Selected Licensing Commitment COLR Parameter Section Number (Section 1.1 Number) 16.9.14 Borated Water Source - Shutdown Borated Water Volume and 2.15 6,7,8,12,14,16 Cone. for BAT/RWST 16.9.11 Borated Water Source - Operating Borated Water Volume and 2.16 6,7,8,12,14,16 Cone. for BAT/RWST 16.9.7 Standby Shutdown System Standby Makeup Pump Water 2.17 6,7,8,12,14,16 Suooly

MCEI-0400-318 Page 5 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 1.1 Analytical Methods The analytical methods used to determine core operating limits for parameters identified in Technical Specifications and previously reviewed and approved by the NRC as specified in Technical Specification 5.6.5 are as follows.

l. 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," (Jj_ 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).

Revision2 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," (Jj_ Proprietary).

Revision: Volume 1(Revision2) and Volumes 2-5 (Revision 1)

Report Date: March 1998

5. BAW-10 l 68P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B&W Proprietary).

Revision I SER Date: January 22, 1991 Revision2 SER Dates: August 22, 1996 and November 26, 1996 Revision 3 SER Date: June 15, 1994 Not Used

MCEI-0400-318 Page6 Revision2 McGuire 2 Cycle 24 Core Operating Limits Report 1.1 Analytical Methods (continued)

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

Revision Sa Report Date: October 2012

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

Revision Oa Report Date: May 2009 Note: The WLQP 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. MF3119, MF3120, MF3121, and MF3122), March 25, 2015.

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

Revision4b Report Date: September 2010

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

Revision 2a Report Date: December 2008

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

Revision 4a Report Date: December 2008

11. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TAC03," (DPC Proprietary).

Revision 0 Report Date: April 3, 1995 Not Used

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

Revision 3a Report Date: September 2011

MCEI-0400-318 Page?

Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 1.1 Analytical Methods (continued)

13. DPC-NE-1004A, "Nuclear Design Methodology Using CASM0-3/SIMULATE-3P ."

Revision la Report Date: January 2009 Not Used

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

Revision 2a Report Date: December 2009

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

• Revision la Report Date: June 2009

16. DPC-NE-1005-PA, "Nuclear Design Methodology Using CASM0-4 / SIMULATE-3 MOX,

(DPC Proprietary).

Revision l Repo1t Date: November 12, 2008

17. DPC-NE-1007-P A, "Conditional Exemption of the BOC MTC Measurement Methodology,"

• (DPC and W Proprietary)

Revision 0 Report Date: April 2015

MCEI-0400-318 Page 8 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 2.0 Operating Limits Cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC approved methodologies specified in Section 1.1.

2.1 Reactor Core Safety Limits (TS 2.1.1) 2.1.1 The Reactor Core Safety Limits are shown in Figure I.

2.2 Shutdown Margin - SDM (TS 3.1.1, TS 3.1.4, TS 3.1.5, TS 3.1.6 and TS 3.1.8) 2.2.1 For TS 3 .1.1, SDM shall be> 1.3% L\KIK in MODE 2 with k-eff < 1.0 and in MODES 3 and 4.

2.2.2 For TS 3.1.1, SDM shall be~ 1.0% L\KIK in MODE 5.

2.2.3 For TS 3.1.4, SDM shall be~ 1.3% L\KIK in MODES 1 and 2.

2.2.4 For TS 3.1.S, SDM shall be~ 1.3% L\KIK in MODE 1 and MODE 2 with any control bank not fully inserted.

2.2.5 For TS 3.1 .6, SDM shall be~ 1.3% L\KIK in MODE 1 and MODE 2 with K-eff 2: 1.0.

2.2.6 For TS 3 .1.8, SDM shall be~ 1.3% L\KIK in MODE 2 during PHYSICS TESTS.

MCEI-0400-318 Page9 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation DO NOT OPERA TE IN Tms AREA 660 1 - - - - - - - + - - - - - l - - - - - - + ---~

~ 630 1------+---""'---+------f-----""'o~----~\-*----l r:n.

~ 620 ACCEPTABLE OPERATION 580 '--~~~--'-~~~~-'-~~~~_._~~~~"--~~~--'~~~~--'

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

MCEI-0400-318 Page 10 Revision2 McGuire 2 Cycle 24 Core Operating Limits Report 2.3 Moderator Temperature Coefficient- MTC (TS 3.1.3) 2.3.1 The 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 AKIK/°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 AKIK/°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. l .3 .2 is not required to be performed.

2.3.4 60 PPM MTC Surveillance Limit is:

60 PPM ARO, equilibrium RTP MTC shall be less negative than or equal to

-4.125E-04 AKIK/°F.

Where: BOC Beginning of Cycle (burnup corresponding to the most positive MTC)

EOC = End of Cycle ARO= All Rods Out HZP =Hot Zero 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.

MCEI-0400-318 Page 11 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 0.9 Unacceptable Operation

..... 0.8

.....=

~

(.I

§Cl 0.7 lj-4

~

u~=-

~ 0 0.6

~~

t <l 0.5 Acceptable Operation c:i..""' 0.4 e 'T

~~ 0.3

""cie.

C'¢ 0.2 "O""

~

Cl

~ 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 OP/2/A/6100/22 Unit 2 Data Book for details.

MCEI-0400-318 Page 12 Revision 2 McGuire 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 /

=it: 180 /

/

Control Bank B - (Minimum= 222) /

/

/

co: .A (100%, 161) i=

~-

/

]

160 R co%, 163) ,, I/

/

,, /

/

~g. 140 ~

/ /

~

~

120 , ,,

/

Control Bank C

,,, /

r

/

=

0

100

/

/

/ - ,,

/

/

~

0 _,

=

e0 80

/

/

/

/

,,, - Control Bank D f== ~

~ 60

....."'=

/ / I 40 1=1

,, /

"!;I (0%,47) 0

~

20 ]Fully Inserted ~

/

(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 <JOO}

Bank CC RIL 2.3(P) +47 {O P 76.J} for CC RIL = 222 {76.J < P <JOO}

Bank CB RIL = 2.3(P) + J 63 {O < P < 25. 7} for CB RIL 222 {25. 7 < P < JOO}

where P = %Rated Thermal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.

Refer to OP/2/A/6100/22 Unit 2 Data Book for details.

MCEI-0400-318 Page 13 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Table 1 RCCA Withdrawal Steps and Sequence Fully Withdrawn at 222 Steps Fully Withdrawn at 223 Steps Control Control Control Control Control Control Control Control Bank A BankB Banke BankD Bank A BankB BankC BankD 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 222 Stop 106 0 0 223 Stop 107 0 0 222 116 0 Start 0 223 116 0 Start 0 222 .222 Stop 106 0 223 223 Stop 107 0 222 222 116 0 Start 223 223 116 0 Start 222 222 222 Stop 106 223 223 223 Stoi; 107 Fully Withdrawn at 224 Steps Fully Withdrawn at 225 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB Banke BankD O Start 0 0 0 0 Start 0 0 0 I 16 0 Start 0 0 116 OStart 0 0 224 Stop 108 0 0 . 225 Stop 109 0 0 224 116 0 Start 0 225 116 O Start 0 224 224 Stop 10& 0 225 225 Stop 109 0 224 224 116 0 Start 225 225 116 0 Start 224 224 224 Stop 108 225 225 225 Stop 109 Fully Withdrawn at 226 Steps Fully Withdrawn at 227 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB BankC BankD 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 226 Stop 110 0 0 227 Stop I 11 0 0 226 116 O Start 0 227 116 O Start 0 226 226 Stop 110 0 227 227 Stop 111 0 226 226 116 0 Start 227 227 116 0 Start 226 226 226 Stop 110 227 227 227 Stop 111 Fully Withdrawn at 228 Steps Fully Withdrawn at 229 Steps Control Control Control Control Control Control Control Control Bank A BankB BankC BankD Bank A BankB Banke BankD 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 228 Stop 112 0 0 229 Stop 113 0 0 228 116 0 Start 0 229 116 0 Start 0 228 228 Stop 112 0 229 229 Stop 113 0 228 228 116 0 Start 229 229 116 0 Start 228 228 228 Stop 112 229 229 229 Stop 113 Fully Withdrawn at 230 Steps Fully Withdrawn at 231 Steps Control Control Control Control Control Control Control Control Bank A BankB Banke BankD Bank A BankB BankC BankD 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 230 Stop 114 0 0 231 Stop ll5 0 0 230 116 0 Start 0 231 116 0 Start 0 230 230 Stop 114 0 231 231 Stop 115 0 230 230 ll6 OStart 231 231 116 0 Start 230 230 230 Stop ll4 231 231 231 Stop 115

MCEI-0400-318 Page 14 Revision 2 McGuire 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 ~TP *K(Z)/P for P > 0.5 F ~TP *K(Z)/0.5 for P:::;: 0.5 where, P (Thermal Power)/(Rated Power)

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

2.6.2 F ~TP = 2. 70 x K(BU) 2.6.3 K(Z) is the normalized F Q(X, Y,Z) as a function of core height. The K(Z) function for Westinghouse RFA fuel is provided in Figure 4.

2.6.4 K(BU) is the normalized FQ(X,Y,Z) as a function ofburnup. F ~TP with the K(BU) penalty for Westinghouse RFA fuel is analytically confirmed in cycle-specific reload calculations. K(BU) is set to 1.0 at all burnups.

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

L F~(X,Y,Z)

• M 0 (X,Y,Z) 2.6.5 FQ(X,Y,Z)OP = UMT *MT* TILT

MCEI-0400-318 Page 15 Revision2 McGuire 2 Cycle 24 Core Operating Limits Report where:

FJ {X,Y,Z)OP = Cycle dependent maximum allowable design peaking factor that ensures F Q(X, Y,Z) LOCA limit will be preserved for operation within the LCO limits. FJ (X,Y,z)OP includes allowances for calculation and measurement uncertainties.

F/j (X,Y,Z) Design power distribution for FQ. F/j (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.

~(X,Y,Z) Margin remaining in core location X,Y,Z to the LOCA limit in the transient power distribution. ~(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.

UMf Total Peak Measurement Uncertainty. (UMT = 1.05)

Mf Engineering Hot Channel Factor. (MT= 1.03)

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

L RPS F~(X,Y,Z)

• Mc(X,Y,Z) 2.6.6 FQ(X,Y,Z)

UMf *MT* TILT where:

F~(X,Y,Z)RPS = Cycle dependent maximum allowable design peaking factor that ensures the FQ(X,Y,Z) Centerline Fuel Melt (CFM) limit will be preserved for operation within the LCO limits.

F~(X,Y,Z)RPS includes allowances for calculation and measurement uncertainties.

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

MCEI-0400-318 Page 16 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Mc(X,Y,Z) Margin remaining to the CFM limit in core location X,Y,Z from the transient power distribution. Mc(X,Y,Z) is provided in Appendix Table A-2 for normal operating conditions and in Appendix Table A-5 for power escalation testing during initial startup operation.

UMT = Defined in Section 2.6.5.

MT Defined in Section 2.6.5.

TILT Defined in Section 2.6.5.

2.6.7 KSLOPE = 0.0725 where:

KSLOPE is the adjustment to KI value from the 0Ti1.T trip setpoint required to compensate for each 1% that Ft/ (X,Y,Z) exceeds Ft (X,Y,Z)RPS .

2.6.8 FQ(X,Y,Z) penalty factors for Technical Specification Surveillances 3.2.1.2 and 3.2.1.3 are provided in Table 2.

MCEI-0400-318 Page 17 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for Westinghouse RFA Fuel 1.200 ~---~--------------------~-~

(0.0, l_.OO) (4.0, 1.00) i.ooo .,_ _______ ...,.

1 (12.0, 0.9259)

(4.0, 0.9259) 0.800 g 0.600

~

0.400 Core Height (ft} K(Z) 0.0 1.0

S4.0 1.0 0.200

>4.0 0.9259 12.0 0.9259 0.000 0.0 2.0 4.0 6.0 8.0 10.0 12.0 Core Height (ft)

MCEI-0400-318 Page 18 Revision 3 McGuire 2 Cycle 24 Core Operating Limits Report Table2 FQ(X,Y,Z) and Fm(X,Y) Penalty Factors For Technical Specification Surveillance's 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burn up FQ(X,Y,Z) F,m(X,Y)

(EFPD) Penalty Factor (%) Penalty Factor(%)

0 2.00 2.00 4 2.00 2.00 12 2.51 2.00 25 2.21 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 2.00 465 2.00 2.00 475 2.00 2.00 496 2.00 2.00 511 2.00 2.00 521 2.00 2.00 535 2.00 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups. All cycle burnups outside of the range of the table shall use a 2% penalty factor for both FQ(X,Y,Z) and FAff(X,Y) for compliance with the Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.

MCEI-0400-318, Page 19 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 2.7 Nuclear Enthalpy Rise Hot Channel Factor-FM:I(X,Y) (TS 3.2.2)

F MI steady-state limits referred to in Technical Specification 3.2.2 is defined by the following relationship.

where:

F!ff (X, Y)LCo is the steady-state, maximum allowed radial peak and includes allowances for calculation/measurement uncertainty.

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

p = Thermal Power Rated Thermal Power

• RRH = Thermal Power reduction required to compensate* for each 1% that the measured radial peak, Fk:i (X,Y), exceeds its limit. RRH also is used to scale the MARP limits as a function of power per the FkH (X, Y) Leo equation. (RRH = 3.34 (0.0 < P :S 1.0))

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

L SURV - F~H(X, Y) x MAfl(X, Y) 2.7.2 F.&a(X,Y) - ~~--~~-

UMRxTILT where:

SURV F!ff (X,Y) = Cycle dependent maximum allowable design peaking factor that ensures the F.&H(X,Y) limit will be preserved for operation within the LCO limits. FkH (X,Y) SURV includes allowances for calculation/measurement uncertainty.

MCEI-0400-318 Page 20 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report F~ (X,Y) Design radial power distribution for F nff F~H (X,Y) is provided in Appendix Table A-3 for normal operation and in Appendix Table A-6 for power escalation testing during initial startup operation.

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

MAfI(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 MnH(X,Y).

TILT = Peaking penalty to account for allowable quadrant power tilt ratio of l.02 (TILT 1.035).

2.7.3 RRH is defined in Section 2.7.1.

2.7.4 TRH 0.04 where:

TRH =Reduction in the OTL\T KI setpoint required to compensate for each I%

that the measured radial peak, F~ (X,Y) exceeds its limit.

2.7.5 FnH (X,Y) penalty factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2.

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

MCEI-0400-318 Page 21 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPS)

RFAMARPS Core Axial Peak Ht (ft.) 1.05 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 bl 3.0 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 1.2 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 2.4 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 3.6 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 4.8 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 6.0 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 7.2 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 8.4 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 9.6 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 10.8 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 11.4 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

MCEI-0400-318 Page 22 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits

(-18, 100) (+10, 100) 90 Unacceptable Operation Unacceptable Operation 80

......ii:: Acceptable Operation Q

~ 70

~

......a

..c: 60 E--i "O

~

~

Q

(-36, 50) 50

(+21, 50)

.... 40

=

~

30 20 10

-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 OP/2/A/6100/22 Unit 2 Data Book for more details.

MCEI-0400-318 Page 23 Revision 2 McGuire 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 AT Setpoint Parameter Values Parameter Nominal Tavg at RTP T'::; 585.1°F Nominal RCS Operating Pressure P' = 2235 psig Overtemperature AT reactor trip setpoint K 1 .:S 1.1978 Overtemperature AT reactor trip heatup setpoint K2 = 0.03341°F penalty coefficient Ove11emperature AT reactor trip depressurization K3 = 0.00160llpsi setpoint penalty coefficient Time constants utilized in the lead-lag compensator 'tl 2: 8 sec.

for AT 't2 S 3 sec.

Time constant utilized in the lag compensator for AT 't3 S2 sec.

Time constants utilized in the lead-lag compensator 't4 2: 28 sec.

for Tavg -r5:::; 4 sec.

Time constant utilized in the measured T avg lag 't6 ::; 2 sec.

compensator f 1(Al) "positive" breakpoint = 19.0 %Al f1(AI) "negative" breakpoint =NIA*

f 1(Al) "positive" slope 1.769 %ATof %AI f1 (AI) "negative" slope =NIA*

• The f1(Al) "negative" breakpoint and the f1(AI) "negative" slope are less restrictive than the OPAT f2(AI) negative breakpoint and slope. Therefore, during a transient which challenges the negative imbalance limits, the OPAT f2(AI) limits will result in a reactor trip before the OTAT f1 (Al) limits are reached. This makes implementation of the OTAT f1 (AI) negative breakpoint and slope unnecessary.

MCEI-0400-318 Page 24 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 2.9.2 Overpower AT Setpoint Parameter Values Parameter Nominal Tavg at RTP T" _:::: 585.1°F Overpower AT reactor trip setpoint Overpower AT reactor trip Penalty Ks= 0.02/°F for increasing Tavg Ks 0.0 for decreasing Tavg Overpower AT reactor trip heatup*

• K6 = 0.001179/°F for T > T" setpoint penalty coefficient K6 0.0 for T < T" Time constants utilized in the lead- .,; 1 > 8 sec.

lag compensator for AT ..2 .:::: 3 sec.

Time constant utilized in the lag .,;3 < 2 sec.

compensator for AT Time constant utilized in the .,; 6 < 2 sec.

measured T avg lag 'compensator Time constant utilized in the rate-lag .,;7 > 5 sec.

controller for Tavg f2(AI) "positive" breakpoint =35.0 %Al f 2(AI) "negative" breakpoint =-35.0 %AI f 2(AI) "positive" slope 7.0 %ATc/ %AI fi{AI) "negative" slope 7.0 %AT0/ %Al

MCEI-0400-318 Page 25 Revision 3 McGuire 2 Cycle 24 Core Operating Limits Report 2.10 RCS Pressure, Temperature and Flow Limits for DNB (TS 3.4.1) 2.10.1 RCS pressure, temperature and flow limits for DNB are shown in Table 4.

2.11 Accumulators (TS 3.5.1) 2.11.1 Boron concentration limits during MODES 1 and 2, and MODE 3 with RCS pressure > 1000 psi:

• Parameter Applicable Burnup Limit Accumulator minimum boron 0-200 EFPD 2,475 ppm concentration.

Accumulator minimum boron 200.1 - 250 EFPD 2,475 ppm concentration.

Accumulator minimum boron 250.1 - 300 EFPD 2,398 ppm concentration.

Accumulator minimum boron 300.l - 350 EFPD 2,292 ppm concentration.

Accumulator minimum boron 350.1 -400 EFPD 2,214 ppm concentration.

Accumulator minimum boron 400.1 - 450 EFPD 2,143 ppm concentration.

Accumulator minimum boron 450.1 - 465 EFPD 2,077 ppm concentration.

Accumulator minimum boron 465.1 - 521 EFPD 2,057 ppm concentration.

Accumulator minim um boron 521.1 - 535 EFPD 1,972 ppm concentration.

Accumulator maximum boron 0- 535 EFPD 2,875 ppm concentration.

2.12 Refueling Water Storage Tank - RWST (TS 3.5.4) 2.12.1 Boron concentration limits during MODES 1, 2, 3, and 4:

Parameter RWST minimum boron concentration. 2,675 ppm RWST maximum boron concentration. 2,875 ppm

MCEI-0400-318 Page 26 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report Table4 Reactor Coolant System DNB Parameters No. Operable Parameter Indication Channels Limits

1. Indicated RCS Average Temperature meter 4  ::=:: 587.2 °F meter 3 < 586.9 °P computer 4  ::=:: 587.7 °P computer 3 < 587.5 °P

2. Indicated Pressurizer Pressure *meter 4 > 2212.3 psig meter 3 2:. 2215.0 psig computer 4 2:. 2209 .1 psig computer 3 2:. 2211.3 psig

3. RCS Total Flow Rate 2:. 388,000 gpm

MCEI-0400-318 Page 27 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 2.13 Spent Fuel Pool Boron Concentration (TS 3.7.14) 2.13.1 Minimum boron concentration limit for the spent fuel pool. Applicable when fuel assemblies are stored in the spent fuel pool.

Parameter Spent fuel pool minimum boron concentration. 2,675 ppm 2.14 Refueling Operations - Boron Concentration (TS 3.9.1) 2.14.1 Minimum boron concentration limit for the filled portions of the Reactor Coolant System, refueling canal, and refueling cavity for MODE 6 conditions. The minimum boron concentration limit and plant refueling procedures ensure that core Keff remains within MODE 6 reactivity requirement ofKeff,:::: 0.95.

Parameter Minimum boron concentration of the Reactor Coolant 2,675 ppm System, the refueling canal, and the refueling cavity.

MCEI-0400-318 Page 28 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 2.15 Borated Water Source-Shutdown (SLC 16.9.14) 2.15.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODE 4 with any RCS cold leg temperature < 300 °F and MODES 5 and 6.

Parameter Note: When cycle burnup is > 476 EFPD, Figure 6 may be used to determine required BAT minimum level.

BAT minimum contained borated water volume 10,599 gallons 13.6% Level BAT minimum boron concentration 7,000 ppm BAT minimum water volume required to 2,300 gallons maintain SDM at 7,000 ppm RWST minimum contained borated water 47,700 gallons volume 41 inches RWST minimum boron concentration 2,675 ppm RWST minimum water volume required to 8,200 gallons maintain SDM at 2,675 ppm

MCEI-0400-318 Page 29 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report 2.16 Borated Water Source~ Operating (SLC 16.9.11) 2.16.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODES l, 2, 3, and MODE 4 with all RCS cold leg temperature > 300 °F.

Parameter Note: When cycle bumup is > 476 EFPD, Figure 6 may be used to determine required BAT minimum level.

BAT minimum contained borated water volume 22,049 gallons 38.0% Level BAT minimum boron concentration 7,000 ppm BAT minimum water volume required to 13,750 gallons maintain SDM at 7,000 ppm RWST minimum contained borated water volume 96,607 gallons 103.6 inches RWST minimum boron concentration 2,675 ppm RWST maximum boron concentration (TS 3.5.4) 2,875 ppm RWST minimum water volume required to 57,107 gallons maintain SDM at 2,675 ppm 2.17

• Standby Shutdown System -(SLC-16.9.7) 2.17.1 Minimum boron concentration limit for the spent fuel pool required for Standby Makeup Pump Water Supply. Applicable for MODES 1, 2, and 3.

Parameter Spent fuel pool minimum boron concentration for TR 2,675 ppm 16.9.7.2.

MCEI-0400-318 Page 30 Revision2 McGuire 2 Cycle 24 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus RCS Boron Concentration (Valid When Cycle Burnup is > 476 EFPD)

This figure includes additional volumes listed in SLC 16.9.14 and 16.9.11 40.0

!I I j I I i I i 35.0 ............

I RCS Boron II Concentration BAT Level

!eem>  !%level) i 0 < 300 37.0 I

30.0 I  ! 300 <500 33.0 ,. . . . .

I i 500 < 700 28.0 700 < 1000 23.0 1000 < 1300 13.6

=:_ 1300 8.7 25.0 *-******* ***********

cu t

,.;i I

~ 20.0 ....................... .......... *****- ***h**** ..********

] I I

,.;i '

~ 15.0 **********

10.0 *************************

I i I* r I 5.0 ...............

I 0.0 I I I 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 RCS Boron Concentration (ppmb)

MCEI-0400-318 Page 31 Revision 2 McGuire 2 Cycle 24 Core Operating Limits Report NOTE: Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance. This data was generated in the McGuire 2 Cycle 24 Maneuvering Analysis calculation file, MCC-1553.05-00-0613. Due to the size of the monitoring factor data, Appendix A is controlled electronically within Duke and is not included in the Duke internal copies of the COLR. The Plant Nuclear Engineering 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 included in the COLR copy transmitted to the NRC.