ML20125A197
| ML20125A197 | |
| Person / Time | |
|---|---|
| Site: | Mcguire |
| Issue date: | 03/31/2020 |
| From: | Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20125A195 | List: |
| References | |
| NTM 02245502, RA-20-0121 MCEI-0400-395, Rev 0 | |
| Download: ML20125A197 (32) | |
Text
Facility Code : MC Applicable Facilities :
Document Number : MCEI-0400-395 Document Revision Number : 000 Document EC Number :
Change Reason : NTM 02245502 Document Title : McGuire 2 Cycle 27 Core Operating Limits Report Thompson, Chad A Originator 3/5/2020 Klein, Casey Concurrent Originator 3/5/2020 Soliman, Kristie M QA 3/5/2020 Siry, Steve Safety Analysis Verifier 3/5/2020 Elkins, Jason R Site Impact Review 3/5/2020 Robinson, Duncan Approver 3/5/2020 Notes :
MCEI-0400-395 Page 1 Revision 0 McGuire Unit 2 Cycle 27 Core Operating Limits Report Revision 0 March 2020 Calculation Number: MCC-1553.05-00-0681, Revision 0 Reload 50.59 # 02307736 QA Condition 1 The information presented in this report has been prepared and issued in accordance with McGuire Technical Specification 5.6.5.
MCEI-0400-395 Page 2 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report Implementation Instructions for Revision 0 Revision Description and CR Tracking Revision 0 of the McGuire Unit 2 Cycle 27 COLR contains limits specific to the reload core.
There is no CR associated with this revision.
Implementation Schedule The McGuire Unit 2 Cycle 27 COLR requires the reload 50.59 (AR #02307736) be approved prior to implementation and fuel loading.
Revision 0 may become effective any time during NO MODE between cycles 26 and 27, but must become effective prior to entering MODE 6 which starts cycle 27. The McGuire Unit 2 Cycle 27 COLR will cease to be effective during No MODE between cycles 27 and 28.
Data Files to be Implemented No data files are transmitted as part of this document.
Additional Information CDR was performed by Safety Analysis for COLR Sections 1.1, 2.1, and 2.9 - 2.17.
MNS Reactor Engineering performed site inspection in accordance with AD-NF-ALL-0807 and AD-NF-NGO-0214.
MCEI-0400-395 Page 3 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected COLR 0 March 2020 1-31, Appendix A* M2C27 COLR, Rev. 0
- 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-0400-395 Page 4 Revision 0 McGuire 2 Cycle 27 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 6,7,8,9,10,12,15,16,18, Pressure Safety Limits 19 3.1.1 Shutdown Margin Shutdown Margin 2.2 6,7,8,12,14,15,16,18,19 3.1.3 Moderator Temperature Coefficient MTC 2.3 6,7,8, 14,16, 17 3.1.4 Rod Group Alignment Limits Shutdown Margin 2.2 6,7,8,12,14,15,16,18,19 3.1.5 Shutdown Bank Insertion Limits Shutdown Margin 2.2 6,7,8,12,14,15,16,18,19 3.1.5 Shutdown Bank Insertion Limits Shutdown Bank Insertion 2.4 2,4,6,7,8,9,10,12,14,15, Limit 16,18,19 3.1.6 Control Bank Insertion Limits Shutdown Margin 2.2 6,7,8,12,14,15,16,18,19 3.1.6 Control Bank Insertion Limits Control Bank Insertion 2.5 2,4,6,7,8,9,10,12,14,15, Limit 16,18,19 3.1.8 Physics Tests Exceptions Shutdown Margin 2.2 6,7,8,12,14,15,16,18,19 3.2.1 Heat Flux Hot Channel Factor Fq, AFD, OTT and 2.6 2,4,6,7,8,9,10,12,15,16, Penalty Factors 18,19 3.2.2 Nuclear Enthalpy Rise Hot Channel FH, AFD and 2.7 2,4,6,7,8,9,10,12,15,16, Factor Penalty Factors 18,19 3.2.3 Axial Flux Difference AFD 2.8 2,4,6,7,8,15,16 3.3.1 Reactor Trip System Instrumentation OTT and OPT 2.9 6,7,8,9,10,12,15,16,18, Setpoints Constants 19 3.4.1 RCS Pressure, Temperature, and Flow RCS Pressure, 2.10 6,7,8,9,10,12,18,19 DNB limits Temperature and Flow 3.5.1 Accumulators Max and Min Boron Conc. 2.11 6,7,8,14,16 3.5.4 Refueling Water Storage Tank Max and Min Boron Conc. 2.12 6,7,8,14,16 3.7.14 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.13 6,7,8,14,16 3.9.1 Refueling Operations - Boron Min Boron Concentration 2.14 6,7,8,14,16 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 6,7,8,14,16 Conc. for BAT/RWST 16.9.11 Borated Water Source - Operating Borated Water Volume and 2.16 6,7,8,14,16 Conc. for BAT/RWST 16.9.7 Standby Shutdown System Standby Makeup Pump Water 2.17 6,7,8,14,16 Supply
MCEI-0400-395 Page 5 Revision 0 McGuire 2 Cycle 27 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.
- 1. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, (W Proprietary).
Revision 0 Report Date: July 1985 Not Used
- 2. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," (W Proprietary).
Revision 0 Report Date: August 1985 Addendum 2, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, (W Proprietary). (Referenced in Duke Letter DPC-06-101)
Revision 1 July 1997
- 3. WCAP-10266-P-A, The 1981 Version Of Westinghouse Evaluation Model Using BASH Code, (W Proprietary).
Revision 2 Report Date: March 1987 Not Used
- 4. WCAP-12945-P-A, Volume 1 and Volumes 2-5, Code Qualification Document for Best-Estimate Loss of Coolant Analysis, (W Proprietary).
Revision: Volume 1 (Revision 2) and Volumes 2-5 (Revision 1)
Report Date: March 1998
- 5. BAW-10168P-A, B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (B&W Proprietary).
Revision 1 SER Date: January 22, 1991 Revision 2 SER Dates: August 22, 1996 and November 26, 1996 Revision 3 SER Date: June 15, 1994 Not Used
MCEI-0400-395 Page 6 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report 1.1 Analytical Methods (continued)
Revision 5a Report Date: October 2012
- 7. DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary).
Revision 1 Report Date: March 2015
Revision 4c Report Date: February 2019
- 9. DPC-NE-2004P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," (DPC Proprietary).
Revision 2a Report Date: December 2008
- 10. DPC-NE-2005P-A, "Thermal Hydraulic Statistical Core Design Methodology," (DPC Proprietary).
Revision 5 Report Date: March 2016
- 11. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3," (DPC Proprietary).
Revision 0 Report Date: April 1995 Not Used
- 12. DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report, (DPC Proprietary).
Revision 3c Report Date: March 2017
- 13. DPC-NE-1004A, "Nuclear Design Methodology Using CASMO-3/SIMULATE-3P."
Revision 1a Report Date: January 2009 Not Used
MCEI-0400-395 Page 7 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report 1.1 Analytical Methods (continued)
- 14. DPC-NF-2010-A, "Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design."
Revision 2a Report Date: December 2009
- 15. DPC-NE-2011-PA, "Duke Power Company Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," (DPC Proprietary).
Revision 1a Report Date: June 2009
- 16. DPC-NE-1005-PA, "Nuclear Design Methodology Using CASMO-4 / SIMULATE-3 MOX,"
(DPC Proprietary).
Revision 1 Report Date: November 2008
- 17. DPC-NE-1007-PA, "Conditional Exemption of the EOC MTC Measurement Methodology, (DPC and W Proprietary)
Revision 0 Report Date: April 2015
- 18. WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, (W Proprietary).
Revision 0 Report Date: April 1995
- 19. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, (W Proprietary).
Revision 0 Report Date: July 2006
MCEI-0400-395 Page 8 Revision 0 McGuire 2 Cycle 27 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 1.
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% K/K 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% K/K in MODE 5.
2.2.3 For TS 3.1.4, SDM shall be > 1.3% K/K in MODES 1 and 2.
2.2.4 For TS 3.1.5, SDM shall be > 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 > 1.3% K/K in MODE 1 and MODE 2 with K-eff > 1.0.
2.2.6 For TS 3.1.8, SDM shall be > 1.3% K/K in MODE 2 during PHYSICS TESTS.
MCEI-0400-395 Page 9 Revision 0 McGuire 2 Cycle 27 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
MCEI-0400-395 Page 10 Revision 0 McGuire 2 Cycle 27 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 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 a MTC measurement in accordance with SR 3.1.3.2 is not required to be performed.
2.3.4 60 PPM MTC Surveillance Limit is:
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 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-0400-395 Page 11 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 0.9 Unacceptable Operation 0.8 Moderator Temperature Coefficient 0.7 0.6 (1.0E-04 K/K/F) 0.5 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 OP/2/A/6100/22 Unit 2 Data Book for details.
MCEI-0400-395 Page 12 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power Fully Withdrawn (Maximum = 231) 231 220 200 Fully Withdrawn Rod Insertion Position (Steps Withdrawn)
(Minimum = 222) 180 Control Bank B (100%, 161) 160 (0%, 163) 140 Control Bank C 120 100 80 Control Bank D 60 40 (0%, 47) 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 < 100}
Bank CC RIL = 2.3(P) +47 {0 < P < 76.1} for CC RIL = 222 {76.1 < P < 100}
Bank CB RIL = 2.3(P) +163 {0 < P < 25.7} for CB RIL = 222 {25.7 < P < 100}
where P = %Rated Thermal Power 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-0400-395 Page 13 Revision 0 McGuire 2 Cycle 27 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 Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 222 Stop 106 0 0 223 Stop 107 0 0 222 116 0 Start 0 223 116 0 Start 0 222 222 Stop 106 0 223 223 Stop 107 0 222 222 116 0 Start 223 223 116 0 Start 222 222 222 Stop 106 223 223 223 Stop 107 Fully Withdrawn at 224 Steps Fully Withdrawn at 225 Steps Control Control Control Control Control Control Control Control Bank A Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 224 Stop 108 0 0 225 Stop 109 0 0 224 116 0 Start 0 225 116 0 Start 0 224 224 Stop 108 0 225 225 Stop 109 0 224 224 116 0 Start 225 225 116 0 Start 224 224 224 Stop 108 225 225 225 Stop 109 Fully Withdrawn at 226 Steps Fully Withdrawn at 227 Steps Control Control Control Control Control Control Control Control Bank A Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 226 Stop 110 0 0 227 Stop 111 0 0 226 116 0 Start 0 227 116 0 Start 0 226 226 Stop 110 0 227 227 Stop 111 0 226 226 116 0 Start 227 227 116 0 Start 226 226 226 Stop 110 227 227 227 Stop 111 Fully Withdrawn at 228 Steps Fully Withdrawn at 229 Steps Control Control Control Control Control Control Control Control Bank A Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 228 Stop 112 0 0 229 Stop 113 0 0 228 116 0 Start 0 229 116 0 Start 0 228 228 Stop 112 0 229 229 Stop 113 0 228 228 116 0 Start 229 229 116 0 Start 228 228 228 Stop 112 229 229 229 Stop 113 Fully Withdrawn at 230 Steps Fully Withdrawn at 231 Steps Control Control Control Control Control Control Control Control Bank A Bank B Bank C Bank D Bank A Bank B Bank C Bank D 0 Start 0 0 0 0 Start 0 0 0 116 0 Start 0 0 116 0 Start 0 0 230 Stop 114 0 0 231 Stop 115 0 0 230 116 0 Start 0 231 116 0 Start 0 230 230 Stop 114 0 231 231 Stop 115 0 230 230 116 0 Start 231 231 116 0 Start 230 230 230 Stop 114 231 231 231 Stop 115
MCEI-0400-395 Page 14 Revision 0 McGuire 2 Cycle 27 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, 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 QRTP = 2.70 x K(BU) 2.6.3 K(Z) is the normalized FQ(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 of burnup. F QRTP 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:
D L FQ(X,Y,Z)
- MQ(X,Y,Z) 2.6.5 FQ(X,Y,Z)OP = UMT
- TILT
MCEI-0400-395 Page 15 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report where:
FQL (X,Y,Z)OP = Cycle dependent maximum allowable design peaking factor that ensures FQ(X,Y,Z) LOCA limit will be preserved for operation within the LCO limits. FQL (X,Y,Z)OP 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.
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
- TILT where:
L FQ(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.
L FQ(X,Y,Z)RPS includes allowances for calculation and measurement uncertainties.
D FQ(X,Y,Z) = Defined in Section 2.6.5.
MCEI-0400-395 Page 16 Revision 0 McGuire 2 Cycle 27 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 K1 value from the OTT trip setpoint required to RPS compensate for each 1% that FQM (X,Y,Z) 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.
MCEI-0400-395 Page 17 Revision 0 McGuire 2 Cycle 27 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, 1.00) (4.0, 1.00) 1.000 (12.0, 0.9259)
(4.0, 0.9259) 0.800 K(Z) 0.600 0.400 Core Height (ft) K(Z) 0.0 1.0 0.200 4.0 1.0
> 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-0400-395 Page 18 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report Table 2 FQ(X,Y,Z) and FH(X,Y) Penalty Factors For Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burnup 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 2.27 2.28 125 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 2.00 485 2.00 2.00 496 2.00 2.00 499 2.00 2.00 504 2.00 2.00 514 2.00 2.00 524 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 FH(X,Y) for compliance with the Technical Specification Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.
MCEI-0400-395 Page 19 Revision 0 McGuire 2 Cycle 27 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 is 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% that the measured radial peak, FMH (X,Y), exceeds its limit.
(RRH = 3.34 (0.0 < P < 1.0))
The following parameters are required for core monitoring per the surveillance requirements of Technical Specification 3.2.2.
SURV FDH (X, Y) x M H (X, Y) 2.7.2 FLH (X,Y) =
UMR x TILT where:
SURV FLH (X,Y) = Cycle dependent maximum allowable design peaking factor that ensures the FH(X,Y) limit will be preserved for operation SURV within the LCO limits. F L (X,Y) includes allowances for H
calculation/measurement uncertainty.
MCEI-0400-395 Page 20 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report D D FH (X,Y) = Design radial 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.
MH(X,Y) = The 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 the OTT K1 setpoint required to compensate for each 1%
that the 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 The Axial Flux Difference (AFD) Limits are provided in Figure 5.
MCEI-0400-395 Page 21 Revision 0 McGuire 2 Cycle 27 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
MCEI-0400-395 Page 22 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits
(-18, 100) (+10, 100) 100 90 Unacceptable Operation Unacceptable Operation 80 Percent of Rated Thermal Power Acceptable Operation 70 60 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 OP/2/A/6100/22 Unit 2 Data Book for more details.
MCEI-0400-395 Page 23 Revision 0 McGuire 2 Cycle 27 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 Value Nominal Tavg at RTP T < 585.1°F Nominal RCS Operating Pressure P = 2235 psig Overtemperature T reactor trip setpoint K1 < 1.1978 Overtemperature T reactor trip heatup setpoint K2 = 0.0334/oF penalty coefficient Overtemperature T reactor trip depressurization K3 = 0.001601/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 < 2 sec.
Time constants utilized in the lead-lag compensator 4 > 28 sec.
for Tavg 5 < 4 sec.
Time constant utilized in the measured Tavg lag 6 < 2 sec.
compensator f1(I) "positive" breakpoint = 19.0 %I f1(I) "negative" breakpoint = N/A*
f1(I) "positive" slope = 1.769 %T0/ %I f1(I) "negative" slope = N/A*
- The f1(I) "negative" breakpoint and the f1(I) "negative" slope are less restrictive than the OPT f2(I) negative breakpoint and slope. Therefore, during a transient which challenges the negative imbalance limits, the OPT f2(I) limits will result in a reactor trip before the OTT f1(I) limits are reached. This makes implementation of the OTT f1(I) negative breakpoint and slope unnecessary.
MCEI-0400-395 Page 24 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report 2.9.2 Overpower T Setpoint Parameter Values Parameter Value Nominal Tavg at RTP T < 585.1°F Overpower T reactor trip setpoint K4 < 1.0864 Overpower T reactor trip Penalty K5 = 0.02/°F for increasing Tavg K5 = 0.0 for decreasing Tavg Overpower T 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 T 2 < 3 sec.
Time constant utilized in the lag 3 < 2 sec.
compensator for T Time constant utilized in the 6 < 2 sec.
measured Tavg lag compensator Time constant utilized in the rate-lag 7 > 5 sec.
controller 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
MCEI-0400-395 Page 25 Revision 0 McGuire 2 Cycle 27 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 - 300 EFPD 2,475 ppm concentration.
Accumulator minimum boron 300.1 - 400 EFPD 2,353 ppm concentration.
Accumulator minimum boron 400.1 - 514 EFPD 2,181 ppm concentration.
Accumulator minimum boron 514.1 - 524 EFPD 2,015 ppm concentration.
Accumulator maximum boron 0 - 524 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 Limit RWST minimum boron concentration. 2,675 ppm RWST maximum boron concentration. 2,875 ppm
MCEI-0400-395 Page 26 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters No. Operable Parameter Indication Channels Limits
- 1. Indicated RCS Average Temperature meter 4 < 587.2 ºF meter 3 < 586.9 ºF computer 4 < 587.7 ºF computer 3 < 587.5 ºF
- 2. Indicated Pressurizer Pressure meter 4 > 2212.3 psig meter 3 > 2215.0 psig computer 4 > 2209.1 psig computer 3 > 2211.3 psig
- 3. RCS Total Flow Rate > 388,000 gpm
MCEI-0400-395 Page 27 Revision 0 McGuire 2 Cycle 27 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 Limit 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 of Keff < 0.95.
Parameter Limit Minimum boron concentration of the Reactor Coolant 2,675 ppm System, the refueling canal, and the refueling cavity.
MCEI-0400-395 Page 28 Revision 0 McGuire 2 Cycle 27 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 Limit Note: When cycle burnup is > 469 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 7,150 ppm BAT minimum water volume required to 2,300 gallons maintain SDM at 7,150 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-0400-395 Page 29 Revision 0 McGuire 2 Cycle 27 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 1, 2, 3, and MODE 4 with all RCS cold leg temperature > 300 °F.*
- Note: The SLC 16.9.11 applicability is down to Mode 4 temperatures of > 300°F. The minimum volumes calculated support cooldown to 200°F to satisfy UFSAR Chapter 9 requirements.
Parameter Limit Note: When cycle burnup is > 469 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 7,150 ppm BAT minimum water volume required to 13,750 gallons maintain SDM at 7,150 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 Limit Spent fuel pool minimum boron concentration for TR 2,675 ppm 16.9.7.2.
MCEI-0400-395 Page 30 Revision 0 McGuire 2 Cycle 27 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus RCS Boron Concentration (Valid When Cycle Burnup is > 469 EFPD)
This figure includes additional volumes listed in SLC 16.9.14 and 16.9.11 40.0 RCS Boron 35.0 Concentration BAT Level (ppm) (%level) 0 < 300 37.0 300 < 500 33.0 30.0 500 < 700 28.0 700 < 1000 23.0 1000 < 1300 13.6 25.0
> 1300 8.7 BAT Level (% Level) 20.0 Acceptable 15.0 10.0 Unacceptable Operation 5.0 0.0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 2800 RCS Boron Concentration (ppmb)
MCEI-0400-395 Page 31 Revision 0 McGuire 2 Cycle 27 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 27 Maneuvering Analysis calculation file, MCC-1553.05-00-0676. Due to the size of the monitoring factor data, Appendix A is controlled electronically within the Duke document management system and is not included in the Duke internal copies of the COLR. The Plant Reactor Engineering and Support Systems section will control this information via computer file(s) and should be contacted if there is a need to access this information.
Appendix A is included in the COLR copy transmitted to the NRC.