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#REDIRECT [[CNS-18-027, Submittal of Core Operating Limits Report (COLR) for Cycle 23 Reload Core - Implementation Instructions for Revision 0]]
| number = ML18134A005
| issue date = 05/11/2018
| title = Catawba Nuclear Station, Unit 2, Submittal of Core Operating Limits Report (Colr) for Cycle 23 Reload Core - Implementation Instructions for Revision 0
| author name =
| author affiliation = Duke Energy Carolinas, LLC
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000414
| license number =
| contact person =
| case reference number = CNC-1553.05-00-0666, Rev. 0, CNS-18-027
| document report number = CNEI-0400-332, Rev 0
| package number = ML18134A002
| document type = Fuel Cycle Reload Report
| page count = 31
}}
 
=Text=
{{#Wiki_filter:Catawba 2 Cycle 23 Core Operating Limits Report Revision 0 February 2018
 
==Reference:==
 
CNC-1553.05-00-0666 , Rev. 0 Reload 50.59 #: 02181851 QA CONDITION 1 CNEI-0400-332 Page 1 Revision 0 The information presented in this report has been prepared and issued in accordance with Catawba Technical Specification 5.6.5.
Catawba 2 Cycle 23 Core Operating Limits Report Implementation Instructions for Revision 0 Revision Description and CR Tracking CNEI-0400-332 Page2 Revision 0 Revision O of the Catawba Unit 2 Cycle 23 COLR contains limits specific to the reload core. There is no CR associated with this revision.
Implementation Schedule The Catawba Unit 2 Cycle 23 COLR requires the reload 50.59 (AR #02181851) be approved prior to implementation and fuel loading. Revision O may become effective any time during No MODE between cycles 22 and 23 but must become effective prior to entering MODE 6 which starts cycle 23. The Catawba Unit 2 Cycle 23 COLR will cease to be effective during No MODE between cycle 23 and 24. 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 -2.18. CNS Reactor Engineering performed site inspection in accordance with AD-NF-ALL-0807 and AD-NF-NG0-0214.
Revision 0 CNEI-0400-332 Page3 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report Effective Date February 2018 REVISION LOG Pages Affected 1-31, Appendix A* COLR C2C23 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.
Catawba 2 Cycle 23 Core Operating Limits Report CNEI-0400-332 Page4 Revision 0 1.0 Core Operating Limits Report TS Section 2.1.1 3.1.1 3.1.3 3.1.4 3.1.5 3.1.6 3.1.8 3.2.1 3.2.2 3.2.3 3.3.l 3.3.9 3.4.l 3.5.l 3.5.4 3.7.15 3.9.1 5.6.5 SLC Section 16.7-9 16.9-11 16.9-12 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.
COLR NRC Approved Technical Specifications COLR Parameter Section Methodology (Section I.I Number) Reactor Core Safety Limits RCS Temperature and 2.1 6, 7, 8, 9, 10, 12, 15, Pressure Safety Limits 16, 19,20 Shutdown Margin Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19,20 Moderator Temperature Coefficient MTC 2.3 6, 7, 8, 14, 16, 18 Rod Group Alignment Limits Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19,20 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 Control Bank lnse1iion Limit Shutdown Margin 2.2 2, 4, 6, 7, 8, 9, 10 Rod Insertion Limits 2.5 12, 14, 15, 16, 19, 20 Physics Tests Exceptions Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19,20 Heat Flux Hot Channel Factor FQ 2.6 2, 4, 6, 7, 8, 9, 10, AFD 2.8 12, 15, 16, 19, 20 OT~T 2.9 Penalty Factors 2.6 Nuclear Enthalpy Rise Hot Channel Fm 2.7 2, 4, 6, 7, 8, 9, 10 Factor Penalty Factors 2.7 12, 15, 16, 19, 20 Axial Flux Difference AFD 2.8 2, 4, 6, 7, 8, 15, 16 Reactor Trip System Instrumentation OT~T 2.9 6, 7, 8, 9, 10, 12 OP~T 2.9 15, 16, 19, 20 Boron Dilution Mitigation Sy stem Reactor Makeup Water Flow 2.10 6, 7, 8, 14, 16 Rate RCS Pressure, Temperature and Flow limits RCS Pressure, Temperature 2.11 6, 7, 8, 9, 10, 12, forDNB and Flow 19,20 Accumulators Max and Min Boron Cone. 2.12 6, 7, 8, 14, 16 Refueling Water Storage Tank Max and Min Boron Cone. 2.13 6, 7, 8, 14, 16 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.14 6, 7, 8, 14, 16 Refueling Operations
-Boron Concentration Min Boron Concentration 2.15 6, 7, 8, 14, 16 Core Operating Limits Report (COLR) Analytical Methods 1.1 None The Selected Licensee Commitments that reference this report are listed below COLR NRC Approved Selected Licensing COLR Parameter Section Methodology (Section Commitment I.I Number) Standby Shutdown System Standby Makeup Pump Water Supply 2.16 6, 7, 8, 14, 16 Boration Systems-Borated Water Borated Water Volume and Cone. for 2.17 6, 7, 8, 14, 16 Source -Shutdown BAT/RWST Boration Sy stems-Borated Water Borated Water Volume and Cone. for 2.18 6, 7, 8, 14, 16 Source -Operating BAT/RWST CNEI-0400-332 Page 5 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report 1.1 Analytical Methods Analytical methods used to determine core operating limits for parameters identified in Technical Specifications and previously reviewed and approved by the NRC as specified in Technical Specification 5.6.5 are as follows. 1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," (W Proprietary).
Revision 0 Report Date: July 1985 Not Used for C2C23 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 for C2C23 4. WCAP-12945-P-A, Volume 1 and Volumes 2-5, "Code Qualification Document for Estimate Loss of Coolant Analysis," (W Proprietary).
Revision:
Volume 1 (Revision
: 2) and Volumes 2-5 (Revision
: 1) Report Date: March 1998 5. BAW-10168P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B&W Proprietary). Revision 1 SER Date: January 22, 1991 Revision 2 SER Dates: August 22, 1996 and November 26 , 1996. Revision 3 SER Date: June 15, 1994. Not Used for C2C23 CNEI-0400-332 Page6 Revision 0 Catawba 2 Cycle 23 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 l Report Date: March 2015 8. DPC-NE-3002-A, "UFSAR Chapter 15 System Transient Analysis Methodology".
Revision 4b Report Date: September 2010 9. DPC-NE-2004P-A, "Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-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: October 2016 11. DPC-NE-2008-P A, "Fuel Mechanical Reload Analysis Methodology Using T AC03," (DPC Proprietary).
Revision 0 Report Date: April 1995 Not Used for C2C23 12. DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report," (DPC Proprietary).
Revision 3c Report Date: March 2017 13. DPC-NE-1004-A, ''Nuclear Design Methodology Using CASM0-3/SIMULATE-3P." Revision la Report Date: January 2009 Not Used for C2C23 CNEI-0400-332 Page7 Revision 0 Catawba 2 Cycle 23 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 for Core Operating Limits of Westinghouse Reactors," (DPC Proprietary).
Revision la Report Date: June 2009 16. DPC-NE-1005-PA, "Duke Power Nuclear Design Methodology Using CASM0-4 / SIMULATE-3 MOX", (DPC Proprietary).
Revision 1 Report Date: November 12, 2008 17. BAW-10231P-A, "COPERNIC Fuel Rod Design Computer Code" (Framatome ANP Proprietary)
Revision 1 SER Date: January 14, 2004 Not Used for C2C23 18. DPC-NE-1007-PA, "Conditional Exemption of the EOC MTC Measurement Methodology", (DPC and W Proprietary)
Revision 0 Report Date: April 2015 19. WCAP-12610-P-A, "VANTAGE+
Fuel Assembly Reference Core Report," (W Proprietary).
Revision 0 Report Date: April 1995 20. WCAP-12610-P-A&
CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLOŽ," (W Proprietary).
Revision 0 Report Date: July 2006 Catawba 2 Cycle 23 Core Operating Limits Report 2.0 Operating Limits CNEI-0400-332 Page 8 Revision 0 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% Af(/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% L1K/K in MODE 5. 2.2.3 For TS 3.1.4, SDM shall be greater than or equal to 1.3% Af(/K in MODE 1 and MODE2. 2.2.4 For TS 3.1.5, SDM shall be greater than or equal to 1.3% L1KIK 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% Af(/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% L1K/K in MODE 2 during PHYSICS TESTS. 
,.-., r.I.s 0 '--' bl) f--< r:/1 u Catawba 2 Cycle 23 Core Operating Limits Report Figure 1 Reactor Core Safety Limits Four Loops in Operation CNEI-0400-332 Page9 Revision 0 DO NOT OPERA TE IN THIS AREA 660 1---~~~~-1--~~~~-1--~~~----+
~~------1 650 630 620 610 600 ACCEPT ABLE OPERATION 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power CNEI-0400-3 3 2 Page 10 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report 2.3 ModeratorTemperature 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 Af(/K/&deg;F.
EOC, ARO, RTP MTC shall be less negative than the -4.3E-04 ~K/K/&deg;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/&deg;F. 2.3.3 The Revised Predicted near-EOC 300ppm ARO RTP MTC shall be calculated using the procedure contained in DPC-NE-1007-P A. 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/&deg;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-0400-332 Page 11 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 0.9 -Unacceptable Operation
= 0.8 .... !.C ..... 0.7 0 ~, 0.6 0.5 i.': <I "" s"~ 0.4 =: Acceptable Operation lo.,.... o-0.3 -i.': 0.2 "O 0 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.
Catawba 2 Cycle 23 Core Operating Limits Report Figure 3 CNEI-0400-3 3 2 Page 12 Revision 0 Control Bank Insertion Limits Versus Percent Rated Thermal Power Full y Withdrawn (Maximum =231~ f}J .... ---------,_ -/' /' /' ,, /' -I 200 /' /' ,__ i 180 ts: I,,, ] 160 -140 !. Full y Withdra wn /' /' -(Minimum= 222) -/' Contro!BankB
-/' /' / (1 00%, 1 6 1) F /' /' (0%, 1 63) /' /' " -/ / /' /' / ---120 ControlBankC V /' /' ,, /' C /' /' 0 /' /' 100 /' ;' /' /' "' 0 80 C /' /' /' /' 0 .:: --C ontrolBankD
/' ;' I,,, 60 /' /' ;' "' .s 'C 40 0 20 /' /' (0%, 4 7) /' /' /' / -]full y In s ert e /' /' (30%, 0) /' 0 -0 10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power The Rod In s ertion L imit s (RI L) for Control Bank D (C D), Control B a nk C (C C), a nd C ontrol Bank B (C B) can b e calculat e d by: Bank C D RIL = 2.3(P)-69 {30 < P < 100) Bank CC RIL = 2.3(P) +4 7 {O:::. P < 7 6.1) for CC RIL = 22 2 {7 6.1 < P < 100) Bank CB RIL = 2.3(P) + 163 {O < P < 25. 7) fo r C B RIL = 2 2 2 {2 5. 7 < P < JOO) wh e re P = % of Rat e d Th e rmal P o w e r NOTE: Compli a nc e with T e chnical Sp e cificati o n 3.1.3 may r e quir e rod w ithdra wa l limit s. Refer to the Unit 2 ROD manual for details.
CNEI-0400-332 Page 13 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report Table 1 Control Bank Withdrawal Steps and Sequence Fully Withdrawn at 222 Steps Fully Withdrawn at 223 Steps Control Control Control Control Control Control Control Control 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 0 222 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 BankB BankC BankD Bank A BankB 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 BankC BankD Bank A Bank B BankC BankD 0 Start 0 0 0 0 Start 0 0 0 116 o*start 0 0 116 0 Start 0 0 226 Stop 110 0 0 227 Stop Ill 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 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 2 28 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 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 2 30 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 Catawba 2 Cycle 23 Core Operating Limits Report 2.6 Heat Flux Hot Channel Factor -FQ(X,Y,Z) (TS 3.2.1) CNEI-0400-332 Page 14 Revision 0 2.6.1 FQ(X,Y ,Z) steady-state limits are defined by the following relationships:
where, F RTP *K(Z)/P Q F ~TP *K(Z)/0.5 Thermal Power p = Rated Thermal Power for P > 0.5 for P < 0.5 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 ~TI' = 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 i rP with the K(BU) penalty for Westinghouse RFA fuel is analytically confirmed in 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: F 0 (X Y ,Z)
* M (XY Z) 2 6 5 [F L(XY,Z)]OP
= Q ' Q ' ' . . Q ' UMT
* MT
* TILT where: [ Ft (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. Ft (X,Y ,Z)0 P includes allowances for calculation and measurement uncertainties.
F f (X,Y ,Z) = Design power distribution for F Q. F f (X , Y ,Z) is provided in Appendix Table A-1 for normal operatin g conditions and in Appendix Table A-4 for power escalation testing during initial startup operation.
CNEI-0400-332 Page 15 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report MQ(X,Y :Z) = Margin remaining in core location X,Y ;z to the LOCA limit m 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) L RPS_ F~(X,Y,Z)
* Mc(X,Y,Z) 2*6*6 [FQ(X,Y :Z)] -UMT *MT* TILT where: [F~(X,Y,Z)]RPS
= Cycle dependent maximum allowable design peaking factor that ensures FQ(X,Y :Z) Centerline Fuel Melt (CFM) limit is not exceeded for operation within AFD, RIL, and QPTR limits. [F~(X,Y :Z)]RPS includes allowances for calculation and measurement unce1iainties.
D FQ(X,Y :Z) = Defmed 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 = Defmed in Section 2.6.5. MT = Defmed in Section 2.6.5. TILT = Defmed in Section 2.6.5.
Catawba 2 Cycle 23 Core Operating Limits Report 2.6.7 KSLOPE= 0.0725 where: CNEI-0400-332 Page 16 Revision 0 KSLOPE= adjustment to Ki value from OT~T trip setpoint required to compensate for each 1 % measured FQM (X,Y :Z) RPS exceeds [ Ft (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.
1.200 Catawba 2 Cycle 23 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for RFA Fuel (0.0, 1.00) (4.0, 1.00) CNEI-0400-332 Page 17 Revision 0 1.000 ----------
(12.0, 0.9259) (4.0, 0.9259) 0.800 8 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 ----1* -~--~---~----~---~~---~
0.0 2.0 4.0 6.0 8.0 10.0 12.0 Core Height (ft)
Catawba 2 Cycle 23 Core Operating Limits Report Table 2 FQ(X,Y,Z) and FLrn(X,Y)
Penalty Factors For Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2 Bumup FQ(X,Y,Z)
F L rn(X,Y) CNEI-0400-332 Page 18 Revision 0 (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 2.00 492 2.00 2.00 497 2.00 2.00 507 2.00 2.00 517 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 F 1 rn(X,Y) for compliance with Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.
CNEI-0400-332 Page 19 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report 2.7 Nuclear Enthalpy Rise Hot Channel Factor-F LiH(X,Y) (TS 3.2.2) F ~H steady-state limits referred to in Technical Specification 3.2.2 are defined by the following relationship.
where: [F~8 (X, Y)tco 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, F: (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. L S U R V 2. 7.2 [ F , rn (X , Y)] = FiH (X, Y)
* M L'.H (X , Y) UMR *TILT where: L SURV _ [FL'iH (X , Y)] -Cycle dependent maximum allowable design peaking factor that ensures F L'iiX,Y) limit is not exceeded for operation SURV
* l d within AFD RIL and QPTR limits FL (X Y) me u e s ' ' . L'.H ' allowances for calculation and measurement uncertainty. F~H (X , Y) = Design power distribution for F L'iH* 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.
CNEI-0400-332 Page20 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report Mt.iX,Y) == Margin remaining in core location X,Y relative to the Operational DNB limits in the transient power distribution.
M~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.
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 Mt.tt<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 OT~ T K I setpoint required to compensate for each 1 % measured radial peak, F~ (X,Y) exceeds its limit. 2. 7 .5 F t.H(X,Y) Penalty Factors for Teclmical Specification Surveillance 3.2.2.2 ar e 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.
Core Height (ft) 1.05 0.12 1.8092 1.20 1.8102 2.40 1.8093 3.60 1.8098 4.80 1.8097 6.00 1.8097 7.20 1.8070 8.40 1.8073 9.60 1.8072 10.80 1.7980 ll.40 I. 7892 Catawba 2 Cycle 23 Core Operating Limits Report Table 3 1.1 1.8553 1.8540 1.8525 1.8514 1.8514 1.8514 1.8438 1.8319 1.8102 1.7868 1.7652 Maximum Allowable Radial Peaks (MARPs) 1.2 1.3 1.9248 1.9146 1.9248 1.9146 1.9312 1.9146 1.9204 1.9146 1.9058 1.9146 1.8921 1.9212 1.8716 1.8930 1.8452 1.8571 1.8093 1.7913 I. 7611 1.7163 1.7250 1.6645 RF A Fuel MARPs 100% Full Power Axial Peak 1.4 1.5 1.6 1.7 1.9179 2.0621 2.0498 2.0090 1.9179 2.1073 2.0191 1.9775 1.9179 2.0735 1.9953 1.9519 1.9179 2.0495 1.9656 1.9258 1.9179 2.0059 1.9441 1.9233 1.9179 1.9336 1.8798 1.8625 1.8872 1.8723 1.8094 1.7866 1.8156 1.7950 1.7359 1.7089 1.7375 I. 7182 1.6572 1.6347 1.6538 1.6315 1.5743 1.5573 1.6057 1.5826 1.5289 1.5098 1.8 1.9 1.9333 1.8625 1.9009 1.8306 1.8760 1.8054 1.8524 1.7855 1.8538 1.7836 1.8024 1.7472 1.7332 1.6812 1.6544 1.6010 1.5808 1.5301 1.5088 1.4624 1.4637 1.4218 CNEI-0400-3 3 2 Page 21 Revision 0 2.1 3 3.25 1.7780 1.3151 1.2461 1.7852 1.3007 1.2235 I. 7320 1.4633 1.4616 1.6996 1.4675 1.3874 1.6714 1.2987 1.2579 1.6705 1.3293 1.2602 1.5982 1.2871 1.2195 1.5127 1.2182 1.1578 1.4444 1.1431 1.0914 1.3832 1.1009 1.0470 1.3458 1.0670 1.0142 CNEI-0400-332 Page22 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits (-20, 100) Unacceptable Operation 90 80 70 bl 60 . Accepta e Opera on (-36 , 50) 40 20 -10 50 40 30 20 10 0 Unacceptable Operation
(+21 , 50) 10 20 30 40 Axi a l Flux Difference
(% Delt a I) 50 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.
CNEl-0400-3 3 2 Page 23 Revision 0 Catawba 2 Cycle 23 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 Tavg at RTP Nominal RCS Operating Pressure Overtemperature 6 T reactor trip setpoint Overtemperature 6 T reactor trip heatup setpoint penalty coefficient Overtemperature 6 T reactor trip depressurization setpoint penalty coefficient Time constants utilized in the lead-lag compensator for6T Time constant utilized in the lag compensator for 6 T Time constants utilized in the lead-lag compensator for Tavg Time constant utilized in the measured Tavg lag compensator f1 (61) "positive" breakpoint f1 (61) "negative" breakpoint f 1 (M) "positive" slope f 1 (61) "negative" slope Nominal Value T' < 590.8 &deg;F P' = 2235 psig K1 = 1.1953 K2 = 0.031631&deg;F K3 = 0.001414lpsi 11=8sec. 12=3sec. 13=0sec. 14 = 22 sec. 15 = 4 sec. 16=0sec. =3.0 %61 =NIA' = 1.525 %6Tol%61 =NIA' f1(M) negative breakpoints and slopes for0TL'1Tare less restrictive than OPL'1Tfi(M)negativebreakpoint and slope. Therefore, during a transient which challenges negative imbalance limits, OPL'1T fz(M) limits will result in a reactor trip before OTL'1T f1(M) limits are reached. This makes implementation of an OTL'1Tf 1 (M) negative breakpoint ands lope unnecessary.
CNEI-0400-332 Page 24 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report 2.9.2 Overpower
~T Setpoint Parameter Values Parameter Nominal Tavg at RTP Overpower T reactor trip setpoint Overpower T reactor trip penalty Overpower T reactor trip heatup setpoint penalty coefficient Time constants utilized in the lead-lag compensator for ~T Time constant utilized in the lag compensator for ~T Tin1e constant utilized in the measured Tavg lag compensator Time constant utilized in the rate-lag controller for Tavg fi(~I) "positive" breakpoint fi(~I) "negative" breakpoint fi( ~I) "pas itive" slope fi(~I) "negative" slope Nominal Value T" < 590.8 &deg;F Ks= 0.02 / &deg;F for increasing Tavg Ks= 0.00 / &deg;F for decreasing Tavg K 6 = 0.001291/0F for T > T" K6 = 0.0 /&deg;F for T,::: T" 1 1 = 8 sec. 1 2 = 3 sec. 1 3 = 0 sec. 1 6 = 0 sec. 1 7 = 10 sec. = 35.0 %~I =-35.0 %~I = 7.0 %~Tof %~I Catawba 2 Cycle 23 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 MODE3 MODE4or 5 Limit :=: 80 gpm :=: 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) CNEI-0400-332 Page25 Revision 0 2.12.1 Boron concentration limits during MODES 1 and 2, and MODE 3 with RCS pressure > 1000 psi: Parameter Applicable Burnup Limit Accumulator minimum boron concentration.
0-200 EFPD 2,500 ppm Accumulator minimum boron concentration.
200.1 -250 EFPD 2,500 ppm Accumulator minimum boron concentration.
250.1 -300 EFPD 2,447 ppm Accumulator minimum boron concentration.
300.1 -350 EFPD 2,336 ppm Accumulator minimum boron concentration.
350.1 -400 EFPD 2,256 ppm Accumulator minimum boron concentration.
400.1 -450 EFPD 2,183 ppm Accumulator minimum boron concentration.
450.1 -475 EFPD 2,114 ppm Accumulator minimum boron concentration.
475.1 -507 EFPD 2,079 ppm Accumulator minimum boron concentration.
507.1-517 EFPD 2,031 ppm Accumulator maximum boron concentration.
0-517 EFPD 3,075 ppm Catawba 2 Cycle 23 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters No. Operable PARAMETER INDICATION CHANNELS 1. Indicated RCS Average Temperature meter 4 meter 3 computer 4 computer 3 2. Indicated Pressurizer Pressure meter 4 meter 3 computer 4 computer 3 3. RCS Total Flow Rate CNEI-0400-332 Page26 Revision 0 LIMITS :S 589.6 &deg;F :S 589.3 &deg;F :S 590.1 &deg;F :S 589.9 &deg;F 2209.8 psig 2212. l psig 2205.8 psig 2207.5 psig 390,000 gpm Catawba 2 Cycle 23 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 R WST minimum boron concentration.
R WST maximum boron concentration.
2.14 Spent Fuel Pool Boron Concentration (TS 3.7.15) CNEI-0400-332 Page 27 Revision 0 Limit 2,700 ppm 3,075 ppm 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 Keffremains within MODE 6 reactivity requirement ofKeff:::
0.95. Parameter Minimum boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity. Limit 2,700 ppm CNEI-0400-3 32 Page 28 Revision 0 Catawba 2 Cycle 23 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 Spent fuel pool minimum boron concentration for TR 16.7-9.3.
2.17 Borated Water Source -Shutdown (SLC 16.9-11) Limit 2,700 ppm 2.17.1 Volume and boron concentrations for the Boric Acid Tank(BAT) and the Refueling Water Storage Tank (RWST) during MODE 4 with any RCS cold leg temperature.:::
210&deg;F, and MODES 5 and 6. Parameter BAT minimum boron concentration Volume of 7,000 ppm boric acid solution required to maintain SDM at 68 &deg;F Limit 7,000 ppm 2000 gallons NOTE: When cycle burnup is> 446 EFPD, Figure 6 may be used to determine required BAT minimum level. BAT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-11) R WST minin1um boron concentration Volume of 2,700 ppm boric acid solution required to maintain SDM at 68 &deg;F RWST Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-11) 13,086 gallons (14.9% level) 2,700 ppm 7,000 gallons 48,500 gallons (8.7% level)
CNEI-0400-332 Page 29 Revision 0 Catawba 2 Cycle 23 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 &deg;F *.
* NOTE: The SLC 16.9-12 applicability is down to MODE 4 temperatures of > 210&deg;F. The minimum volumes calculated support cooldown to 200&deg;F to satisfy UFSAR Chapter 9 requirements.
Parameter BAT minimum boron concentration Volume of 7,000 ppm boric acid solution required to maintain SDM at 210&deg;F Limit 7,000 ppm 13 , 500 gallons NOTE: When cycle burnup is> 446 EFPD, Figure 6 may be used to determine required BAT minimum level. BAT Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-12) R WST minimum boron concentration Volume of 2,700 ppm boric acid solution required to maintain SDM at 210 &deg;F RWST Minimum Shutdown Volume (Includes the additional volumes listed in SLC 16.9-12) 25,200 gallons (45.8% level) 2,700 ppm 57,107 gallons 98,607 gallons (22.0% level) 50.0 45.0 40.0 35.0 -l 30.0 QI ...J ::-25.0 QI ...J 1-20.0 a!i 15.0 10.0 5.0 0.0 CNEI-0400-332 Page 30 Revision 0 Catawba 2 Cycle 23 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is> 446 EFPD) This figure includes additional volumes listed in SLC 16.9-11 and 16.9-12 0 RCS Boron Concentration BAT Level I (ppm) (%level) I 0 < 300 43.0 300 < 500 40.0 500 < 700 37.0 700 < 1000 30.0 1000 < 1300 14.9 1300 < 2700 9.8 > 2700 9.8 Unacceptable Operation Acceptable Operation 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 Primary Coolant Boron Concentration (ppmb)
Catawba 2 Cycle 23 Core Operating Limits Report Appendix A Power Distribution Monitoring Factors CNEI-0400-332 Page 31 Revision 0 Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance.
This data was generated in the Catawba 2 Cycle 23 Maneuvering Analysis calculation file, CNC-1553.05-00-0660.
Due to the size of monitoring factor data, Appendix A is controlled electronically within Duke and is not included in Duke internal copies of the COLR. Catawba Reactor and Electrical Systems Engineering controls monitoring factor via computer files and should be contacted if there is a need to access this information.
Appendix A is included in the COLR transmitted to the NRC.}}

Latest revision as of 08:33, 18 April 2019

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