RA-23-0084, Core Operating Limits Report (COLR) for Unit 1 Cycle 27 Reload Core, Rev. 4
| ML23075A216 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/29/2023 |
| From: | Flippin N Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RA-23-0084 | |
| Download: ML23075A216 (1) | |
Text
Nicole Flippin d_~ DUKE Vice President
~ ENERGY Catawba Nuclear Station Duke Energy CN01VP I 4800 Concord Road York, SC 29745 o: 803.701 .3340 f : 803.701 .3221 Nicole.Flippin@duke-energy.com RA-23-0084 March 29, 2023 ATTN : Document Control Desk U.S. Nuclear Regulatory Commission Washington , DC 20555-0001
Subject:
Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Unit 1 Facility Operating License Number NPF-35 Docket Number 50-413 Core Operating Limits Report (COLR) for Unit 1 Cycle 27 Reload Core, Rev. 4 Pursuant to Catawba Technical Specification 5.6.5.d, please find attached an information copy of the subject revised (Revision 4) COLR.
The power distribution monitoring factors from Appendix A of Revision 3 remain valid and are not transmitted as part of Revision 4.
This letter and enclosed COLR do not contain any regulatory commitments .
Please direct any questions or concerns to Ari Tuckman , Regulatory Affairs, at (803) 701-3771 .
Sincerely, Nicole Flippin Vice President, Catawba Nuclear Station Enclosure (Catawba Unit 1 Cycle 27 COLR , Rev. 4) www.duke-energy.com
U.S. Nuclear Regulatory Commission RA-23-0084 March 29, 2023 Page 12 xc (with enclosure):
L. Dudes, Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257 Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station S. Williams, Project Manager U.S. Nuclear Regulatory Commission 11555 Rockville Pike Mailstop O-881A Rockville, MD 20852
U.S. Nuclear Regulatory Commission RA-23-0084 March 29, 2023 Page 13 bxc (with enclosure):
ELL-EC2ZF bxc (without enclosure):
M.B. Hare W. Murphy NCMPA-1 NCEMC PMPA
Enclosure Catawba Unit 1 Cycle 27 COLR, Rev. 4 (excluding Attachment A)
CNEI-0400-3 79 Page 1 Revision 4 Catawba 1 Cycle 27 Core Operating Limits Report Revision 4 March 2023
Reference:
CNC-1553.05-00-0711, Rev. 4 Reload 50.59 #02403601 QA Condition 1 The information presented in this report has been prepared and issued in accordance with Catawba Technical Specification 5.6.5.
CNEI-0400-379 Page2 Revision 4 Catawba 1 Cycle 27 Core Operating Limits Report Implementation Instructions for Revision 4 Revision Description and CR Tracking Revision 4 of the Catawba 1 Cycle 27 COLR contains limits applicable to the cycle 27 core. The following changes are made in this revision.
- COLR Reference 18 is being revised to allow for the adoption of the revised EOC MTC test exemption criteria in Revision 1 ofDPC-NE-1007-P-A. (COLR Section 1.1)
- The rod repositioning schedule was updated from Revision 7 to 9 to reflect current guidance.
- Tau3 and tau6 associated with the OT.AT and OP.AT trip function and K4, tau3 and tau6 associated with the OP.AT trip function are revised to reduce the potential for spurious trip signals (COLR Sections 9.2.1 and 9.2.2). The OP.AT f2(.AI) breakpoints are revised to increase analysis margin.
Power distribution monitoring factors contained in Appendix A of Revision 3 remain valid and are not transmitted as part of this revision.
Implementation Schedule Implementation of Revision 4 to the Catawba 1 Cycle 27 COLR may be performed any time upon receipt.
The Catawba 1 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 A CDR was performed by Safety Analysis for COLR Section 2.9.
CNS Reactor Engineering performed site inspection in accordance with AD-NF-ALL-0807 and AD-NF-NGO-0214.
CNEI-0400-379 Page 3 Revision 4 Catawba 1 Cycle 27 Core Operating Limits Report REVISION LOG Revision Effective Date Pages Affected COLR 0 September 2021 1-31, Appendix A* C1C27 COLR, Rev. 0 1 October 2021 1-31 C1C27 COLR, Rev. 1 2 October 2021 1-31 C1C27 COLR, Rev. 2 3 October 2021 1-31, Appendix A* C1C27 COLR, Rev. 3 4 March 2023 1-3, 7, 13, 23 and 24 CIC27 COLR, Rev. 4
- Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance and is not uploaded as part of the EI body. However, Appendix A is uploaded into the document management system, for ease of transmittal to the NRC.
CNEI-0400-379 Page4 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 1.0 Core Operating Limits Report This Core Operating Limits Report (COLR) has been prepared in accordance with requirements of Technical Specification 5.6.5. Technical Specifications that reference this report are listed below along with the NRC approved analytical methods used to develop and/or determine COLR parameters identified in Technical Specifications.
TS COLR NRC Approved Section Technical Specifications COLR Parameter Section Methodology (Section 1.1 Number) 2.1.1 Reactor Core Safety Limits RCS Temperature and Pressure 2.1 6, 7, 8, 9, 10, 12, 15, Safety Limits 16, 19,20 3.1.1 Shutdown Margin Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19, 20 3.1.3 Moderator Temperature Coefficient MTC 2.3 6, 7, 8, 14, 16, 18 3.1.4 Rod Group Ali!mment Limits Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19, 20 3.1.5 Shutdown Bank Insertion Limit Shutdown Margin 2.2 2, 4, 6, 7, 8, 9, 10 Rod Insertion Limits 2.4 12, 14, 15, 16, 19, 20 3.1.6 Control Bank Insertion Limit Shutdown Margin 2.2 2, 4, 6, 7, 8, 9, 10 Rod Insertion Limits 2.5 12, 14, 15, 16, 19, 20 3.1.8 Physics Tests Exceptions Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16, 19, 20 3.2.1 Heat Flux Hot Channel Factor FQ 2.6 2, 4, 6, 7, 8, 9, 10, AFD 2.8 12, 15, 16, 19, 20 OTLlT 2.9 Penalty Factors 2.6 l 3.2.2 Nuclear Enthalpy Rise Hot Channel FLlH 2.7 2, 4, 6, 7, 8, 9, 10 Factor Penaltv Factors 2.7 12, 15, 16, 19, 20 3.2.3 Axial Flux Difference AFD 2.8 2, 4, 6, 7, 8, 15, 16 3.3.1 Reactor Trip System Instrumentation OTLlT 2.9 6, 7, 8, 9, 10, 12 OPLlT 2.9 15, 16, 19, 20 3.3.9 Boron Dilution Mitigation System Reactor Makeup Water Flow Rate 2.10 6, 7, 8, 14, 16 3.4.1 RCS Pressure, Temperature and Flow RCS Pressure, Temperature and 2.11 6, 7, 8, 9, 10, 12, limits for DNB Flow 19,20 3.5.1 Accumulators Max and Min Boron Cone. 2.12 6, 7, 8, 14, 16 3.5.4 Refueling Water Storage Tank Max and Min Boron Cone. 2.13 6, 7, 8, 14, 16 3.7.15 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.14 6, 7, 8, 14, 16 3.9.1 Refueling Operations - Boron Min Boron Concentration 2.15 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 SLC COLR NRC Approved Section Selected Licensee Commitment COLR Parameter Section Methodology (Section 1.1 Number) 16.7-9 Standby Shutdown System Standby Makeup Pump Water SuoPl).:'. 2.16 6, 7, 8, 14, 16 16.9-11 Boration Systems - Borated Water Borated Water Volume and Cone. for 2.17 6, 7, 8, 14, 16 Source - Shutdown BAT/RWST 16.9-12 Boration Systems - Borated Water Borated Water Volume and Cone. for 2.18 6, 7, 8, 14, 16 Source - Ooerating BAT/RWST
CNEI-0400-379 PageS Revision 3 Catawba 1 Cycle 27 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," CJ!.. Proprietary).
Revision 0 Report Date: July 1985 Not Used
- 2. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code," CJ!.. 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," CJ!.. 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," CJ!.. 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 Revision2 SER Dates: August 22, 1996 and November 26, 1996 Revision 3 SER Date: June 15, 1994 Not Used
CNEI-0400-3 79 Page6 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 1.1 Analytical Methods (continued)
- 6. DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," (Duke Energy Proprietary).
Revision Sa Report Date: October 2012
- 7. DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (Duke Energy Proprietary).
Revision 1 Report Date: March 2015
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," (Duke Energy Proprietary).
Revision 2a Report Date: December 2008
- 10. DPC-NE-2005P-A, "Thermal Hydraulic Statistical Core Design Methodology," (Duke Energy Proprietary).
Revision 6 Report Date: September 2020
- 11. DPC-NE-2008P-A, "Fuel Mechanical Reload Analysis Methodology Using TACO3," (Duke Energy Proprietary).
Revision 0 Report Date: April 1995 Not Used
- 12. DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report," (Duke Energy Proprietary).
Revision 3c Report Date: March 2017
- 13. DPC-NE-1004-A, "Nuclear Design Methodology Using CASMO-3/SIMULATE-3P."
Revision la Report Date: January 2009 Not Used
CNEI-0400-379 Page7 Revision 4 Catawba 1 Cycle 27 Core Operating Limits Report 1.1 Analytical Methods (continued)
- 14. DPC-NF-2010-A, "Nuclear Physics Methodology for Reload Design."
Revision 2a Report Date: December 2009
- 15. DPC-NE-2011-PA, "Nuclear Design Methodology Report for Core Operating Limits of Westinghouse Reactors," (Duke Energy Proprietary).
Revision la Report Date: June 2009
- 16. DPC-NE-1005-PA, "Nuclear Design Methodology Using CASMO-4 / SIMULATE-3 MOX,"
(Duke Energy Proprietary).
Revision 1 Report Date: November 2008
- 17. BAW-10231P-A, "COPERNIC Fuel Rod Design Computer Code" (Framatome ANP Proprietary)
Revision 1 SER Date: January 14, 2004 Not Used
- 18. DPC-NE-1007-P-A, "Conditional Exemption of the EOC MTC Measurement Methodology,"
(Duke Energy and W Proprietary)
Revision 1 Report Date: December 2022
- 19. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report," QY Proprietary).
Revision 0 Report Date: April 1995
- 20. WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO'," QY Proprietary).
Revision 0 Report Date: July 2006
CNEI-0400-379 Page8 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 2.0 Operating Limits Cycle-specific parameter limits for specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using NRC approved methodologies specified in Section 1.1.
2.1 Reactor Core Safety Limits (TS 2.1.1)
Reactor Core Safety Limits are shown in Figure 1.
2.2 Shutdown Margin - SDM (TS 3.1.1, TS 3.1.4, TS 3.1.5, TS 3.1.6, TS 3.1.8) 2.2.1 For TS 3.1.1, SDM shall be 2'.: 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 2'.: 1.0% Af</K in MODE 5.
2.2.3 For TS 3.1.4, SDM shall be 2'.: 1.3% Af</K in MODE 1 and MODE 2.
2.2.4 For TS 3.1.5, SDM shall be 2'.: 1.3% LiK/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 2'.: 1.3% Af</K in MODE 1 and MODE 2 with Keff2'.:
1.0.
2.2.6 For TS 3.1.8, SDM shall be 2'.: 1.3% Af</K in MODE 2 during PHYSICS TESTS.
CNEI-0400-3 79 Page 9 Revision 3 Catawba 1 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 t _530 C')
~
Cl) 620 0
a:::
610 600 590 ACCEPTABLE OPERATION 580 ..__ ___,___ _ __.__ _ _..,___ ___,___ _ __,__ ____,
0.0 0 .2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power
CNEI-0400-379 Page 10 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 2.3 Moderator Temperature Coefficient- MTC (TS 3.1.3) 2.3.1 Moderator Temperature Coefficient (MTC) Limits are:
MTC shall be less positive than the upper limits shown in Figure 2.
BOC, ARO, HZP MTC shall be less positive than 0.7E-04 dK/K/°F.
EOC, ARO, RTP MTC shall be less negative than the -4.3E-04 dK/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 dK/K/°F.
2.3.3 The Revised Predicted near-EOC 300 ppm ARO RTP MTC shall be calculated using the procedure contained in DPC-NE-1007-PA.
If the Revised Predicted MTC is less negative than or equal to the 300 ppm SR 3.1.3.2 Surveillance Limit, and all benchmark data contained in the surveillance procedure is satisfied, then an MTC measurement in accordance with SR 3 .1.3 .2 is not required to be performed.
2.3.4 60 ppm MTC Surveillance Limit is:
Measured 60 ppm ARO, equilibrium RTP MTC shall be less negative than or equal to -4.125E-04 dK/K/°F.
Where: BOC = Beginning of Cycle (burnup corresponding to 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.
CNEI-0400-379 Page 11 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Figure2 Moderator Temperature Coefficient Upper Limit Versus Power Level 1.0 - , - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ,
Unacceptable Operation Acceptable Operation 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 1 ROD manual for details.
CNEI-0400-3 79 Page 12 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Figure3 Control Bank Insertion Limits Versus Percent Rated Thermal Power F II W"thd u y 1 raw~-
(Maximum= 231) 231 -
220 - - - - ./
- - - - ~
,,.- - - /
., ./ /
., /
200 ./
Fully Withdrawn V
/
i' 1so
./
(Minimum=222)
/ v : Contro!Bank B : ./
./
,a F C'il ./
/
c100%, 161)
! R (0%, 163) - -
./
160
~ ./
./
., ./
!. 140 /
./ ./
Contro!Bank C ,,
~
./ ./
= 120 I I
./ /
V /
0
- = ./ ./
-~ 100 ./
./
., ./
/ ./
ll.
]
= 80
./
1./
./
./ ./
ControlBank D
"' V ./
.5 -
Qj 60 ./ ./
.s-= ,,
./ ./
40 ~ (0%,47)
/
""" I
/
/
Fully Inserted ./
20
_,, (30%,0)
./
0 - 1./
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:
BankCDRIL =2.3(P)-69 {30::::P::::100)
Bank CC RIL = 2.3(P) + 47 {0 ::::P:::: 76.1) for CC RIL = 222 (76.1 < P:::: 100)
Bank CB RIL = 2.3 (P) + 163 (0 :::: P :::: 25. 7) for CB RIL = 222 (25. 7 < P :::: 100}
where P = % of Rated Thermal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.
Refer to the Unit 1 ROD manual for details.
CNEI-0400-379 Page 13 Revision 4 Catawba 1 Cycle 27 Core Operating Limits Report Table 1 Control Bank Withdrawal Sequence Equation Control Control Control Control Bank A Bank B Bank C Bank D O Start 0 0 0 116 0 Start 0 0 CBA Stop CBA- 116 0 0 CBA 116 O Start 0 CBA CBB Stop CBB -116 0 CBA CBB 116 0 Start CBA CBB CBC Stop CBC - 116 Where:
CBA = Fully withdrawn position of Control Bank A CBB = Fully withdrawn position of Control Bank B CBC= Fully withdrawn position of Control Bank C Allowed Control Bank Fully Withdrawn Positions Range from 223 Steps to 231 Steps for frequent RCCA Reposition Required per CNEI-0400-091, Rev. 9, "RCCA Axial Repositioning Schedule for Catawba Nuclear Station."
CNEI-0400-379 Page 14 Revision 3 Catawba 1 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:
Fr*K(Z)/P forP > 0.5 F ~TP *K(Z)/0.5 forP:::; 0.5 where, p = Thermal Power Rated Thermal Power Note: Measured FQ(X,Y,Z) shall be increased by 3% to account for manufacturing tolerances and 5% to account for measurement uncertainty when comparing against the LCO limits. The manufacturing tolerance and measurement uncertainty are implicitly included in the F Q surveillance limits as defined in COLR Sections 2.6.5 and 2.6.6.
RTP 2.6.2 F Q = 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 bumup. 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 bumups.
The following parameters are required for core monitoring per the Surveillance Requirements of Technical Specification 3 .2.1:
L Fg(X,Y,Z)
- M0(X,Y,Z) 2.6.5 [FQ(X,Y,Z)]OP = UMT *MT* TILT where:
[FJ (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. FJ (X, Y,Z)°p includes allowances for calculation and measurement uncertainties.
Ft (X,Y,Z) = Design power distribution for FQ. Ft (X,Y,Z) is provided in Appendix Table A-1 for normal operating conditions and in
CNEI-0400-379 Page 15 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 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)
L RPS Fg(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 uncertainties.
D FQ(X,Y,Z) = Defined in Section 2.6.5.
Mc(X,Y,Z) = Margin remaining to the CFM limit in core location X,Y,Z from the transient power distribution. Mc(X,Y,Z) is provided in Appendix Table A-2 for normal operating conditions and in Appendix Table A-5 for power escalation testing during initial startup operations.
UMT = Defined in Section 2.6.5.
MT = Defined in Section 2.6.5.
TILT = Defined in Section 2.6.5.
CNEI-0400-379 Page 16 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 2.6.7 KSLOPE = 0.0725 where:
KSLOPE = Adjustment to K 1 value from OT~T trip setpoint required to M FL compensate for each 1% FQ (X,Y,Z) exceeds Q (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.
CNEI-0400-3 79 Page 17 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height forRFAFuel 1.200 ~ - - - - - - - - - - - - - - - - - - - - - - - - - - - ,
(0.0, 1.00) 1.000 _ _ _ _ _ _ _(4.0, __ 1.00)
.,.)
(12.0, 0.9259)
(4.0, 0.9259) 0.800
@0.600
~
0.400 Core Height
{ft} K(Z) 0.0 1.000 0.200 ,:S4 1.000
>4 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)
CNEI-0400-379 Page 18 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Table 2 FQ(X,Y,Z) and FaH(X,Y) Penalty Factors For Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2 Burnup FQ(X,Y,Z) FaH(X,Y)
(EFPD) Penalty Factor(%) Penalty Factor (%)
4 2.00 2.00 12 2.00 2.00 25 2.56 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 2.00 482 2.00 2.00 486 2.00 2.00 503 2.00 2.00 513 2.00 2.00 532 2.00 2.00 Note: Linear interpolation is adequate for intermediate cycle burnups.
All cycle bumups outside the range of the table shall use a 2%
penalty factor for both FQ(X,Y,Z) and FaH{X,Y) for compliance with Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.
CNEI-0400-3 79 Page 19 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 2.7 Nuclear Enthalpy Rise Hot Channel Factor - Fm(X,Y) (TS 3.2.2)
FMI steady-state limits referred to in Technical Specification 3.2.2 are defined by the following relationship.
2.7.1 [F,k()C, Y)]Lco= MARP (X,Y) * [ 1.0 + RiH * (1.0 - P)]
where:
[F,k oc, 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% measured radial peak, F~ (X,Y), exceeds the limit.
(RRH = 3.34, 0.0 < P :S 1.0)
The following parameters are required for core monitoring per surveillance requirements of Technical Specification 3.2.2.
2.7.2 [ Flii (X,Y)J8URV = F~ (X, Y) *ML\H (X, Y)
UMR *TILT where:
[ Flii (X,Y)J8URV = Cycle dependent maximum allowable design peaking factor that ensures FMI(X, Y) limit is not exceeded for operation within AFD, RIL, and QPTR limits. Flii (X,Y)SURV includes allowances for calculation and measurement uncertainty.
F~H (X, Y) = Design power distribution for FAH. 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-0400-379 Page 20 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report M,m(X, Y) = Margin remaining in core location X, Y relative to the Operational DNB limits in the transient power distribution.
Mm(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 MliiX, 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 1 setpoint required to compensate for each 1% that the measured radial peak, F~ (X, Y) exceeds its limit.
2.7.5 FiiH(X,Y) Penalty Factors for Technical Specification Surveillance 3.2.2.2 are provided in Table 2.
2.8 Axial Flux Difference - AFD (TS 3.2.3) 2.8.1 Axial Flux Difference (AFD) Limits are provided in Figure 5.
CNEI-0400-379 Page 21 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Table3 Maximum Allowable Radial Peaks (MARPs)
RFA Steady State Limiting Value Between Loss of Flow Accident (LOFA) MARPs and FAHwcA Core Axial Peak Height ft 1.05 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 2.1 3 3.25 0.12 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.3151 1.2461 1.20 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.3007 1.2235 2.40 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.4633 1.4616 3.60 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.4675 1.3874 4.80 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.2987 1.2579 6.00 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.3293 1.2602 7.20 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.5982 1.2871 1.2195 8.40 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6010 1.5127 1.2182 1.1578 9.60 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.5808 1.5301 1.4444 1.1431 1.0914 10.80 1.6058 1.6058 1.6058 1.6058 1.6058 1.6058 1.5743 1.5573 1.5088 1.4624 1.3832 1.1009 1.0470 11.40 1.6058 1.6058 1.6058 1.6058 1.6057 1.5826 1.5289 1.5098 1.4637 1.4218 1.3458 1.0670 1.0142
CNEI-0400-379 Page 22 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits
(-20, 100) (+10, 100)
Unacceptable Operation Unacceptable Operation 90 80
~
0 70 r:i..
-a 5
..c:
Acceptable Operation 60 E-<
'"0 50 tU (-36, 50) (+21, 50)
~
..... 40
=
0 tU g 30 tU r:i..
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 the Unit 1 ROD manual for operational AFD limits.
CNEI-0400-379 Page 23 Revision 4 Catawba 1 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 dT Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T' :S 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 = 0.033401°F penalty coefficient Overtemperature ~T reactor trip depressurization K3 = 0.00160llpsi setpoint penalty coefficient Time constants utilized in the lead-lag compensator 1:1 = 8 sec.
for /)._T 1:2 = 3 sec.
Time constant utilized in the lag compensator for /)._T 1:3 ::S 1.8 sec.
Time constants utilized in the lead-lag compensator 1:4 = 22 sec.
for Tavg 1:5 = 4 sec.
Time constant utilized in the measured Tavg lag *6 ::S 1.8 sec.
compensator f1 (M) "positive" breakpoint = 19.0 %~I f 1(M) "negative" breakpoint =NIA*
f1 (M) "positive" slope = 1.769 %!)._Toi %M f 1(!)._I) "negative" slope =NIA*
- f1 {Af) negative breakpoints and slopes for OT.e1T are less restrictive than OP,1T fz{Af) negative breakpoint and slope. Therefore, during a transient which challenges negative imbalance limits, OPdT fz(Af) limits will result in a reactor trip before OTdT f1 {Af) limits are reached. This makes implementation of an OTdT f1 {Af) negative breakpoint and slope unnecessary.
CNEI-0400-379 Page 24 Revision 4 Catawba 1 Cycle 27 Core Operating Limits Report 2.9.2 Overpower AT Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP T" ~ 585.1 °P Overpower AT reactor trip setpoint K4~ 1.0909 Overpower AT reactor trip penalty Ks= 0.02 I 0 P for increasing Tavg Ks= 0.00 I 0 P for decreasing Tavg Overpower AT reactor trip heatup K6 = 0.001179/0 P for T > T" setpoint penalty coefficient K6 = 0.0 / 0 P forT ~T" Time constants utilized in the lead-lag 't 1 = 8 sec.
compensator for AT 't 2 = 3 sec.
Time constant utilized in the lag 't3 ~ 1.8 sec.
compensator for AT Time constant utilized in the measured 't 6 ~ 1.8 sec.
T avg lag compensator Time constant utilized in the rate-lag controller for T avg fz(AI) "positive" breakpoint (I) 27.0 %Af ~ (fz(Af) "positive")~ 35.0 %Af fz(Af) "negative" breakpoint(!) -27.0 %Af 2'.: (fz(Af) "negative") 2'.: -35.0 %Af fz(AI) "positive" slope = 7.0 %.ATof %Af fz(Af) "negative" slope = 7.0 %ATof %Af (1) Variable f2(~I) breakpoints are specified to allow a phased implementation of the more restrictive
+/- 27 %M limits. The safety analysis for the CIC27 core was analyzed for the+/- 35 %M breakpoints which bounds the transition to the more restrictive +/- 27 %M limits which will be implemented in future core designs.
CNEI-0400-379 Page 25 Revision 3 Catawba 1 Cycle 27 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 < 80 gpm MODE4 or 5 < 70 gpm 2.11 RCS Pressure, Temperature, and Flow Limits for DNB (TS 3.4.1)
RCS pressure, temperature and flow limits for DNB are shown in Table 4.
2.12 Accumulators (TS 3.5.1) 2.12.1 Boron concentration limits during MODES 1 and 2, and MODE 3 with RCS pressure > 1000 psi:
Parameter Applicable Bumup Limit Accumulator minimum boron concentration. 0 - 200 EFPD 2,500 ppm Accumulator minimum boron concentration. 200.1 - 300 EFPD 2,500 ppm Accumulator minimum boron concentration. 300.1 - 400 EFPD 2,297 ppm Accumulator minimum boron concentration. 400.1 - 513 EFPD 2,145 ppm Accumulator minimum boron concentration 513.1 - 532 EFPD 1,972 ppm Accumulator maximum boron concentration. 0- 532 EFPD 3,075 ppm
CNEI-0400-379 Page 26 Revision 3 Catawba 1 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 :::; 587.5 °F meter 3 :::; 587.3 °F computer 4 :::; 588.0 °F computer 3 :::; 587.9 °F
- 2. Indicated Pressurizer Pressure meter 4 2'.: 2206.9 psig meter 3 2'.: 2208. 7 psig computer 4 2'.: 2204.0 psig computer 3 2'.: 2205.4 psig
- 3. RCS Total Flow Rate > 384,000 gpm
CNEI-0400-379 Page 27 Revision 3 Catawba 1 Cycle 27 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 RWST minimum boron concentration. 2,700 ppm R WST maximum boron concentration. 3,075 ppm 2.14 Spent Fuel Pool Boron Concentration (TS 3.7.15) 2.14.1 Minimum boron concentration limit for the spent fuel pool. Applicable when fuel assemblies are stored in the spent fuel pool.
Parameter Spent fuel pool minimum boron concentration. 2,700 ppm 2.15 Refueling Operations - Boron Concentration (TS 3.9.1) 2.15.1 Minimum boron concentration limit for filled portions of the Reactor Coolant System, refueling canal, and refueling cavity for MODE 6 conditions. The minimum boron concentration limit and plant refueling procedures ensure that core Keff remains within MODE 6 reactivity requirement ofKeff:'.5 0.95.
Parameter Minimum boron concentration of the Reactor Coolant 2,700 ppm System, the refueling canal, and the refueling cavity.
CNEI-0400-379 Page 28 Revision 3 Catawba 1 Cycle 27 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.
2.17 Boration Systems Borated Water Source- Shutdown (SLC 16.9-11) 2.17.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODE 4 with any RCS cold leg temperature::; 210 °P, and MODES 5 and 6.
Parameter BAT minimum boron concentration 7,000 ppm Volume of 7,000 ppm boric acid solution required 2,000 gallons to maintain SDM at 68°P NOTE: When cycle burnup is > 445 EFPD, Figure 6 may be used to determine the required BAT minimum level.
BAT Minimum Shutdown Volume (Includes the 13,086 gallons additional volumes listed in SLC 16.9-11) (14.9%)
R WST minimum boron concentration 2,700 ppm Volume of2,700 ppm boric acid solution required 7,000 gallons to maintain SDM at 68 °P R WST Minimum Shutdown Volume (Includes the 48,500 gallons additional volumes listed in SLC 16.9-11) (8.7%)
CNEI-0400-379 Page 29 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report 2.18 Boration Systems Borated Water Source - Operating (SLC 16.9-12) 2.18.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODES 1, 2, and 3 and MODE 4 with all RCS cold leg temperatures> 210 °F*.
- NOTE: The SLC 16.9-12 applicability is down to MODE 4 temperatures of
> 210°F. The minimum volumes calculated support cooldown to 200°F to satisfy UFSAR Chapter 9 requirements.
Parameter BAT minimum boron concentration 7,000 ppm Volume of7,000 ppm boric acid solution required 13,500 gallons to maintain SDM at 210 °F NOTE: When cycle burnup is> 445 EFPD, Figure 6 may be used to determine the required BAT minimum level.
BAT Minimum Shutdown Volume (Includes the 25,200 gallons additional volumes listed in SLC 16.9-12) (45.8%)
R WST minimum boron concentration 2,700 ppm Volume of2,700 ppm boric acid solution required 57,107 gallons to maintain SDM at 210 °F RWST Minimum Shutdown Volume (Includes the 98,607 gallons additional volumes listed in SLC 16.9-12) (22.0%)
CNEI-0400-379 Page 30 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is> 445 EFPD)
This figure includes additional volumes listed in SLC 16.9-11 and 16.9-12 50.0 I
I I
RCS Boron 45.0 Concentration BAT Level 40.0 I (ppm) 0 < 300
(%level) 43.0 I 300 < 500 40.0 35.0 - 500 < 700 37.0
~ 30.0 Q) !
700 < 1000 1000 < 1300 1300 < 2700 30.0 14.9 9.8
...J
";!. > 2700 9.8 i 25.0 I
~ Unacceptable
...J 20.0 Operation -
Acceptable Operation ID 15.0 I
I I I
10.0 5.0 0.0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 Primary Coolant Boron Concentration (ppmb)
CNEI-0400-379 Page 31 Revision 3 Catawba 1 Cycle 27 Core Operating Limits Report Appendix A Power Distribution Monitoring Factors Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance. This data was generated in the Catawba 1 Cycle 27 Maneuvering Analysis calculation file, CNC-1553.05-00-0705, Rev 1. Due to the size of the monitoring factor data, Appendix A is controlled electronically within the Duke document management system and is not included in the Duke internal copies of the COLR. The Plant Reactor Engineering section will control this information via computer file(s) and should be contacted ifthere is a need to access this information.
Appendix A is available to be transmitted to the NRC.
Filename Cksum I File Size clc27colrei_r3_AppendixA.pdf 2371384185 / 1514470