ML15093A421
| ML15093A421 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/31/2015 |
| From: | Henderson K, Young J Duke Energy Carolinas, Duke Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15093A419 | List: |
| References | |
| CNS-15-033 CNEI-0400-281 | |
| Download: ML15093A421 (34) | |
Text
Kelvin Henderson DUKE Vice President
- ENERGY, Catawba Nuclear Station
/*
Duke Energy CNOIVP I 4800 Concord Road York, SC 29745 o: 803.701.4251 f: 803.701.3221 CNS-1 5-033 March 31, 2015 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
Subject:
Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Unit 2 Docket Number 50-414 Core Operating Limits Report (COLR) for Cycle 21 Reload Core Pursuant to Catawba Technical Specification 5.6.5d., please find attached an information copy and an electronic copy of the subject COLR. This COLR is being submitted to update the limits of the Unit 2 Cycle 21 reload core.
The electronic copy of this COLR is included with this letter in compact disc (CD) format. The electronic copy includes the power distribution monitoring factors.
This letter, the attached COLR, and the included CD do not contain any regulatory commitments.
Please direct any questions or concerns to L.J. Rudy at (803) 701-3084.
Very truly yours, Kelvin Henderson Vice President, Catawba Nuclear Station LJR/s Attachments (paper and CD COLR versions)
Aoo(
A£1UL www.duke-energy.com
U.S. Nuclear Regulatory Commission Page 2 March 31, 2015 xc (with attachments):
V.M. McCree, Region II Administrator U.S. Nuclear Regulatory Commission Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto, III, NRC Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station G.E. Miller, NRC Project Manager U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Mail Stop 8-G9A 11555 Rockville Pike Rockville, MD 20852-2738
Attachments Catawba Unit 2 Cycle 21 COLR (paper and CD COLR versions)
CNEI-0400-281 Page 1 Revision 0 Catawba Unit 2 Cycle 21 Core Operating Limits Report Revision 0 February 2015 Calculation Number: CNC-1553.05-00-0621 Duke Energy Date Prepared By:
Checked By:
Checked By:
Approved By:
J.S. Young a
N.R. Hager 4j.2,
~m T.P. Phelps (Sections 1.1, 2.1 and 2.9 - 2.18)
M. A. Blom Z212-1-01 2/I/"
QA Condition I The information presented in this report has been prepared and issued in accordance with Catawba Technical Specification 5.6.5.
CNEI-0400-281 Page 2 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Implementation Instructions for Revision 0 Revision Description and PIP Tracking Revision 0 of the Catawba Unit 2 Cycle 21 COLR contains limits specific to the reload core.
There is no PIP associated with this revision.
Implementation Schedule The Catawba Unit 2 Cycle 21 COLR requires the reload 50.59 be approved prior to implementation and fuel loading.
Revision 0 may become effective any time during No MODE between cycles 20 and 21 but must become effective prior to entering MODE 6 which starts cycle 21. The Catawba Unit 2 Cycle 21 COLR will cease to be effective during No MODE between cycle 21 and 22.
Data files to be Implemented No data files are transmitted as part of this document.
Engineering Instruction Inspection Waiver Per EDM-130 "Engineering Drawings", the Engineering Instruction (EI) has been waived per Reference "CN -1438.88".
CNEI-0400-28 1 Page 3 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report REVISION LOG Revision 0
Effective Date February 2015 Pages Affected 1-3 1, Appendix A*
- 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.
CNEI-0400-281 Page 4 Revision 0 Catawba 2 Cycle 21 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
[
........*....S a ~l' t Y..!*.i m it s 1
3.1.1 Shutdown Margin Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16 3.1.3 Mo.....d.erator..
Tem peratur.e
.Coefficient.......... M TC 2
6...............................3.
8,..2, 14, 16 3.1.4.. Rod Group Alignment Limits Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16.....
I...........
3.1.5 Shutdown Bank Insertion Limit Shutdown Margin 2.2 2, 4,6, 7,8, 9, Rod Insertion Limits 2.4 10, 12, 14, 15, 16 3.1.6 Control Bank Insertion Limit Shutdown Margin 2.2 2,4,6, 7,8, 9, Rod Insertion Limits 2.5 10, 12, 14, 15, 16 3.1.8 Physics Tests Exceptions Shutdown Margin 2.2 6, 7, 8, 12, 14, 15, 16 3.2.1 Heat Flux Hot Channel Factor FQ 2.6 2, 4,6, 7, 8, 9, 10, AFD 2.8 12, 15, 16 OTAT 2.9 Penalty Factors 2.6 3.2.2 Nuclear Enthalpy Rise Hot Channel FAHI-2.7 2, 4, 6, 7, 8, 9, Factor Penalty Factors 2.7 10, 12, 15, 16
- '3".2.3
~
~
~
..*ii i x ~ f e e c.......................................
A F
......... 2...2
,4,6,7,8, 15.1 3J.2.3 Axial Fl ux Difference AFD 2".8 2, 4, 6, 7, 8, 15, 16 3..1
..e c o T rip te
.n t u e t t o
.O...................A..T...................
I..........
- -~ 1
- - -, 2,.........
3.3.1 Reactor Trip System Instrumentation OTAT
- 296, 7, 8, 9, 10, 12, OPAT 2.9 15, 16 3.3.9 Boron Dilution Mitigation System Reactor Makeup Water Flow Rate 2.10 6, 7, 8, 12, 14, 16 3.4.1 RCS Pressure, Temperature and Flow RCS Pressure, Temperature and Flow 2.11 6, 7, 8, 9, 10, 12
..!m t f D...
i...........
3.5.1 Accumulators
.Max and Min Boron Conc.
2.12 6, 7, 8,12, 14, 16
..3 =.................
.cu t l t r M a. a n d inB.ro
...n c..
...... 2 2
, 8.2, !, !.....
3.5.4 Refueling Water Storage Tank.
_ Max and Min Boron Conc.
2.13 6,7, 8, 12, 14, 16 3.7.15 Spent Fuel Pool Boron Concentration Min Boron Concentration 2.14 6, 7, 8, 12, 14, 16
... i...............................
3.9.1 Refueling Operations - Boron Mm Boron Concentration 2.15 6, 7,8, 12, 14, 16 Concentration C r r t in
.L.m p...
a.ti a
.e t -~ ~
1.............
5.6.5 Core Operating Limits Report Analytical Methods 1.1 None (COLR)
The Selected License Commitments that reference this report are listed below SLC COLR NRC Approved Section Selected Licensing Commitment COLR Parameter Section Methodology (Section 1.1 Number) 16.7-9 Standby Shutdown System Standby Makeup Pump Water Supply 2.16 6, 7, 8, 12, 14, 16 16.9-11 Boration Systems - Borated Water Borated Water Volume and Conc. for 2.17 6, 7, 8, 12, 14, 16
.Source - Shutdown BAT/RWST 16.9-12 Boration Systems - Borated Water Borated Water Volume and Conc. for 2.18 6,7,8,12, 14,16 Source - Operating BAT/RWST
A0 CNEI-0400-281 Page 5 Revision 0 Catawba 2 Cycle 21 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," (WY Proprietary).
Revision 0 Report Date: July 1985 Not Used for C2C21
- 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 I 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 C2C21
- 4. WCAP-12945-P-A, Volume I 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 I 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 C2C21
CNEI-0400-281 Page 6 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 1.1 Analytical Methods (continued)
- 6.
DPC-NE-3000-PA, "Thermal-Hydraulic Transient Analysis Methodology," (DPC Proprietary).
Revision 5a Report Date: October 2012
- 7. DPC-NE-3001-PA, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," (DPC Proprietary).
Revision Oa Report Date: May 2009
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-0 1," (DPC Proprietary).
Revision 2a Report Date: December 2008
- 10. DPC-NE-2005P-A, "Thermal Hydraulic Statistical Core Design Methodology," (DPC Proprietary).
Revision 4a Report Date: December 2008
- 11. DPC-NE-2008-PA, "Fuel Mechanical Reload Analysis Methodology Using TACO3 and GDTACO," (DPC Proprietary).
Revision 2 Report Date: August 2012 Not Used for C2C21
- 12. DPC-NE-2009-P-A, "Westinghouse Fuel Transition Report," (DPC Proprietary).
Revision 3a Report Date: September 2011
- 13. DPC-NE-1004A, "Nuclear Design Methodology Using CASMO-3/S1MULATE-3P."
Revision la Report Date: January 2009 Not Used for C2C21
CNEI-0400-281 Page 7 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 1.1 Analytical Methods (continued)
- 14. DPC-NF-20 10-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-201 I-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 CASMO-4 /
SIMULATE-3 MOX", (DPC Proprietary).
Revision I Report Date: November 12, 2008
- 17. BAW-10231P-A, "COPERNIC Fuel Rod Design Computer Code" (Framatome ANP Proprietary)
Revision I SER Date: January 14, 2004 Not Used for C2C21
CNE1-0400-28 I Page 8 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 2.0 Operating Limits Cycle-specific parameter limits for specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using NRC approved methodologies specified in Section 1.1.
2.1 Reactor Core Safety Limits (TS 2.1.1)
Reactor Core Safety Limits are shown in Figure 1.
2.2 Shutdown Margin - SDM (TS 3.1.1, TS 3.1.4, TS 3.1.5, TS 3.1.6, TS 3.1.8) 2.2.1 For TS 3.1.1, SDM shall be greater than or equal to 1.3% AK/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% AK/K in MODE 5.
2.2.3 For TS 3.1.4, SDM shall be greater than or equal to 1.3% AK/K in MODE I and MODE 2.
2.2.4 For TS 3.1.5, SDM shall be greater than or equal to 1.3% AK/K in MODE I 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% AK/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% AK/K in MODE 2 during PHYSICS TESTS.
CNEI-0400-281 Page 9 Revision 0 Catawba 2 Cycle 21 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 oý 630 228 psia U
620______
610 2100 600195pi 590 ACCEPTABLE OPERATION 580 1
0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power
CNEI-0400-281 Page 10 Revision 0 Catawba 2 Cycle 21 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 AK/K/°F.
EOC, ARO, RTP MTC shall be less negative than the -4.3E-04 AK/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 AK/K/JF.
2.3.3 60 ppm MTC Surveillance Limit is:
Measured 60 ppm ARO, equilibrium RTP MTC shall be less negative than or equal to -4.125E-04 AK/K/°F.
Where:
BOC = Beginning of Cycle (burnup corresponding to most positive MTC)
EOC = End of Cycle ARO = All Rods Out HZP = Hot Zero Thermal Power RTP = Rated Thermal Power ppm = Parts per million (Boron) 2.4 Shutdown Bank Insertion Limit (TS 3.1.5) 2.4.1 Each shutdown bank shall be withdrawn to at least 222 steps. Shutdown banks are withdrawn in sequence and with no overlap.
2.5 Control Bank Insertion Limits (TS 3.1.6) 2.5.1 Control banks shall be within the insertion, sequence, and overlap limits shown in Figure 3. Specific control bank withdrawal and overlap limits as a function of the fully withdrawn position are shown in Table 1.
CNEI-0400-281 Page II Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Figure 2 Moderator Temperature Coefficient Upper Limit Versus Power Level o
1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0.0 i i
I i
I I
I
'i 0
10 20 30 40 50 60 70 80 90 100 Percent of Rated Thermal Power NOTE: Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.
Refer to the Unit 2 ROD manual for details.
CNEI-0400-28 i Page 12 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Figure 3 Control Bank Insertion Limits Versus Percent Rated Thermal Power 231 220 200 180 2
160 140 120 0
- 100 t
60
=
80 0
60 O*
40 0
20 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 Bank CC RIL = 2.3(P) +47 Bank CB RIL = 2.3(P) +163 (30 < P < 100)
(0 < P < 76.1) for CC RIL 222 (76.1 < P < 100)
(0 < P < 25.7) for CB RIL =222 (25.7 < P < 100) where P = % of Rated Thermal Power NOTE:
Compliance with Technical Specification 3.1.3 may require rod withdrawal limits.
Refer to the Unit 2 ROD manual for details.
CNEI-0400-281I Page 13 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Table 1 Control Bank Withdrawal Steps and Sequence Fully Withdrawn at 222 Steps Control Control Control Control Bank A Bank B Bank C Bank D Fully Withdrawn at 223 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 222 Stop 106 0
0 222 116 0OStart 0
222 222 Stop 106 0
2 22 222 116 0OStart 222 222 222 Stop 106 Fully WVithdrawn at 224 Steps Control Control Control Control BankA BankB BankC BankD 0OStart 0
0 0
116 0OStart 0
0 223 Stop 107 0
0 223 116 0OStart 0
223 223 Stop 107 0
223 223 116 0OStart 223 223 223 Stop 107 Fully Withdrawn at 225 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 224 Stop 108 0
0 224 116 0OStart 0
224 224 Stop 108 0
224 224 116 0OStart 224 224 224 Stop 108 Fully WAithdrawn at 226 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 225 Stop 109 0
0 225 116 0OStart 0
225 225 Stop 109 0
225 225 116 0OStart 225 225 225 Stop 109 Fully W~ithdrawn at 227 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 226 Stop hO0 0
0 226 116 0OStart 0
226 226 Stop 110 0
226 226 116 0 Start 226 226 226 Stop 11O Fully W~ithdrawn at 228 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 227 Stop 111 0
0 227 116 0OStart 0
227 227 Stop 111 0
227 227 116 0OStart 227 227 227 Stop IlI Fully Withdrawn at 229 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 228 Stop 112 0
0 228 116 0OStart 0
228 228 Stop 112 0
228 228 116 0 Start 228 228 228 Stop 112 Fully Withdrawn at 230 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 229 Stop 113 0
0 229 116 0OStart 0
229 229 Stop 113 0
229 229 116 0OStart 229 229 229 Stop 113 Fully Withdrawn at 231 Steps Control Control Control Control Bank A Bank B Bank C Bank D 0OStart 0
0 0
116 0OStart 0
0 230OStop 114 0
0 230 116 0OStart 0
230 230OStop 114 0
230 230 116 0OStart 230 230 230 Stop 114 0OStart 0
0 0
116 0OStart 0
0 231 Stop 1I5 0
0 231 116 0OStart 0
231 231 Stop hI5 0
231 231 116 0 Start 231 231 231 Stop 115
CNEI-0400-28 I Page 14 Revision 0 Catawba 2 Cycle 21 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 RTP *K(Z)/P for P > 0.5 F RTP *K(Z)/0.5 for P < 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 FQ surveillance limits as defined in COLR Sections 2.6.5 and 2.6.6.
2.6.2 F "
= 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 'TP with the K(BU) penalty for Westinghouse RFA fuel is analytically confirmed in cycle-specific reload calculation. K(BU) is set to 1.0 at all burnups.
CNEI-0400-281 Page 15 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report The following parameters are required for core monitoring per the Surveillance Requirements of Technical Specification 3.2. 1:
L~
FQ(X,Y,Z)
- MQ(X,Y,Z) 2.6.5
[FQ(XYZ)]°P UMT
- TILT where:
[FC (X,Y,Z)]OP =
F (X,Y,Z)
=
MQ(X,Y,Z)
=
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.
F1 (X,Y,Z)°P includes allowances for calculation and measurement uncertainties.
Design power distribution for FQ. F D (X,Y,Z) is provided in Appendix Table A-I for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial startup operation.
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-I 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)
= Engineering Hot Channel Factor. (MT = 1.03).
TILT = Peaking penalty to account for allowable quadrant power tilt ratio of 1.02. (TILT = 1.035)
[RPS 2.6.6
[FQ(X,Y,)
FQ(X,Y,Z)
- MC(X,Y,Z)
UMT
- TILT where:
[FQ(X,Y,Z)]RP5 =
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.
CNEI-0400-281 Page 16 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report D~XYZ MC(X,Y,Z)
=
D Design power distributions for FQ. FQ(X,Y,Z) is provided in Appendix Table A-I for normal operating conditions and in Appendix Table A-4 for power escalation testing during initial startup operations.
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 =
Measurement Uncertainty (UMT = 1.05)
= Engineering Hot Channel Factor. (MT = 1.03).
TILT =
Peaking penalty to account for allowable quadrant power tilt ratio of 1.02. (TILT = 1.035) 2.6.7 KSLOPE = 0.0725 where:
KSLOPE = adjustment to K1 value from OTAT trip setpoint required to compensate for each 1% measured F,' (X,Y,Z) exceeds [ F RPS (X,Y,Z)]
2.6.8 FQ(X,Y,Z) Penalty Factors for Technical Specification Surveillances 3.2.1.2 and 3.2.1.3 are provided in Table 2.
CNEI-0400-281 Page 17 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Figure 4 K(Z), Normalized FQ(X,Y,Z) as a Function of Core Height for RFA Fuel 1.200 1.000 0.800 0.600 (0.0, 1.00)
(4.0, 1.00)
I (12.0, 0.9259)
(4.0, 0.9259) 0.400 0.200 -
0.000 Core Height (ft)
K(Z) 0.0 1.000
<4 1.000
>4 0.9259 12 0.9259 0.0 2.0 4.0 6.0 Core Height (ft) 8.0 10.0 12.0
CNEI-0400-281 Page 18 Revision 0 Catawba 2 Cycle 21 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 (EFPD) 4 12 25 50 75 100 125 150 175 200 225 250 275 300 325 350 375 400 425 450 475 495 512 517 527 537 FQ(X,Y,Z)
Penalty Factor(%)
2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 FAH(X,Y)
Penalty Factor (%)
2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 2.00 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 FA(X,Y) for compliance with Tech Spec Surveillances 3.2.1.2, 3.2.1.3 and 3.2.2.2.
CNEI-0400-281 Page 19 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 2.7 Nuclear Enthalpy Rise Hot Channel Factor - FALH(X,Y) (TS 3.2.2)
FAH steady-state limits referred to in Technical Specification 3.2.2 are defined by the following relationship.
2.7.1
[FLH(X,Y)]LCO= MARP (X,Y)*
1.0 +
- (1.0- P)I where:
[FHL (X, y)]LCO is the steady-state, maximum allowed radial peak and includes allowances for calculation/measurement uncertainty.
MARP(X,Y) =
Cycle-specific operating limit Maximum Allowable Radial Peaks. MARP(X,Y) radial peaking limits are provided in Table 3.
P Thermal Power Rated Thermal Power RRH = Thermal Power reduction required to compensate for each 1% 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.
SURV FffH (X, Y)
- M An (X, Y) 2.7.2
[ FALH (X,Y)]
-MR*TXLTUMR
- TILT where:
SURV FL (X,Y)]
=
Cycle dependent maximum allowable design peaking factor that ensures FAH(XY) limit is not exceeded for operation within AFD, RIL, and QPTR limits. FL (X,Y)SURV includes allowances for calculational and measurement uncertainty.
D D
F D (X,Y) = Design power distribution for FAH FAH (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-0400-281 Page 20 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report MAH(XY) = Margin remaining in core location X,Y relative to the Operational DNB limits in the transient power distribution.
MAH(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 MAH(XY).
TILT = Peaking penalty to account for allowable quadrant power tilt ratio of 1.02. (TILT = 1.035) 2.7.3 RRH = 3.34 where:
RRH = Thermal Power reduction required to compensate for each 1% measured radial peak, F* (X,Y) exceeds its limit. (0 < P < 1.0) 2.7.4 TRH = 0.04 where:
TRH =
Reduction in OTAT K1 setpoint required to compensate for each 1%
measured radial peak, FAH (X,Y) exceeds its limit.
2.7.5 FAH(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-0400-28 1 Page 21 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Table 3 Maximum Allowable Radial Peaks (MARPs)
RFA Fuel MARPs 100% Full Power Core Height Axial Peak (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 1.8092 1.8553 1.9248 1.9146 1.9179 2.0621 2.0498 2.0090 1.9333 1.8625 1.7780 1.3151 1.2461 1.20 1.8102 1.8540 1.9248 1.9146 1.9179 2.1073 2.0191 1.9775 1.9009 1.8306 1.7852 1.3007 1.2235 2.40 1.8093 1.8525 1.9312 1.9146 1.9179 2.0735 1.9953 1.9519 1.8760 1.8054 1.7320 1.4633 1.4616 3.60 1.8098 1.8514 1.9204 1.9146 1.9179 2.0495 1.9656 1.9258 1.8524 1.7855 1.6996 1.4675 1.3874 4.80 1.8097 1.8514 1.9058 1.9146 1.9179 2.0059 1.9441 1.9233 1.8538 1.7836 1.6714 1.2987 1.2579 6.00 1.8097 1.8514 1.8921 1.9212 1.9179 1.9336 1.8798 1.8625 1.8024 1.7472 1.6705 1.3293 1.2602 7.20 1.8070 1.8438 1.8716 1.8930 1.8872 1.8723 1.8094 1.7866 1.7332 1.6812 1.5982 1.2871 1.2195 8.40 1.8073 1.8319 1.8452 1.8571 1.8156 1.7950 1.7359 1.7089 1.6544 1.6010 1.5127 1.2182 1.1578 9.60 1.8072 1.8102 1.8093 1.7913 1.7375 1.7182 1.6572 1.6347 1.5808 1.5301 1.4444 1.1431 1.0914 10.80 1.7980 1.7868 1.7611 1.7163 1.6538 1.6315 1.5743 1.5573 1.5088 1.4624 1.3832 1.1009 1.0470 11.40 1.7892 1.7652 1.7250 1.6645 1.6057 1.5826 1.5289 1.5098 1.4637 1.4218 1.3458 1.0670 1.0142
CNEI-0400-281 Page 22 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Figure 5 Percent of Rated Thermal Power Versus Percent Axial Flux Difference Limits 0
H 0
C.)
~L)
-50
-40
-30
-20
-10 0
10 20 30 40 50 Axial Flux Difference (% Delta I)
NOTE: Compliance with Technical Specification 3.2.1 may require more restrictive AFD limits. Refer to the Unit 2 ROD manual for operational AFD limits.
CNEI-0400-281 Page 23 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 2.9 Reactor Trip System Instrumentation Setpoints (TS 3.3.1) Table 3.3.1-1 2.9.1 Overtemperature AT Setpoint Parameter Values Parameter Nominal Value Nominal Tavg at RTP Nominal RCS Operating Pressure Overtemperature AT reactor trip setpoint Overtemperature AT reactor trip heatup setpoint penalty coefficient Overtemperature AT reactor trip depressurization setpoint penalty coefficient Time constants utilized in the lead-lag compensator for AT Time constant utilized in the lag compensator for AT Time constants utilized in the lead-lag compensator for Tavg Time constant utilized in the measured Tavg lag compensator fl(Al) "positive" breakpoint fl(AI) "negative" breakpoint fl(AI) "positive" slope fl(A1) "negative" slope T' < 590.8 OF P'= 2235 psig KI = 1.1953 K2 = 0.03163/°F K3 = 0.001414/psi Tj = 8 sec.
T2 = 3 sec.
T3 = 0 sec.
T4 = 22 sec.
T5 = 4 sec.
T6 = 0 sec.
= 3.0 %AI
= N/A*
= 1.525 %AT 0/%A1
= N/A*
fl(AI) negative breakpoints and slopes for OTAT are less restrictive than OPAT fE(AI) negative breakpoint and slope. Therefore, during a transient which challenges negative imbalance limits, OPAT f,(Al) limits will result in a reactor trip before OTAT fl (Al) limits are reached. This makes implementation of an OTAT fl (Al) negative breakpoint and slope unnecessary.
CNEI-0400-281 Page 24 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Overpower AT Setpoint Parameter Values 2.9.2 Parameter Nominal Tavg at RTP Overpower AT reactor trip setpoint Overpower AT reactor trip penalty Overpower AT reactor trip heatup setpoint penalty coefficient Time constants utilized in the lead-lag compensator for AT Time constant utilized in the lag compensator for AT Time constant utilized in the measured Tavg lag compensator Time constant utilized in the rate-lag controller for Tavg f2(AI) "positive" breakpoint f2(AI) "negative" breakpoint f2(AI) "positive" slope f2(AI) "negative" slope Nominal Value T" < 590.8 OF K4 = 1.0819 K5 = 0.02 / OF for increasing Tavg K5 = 0.00 / OF for decreasing Tavg K 6 = 0.001291/OF forT> T" K 6 = 0.0 /OF for T < T"
-i1 = 8 sec.
'U2 = 3 sec.
"C3 = 0 sec.
"6 = 0 sec.
T7 = 10 sec.
= 35.0 %AI
= -35.0 %AI
= 7.0 %ATo/ %AI
= 7.0 %ATo/ %AI
CNEI-0400-281 Page 25 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 2.10 Boron Dilution Mitigation System (TS 3.3.9) 2.10.1 Reactor Makeup Water Pump flow rate limits:
Applicable Mode Limit MODE 3
< 80 gpm MODE 4 or 5
< 70 gpm 2.11 RCS Pressure, Temperature and Flow Limits for DNB (TS 3.4.1)
RCS pressure, temperature and flow limits for DNB are shown in Table 4.
2.12 Accumulators (TS 3.5.1) 2.12.1 Boron concentration limits during MODES I 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 2,425 ppm Accumulator minimum boron concentration.
300.1 - 350 EFPD 2,325 ppm Accumulator minimum boron concentration.
350.1 -400 EFPD 2,244 ppm Accumulator minimum boron concentration.
400.1 -450 EFPD 2,170 ppm Accumulator minimum boron concentration.
450.1 - 527 EFPD 2,101 ppm Accumulator minimum boron concentration.
527.1 - 537 EFPD 1,991 ppm Accumulator maximum boron concentration.
0 - 537 EFPD 3,075 ppm
CNEI-0400-281 Page 26 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Table 4 Reactor Coolant System DNB Parameters No. Operable PARAMETER INDICATION CHANNELS LIMITS
- 1. Indicated RCS Average Temperature meter 4
< 589.6 OF meter 3
< 589.3 OF computer 4
< 590.1 OF computer 3
< 589.9 OF
- 2. Indicated Pressurizer Pressure meter 4
> 2219.8 psig meter 3
> 2222.1 psig computer 4
> 2215.8 psig computer 3
> 2217.5 psig
- 3. RCS Total Flow Rate
> 390,000 gpm
CNEI-0400-281 Page 27 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 2.13 Refueling Water Storage Tank - RWST (TS 3.5.4) 2.13.1 Boron concentration limits during MODES 1, 2, 3, and 4:
Parameter Limit RWST minimum boron concentration.
RWST maximum boron concentration.
2,700 ppm 3,075 ppm 2.14 Spent Fuel Pool Boron Concentration (TS 3.7.15) 2.14.1 Minimum boron concentration limit for the spent fuel pool. Applicable when fuel assemblies are stored in the spent fuel pool.
Parameter Limit Spent fuel pool minimum boron concentration.
2,700 ppm 2.15 Refueling Operations - Boron Concentration (TS 3.9.1) 2.15.1 Minimum boron concentration limit for filled portions of the Reactor Coolant System, refueling canal, and refueling cavity for MODE 6 conditions. The minimum boron concentration limit and plant refueling procedures ensure that core Keff remains within MODE 6 reactivity requirement of Keff < 0.95.
Parameter Limit Minimum boron concentration of the Reactor Coolant System, the refueling canal, and the refueling cavity.
2,700 ppm
g CNEI-0400-281 Page 28 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report 2.16 Standby Shutdown System - (SLC-16.7-9) 2.16.1 Minimum boron concentration limit for the spent fuel pool required for Standby Makeup Pump Water Supply. Applicable for MODES 1, 2, and 3.
Parameter Limit Spent fuel pool minimum boron concentration for TR 16.7-9.3.
2,700 ppm 2.17 Borated Water Source - Shutdown (SLC 16.9-11) 2.17.1 Volume and boron concentrations for the Boric Acid Tank (BAT) and the Refueling Water Storage Tank (RWST) during MODE 4 with any RCS cold leg temperature < 210°F, and MODES 5 and 6.
Parameter Limit BAT minimum boron concentration 7,000 ppm 2000 gallons Volume of 7,000 ppm boric acid solution required to maintain SDM at 68 'F NOTE: When cycle burnup is > 460 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)
RWST minimum boron concentration Volume of 2,700 ppm boric acid solution required to maintain SDM at 68 '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-0400-281 Page 29 Revision 0 Catawba 2 Cycle 21 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°F *.
- NOTE: The SLC 16.9-12 applicability is down to MODE 4 temperatures of
> 210TF. The minimum volumes calculated support cooldown to 200'F to satisfy UFSAR Chapter 9 requirements.
Parameter Limit BAT minimum boron concentration 7,000 ppm Volume of 7,000 ppm boric acid solution required 13,500 gallons to maintain SDM at 210°F NOTE: When cycle burnup is > 460 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)
RWST minimum boron concentration Volume of 2,700 ppm boric acid solution required to maintain SDM at 210 °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)
CNEI-0400-28 1 Page 30 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Figure 6 Boric Acid Storage Tank Indicated Level Versus Primary Coolant Boron Concentration (Valid When Cycle Burnup is > 460 EFPD)
This figure includes additional volumes listed in SLC 16.9-11 and 16.9-12 50.0 45.0 RCS Boron Concentration BAT Level (ppm)
(%level) 40.0 0300 43.0 300 < 500 40.0 35.0 -
500 < 700 37.0 700 < 1000 30.0 3 0.0 -
1 0 0 0..
..1
- 4. 9 1300 <2700 9.8
._J
~ ~~~~
~~.....................
- 25.
> 2700 9.8 Unacceptable 20.0 -
Operation Acceptable Operation 15.0 1 0.0 5.0 0.0 0
200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 2600 Primary Coolant Boron Concentration (ppmb)
CNEI-0400-281 Page 31 Revision 0 Catawba 2 Cycle 21 Core Operating Limits Report Appendix A Power Distribution Monitoring Factors Appendix A contains power distribution monitoring factors used in Technical Specification Surveillance. This data was generated in the Catawba 2 Cycle 21 Maneuvering Analysis calculation file, CNC-1553.05-00-0618. 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 questions concerning this information arise.
Appendix A is included in the COLR transmitted to the NRC.