LR-N23-0033, Core Operating Limits Report Cycle 27
| ML23116A333 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/26/2023 |
| From: | Jennings J Public Service Enterprise Group |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| LR-N23-0033 | |
| Download: ML23116A333 (1) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 0 PSEG NuclearLLC Technical Specification 6.9.1.9 LR-N23-0033 April 26, 2023 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-0001 Salem Generating Station Unit 2 Renewed Facility Operating License DPR-75 NRC Docket No. 50-311
Subject:
Salem Unit 2 Core Operating Limits Report - Cycle 27 In accordance with section 6.9.1.9 of the Salem Unit 2 Technical Specifications, PSEG Nuclear, LLC, submits the enclosed Core Operating Limits Report (COLR) for Salem Unit 2, Cycle 27.
There are no commitments contained in this letter.
Should you have any questions regarding this submittal, please contact Bernadette Cizin at (856) 339 - 2206.
Sincerely, Jennings, Digitally signed by Jennings, Jason Jason Date: 2023.04.26 07:47:08 -04'00' Jason Jennings Director, Site Regulatory Compliance PSEG Nuclear LLC
Enclosure:
Core Operating Limits Report (COLR)
April 26, 2023 Page 2 Technical Specification 6.9.1.9 LR-N23-0033 cc: USNRC Regional Administrator Region 1 USNRC NRR Project Manager Salem USNRC Senior Resident Inspector - Salem NJ Department of Environmental Protection, Bureau of Nuclear Engineering Commitment Coordinator, Salem Generating Station Corporate Commitment Coordinator, PSEG Nuclear, LLC President & Chief Nuclear Officer Site Vice President - Salem Plant Manager - Salem Executive Director, Regulatory Affairs and Nuclear Oversight Director, Site Regulatory Compliance Manager - Nuclear Oversight Manager - Licensing Records Management
LR-N23-0033 Enclosure Salem Unit 2 Core Operating Limits Report (COLR)
Cycle 27
COLR SALEM 2 Revision 11 January 2023 Core Operating Limits Report for Salem Unit 2, Cycle 27 Page 1 of 14
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 2 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 TABLE OF CONTENTS Section Section Title Page Number Number Table of Contents 2 List of Figures 3 1.0 Core Operating Limits Report 4 2.0 Operating Limits 5 2.1 Moderator Temperature Coefficient (Specification 3.1.1.3) 5 2.2 Control Rod Insertion Limits (Specification 3.1.3.5) 6 2.3 Axial Flux Difference (Specification 3.2.1) 6 2.4 Heat Flux Hot Channel Factor - FQ(z) (Specification 3.2.2) 6 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNH (Specification 3.2.3) 9 2.6 Boron Concentration (Specification 3.9.1) 9 3.0 Analytical Methods 10 4.0 References 11
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 3 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 LIST OF FIGURES Figure Figure Title Page Number Number 1 Rod Bank Insertion Limits vs. Thermal Power 12 2 Axial Flux Difference Limits as a Function of Rated Thermal Power 13 3 K(z) - Normalized FQ(z) as a Function of Core Height 14
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 4 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 27 has been prepared in accordance with the requirements of Technical Specification 6.9.1.9. The Technical Specifications affected by this report are listed below along with the NRC-approved methodologies used to develop and/or determine COLR parameters identified in Technical Specifications.
NRC Approved TS COLR Technical Specifications COLR Parameter Methodology Section Section (Section 3.0 Number) 3.1.1.3 Moderator Temperature Coefficient MTC 2.1 3.1, 3.6 Control Rod Insertion 3.1.3.5 Control Rod Insertion Limits 2.2 3.1, 3.6 Limits 3.2.1 Axial Flux Difference AFD 2.3 3.1, 3.2, 3.6 Heat Flux Hot Channel Factor - 3.1, 3.3, 3.4, 3.5, 3.6, 3.2.2 FQ(z) 2.4 FQ(z) 3.7, 3.8, 3.9 Nuclear Enthalpy Rise Hot Channel 3.2.3 FNH 2.5 3.1, 3.5, 3.6, 3.8, 3.9 Factor - FNH 3.9.1 Boron Concentration Boron Concentration 2.6 3.1, 3.6
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 5 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.9.1.9 and in Section 3.0 of this report.
2.1 Moderator Temperature Coefficient (Specification 3.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:
The BOL/ARO/HZP-MTC shall be less positive than or equal to 0 k/k/°F.
The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.4x10-4 k/k/°F.
2.1.2 The MTC Surveillance limit is:
The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.7x10-4 k/k/°F.
where: BOL stands for Beginning of Cycle Life ARO stands for All Rods Out HZP stands for Hot Zero THERMAL POWER EOL stands for End of Cycle Life RTP stands for RATED THERMAL POWER
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 6 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 2.2 Control Rod Insertion Limits (Specification 3.1.3.5) 2.2.1 The control rod banks shall be limited in physical insertion as shown in Figure 1.
2.3 Axial Flux Difference (Specification 3.2.1)
[Constant Axial Offset Control (CAOC) Methodology]
2.3.1 The Axial Flux Difference (AFD) target band shall be (+6%, -9%).
2.3.2 The AFD Acceptable Operation Limits are provided in Figure 2.
2.4 Heat Flux Hot Channel Factor - FQ(z) (Specification 3.2.2)
[Fxy Methodology]
FQRTP FQ(z) P
- K(z) for P > 0.5 FQRTP FQ(z) 0.5
- K(z) for P 0.5 THERMAL POWER where: P = RATED THERMAL POWER 2.4.1 FQRTP = 2.40 2.4.2 K(z) is provided in Figure 3.
2.4.3 FxyL = FxyRTP [1.0 + PFxy (1.0 - P)]
where: from BOL to 8000 MWD/MTU FxyRTP = 2.30 for unrodded upper core planes 1 through 6 1.98 for unrodded upper core planes 7 through 8 1.80 for unrodded upper core planes 9 through 11 1.73 for unrodded upper core planes 12 through 13 1.72 for unrodded upper core planes 14 through 18 1.77 for unrodded upper core planes 19 through 31 1.79 for unrodded lower core planes 32 through 43 1.83 for unrodded lower core planes 44 through 48 1.94 for unrodded lower core planes 49 through 50 1.87 for unrodded lower core planes 51 through 53
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 7 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 1.90 for unrodded lower core planes 54 through 55 1.93 for unrodded lower core planes 56 through 61 2.15 for the core planes containing Bank D control rods PFxy = 0.3 where: from 8000 MWD/MTU to 12000 MWD/MTU FxyRTP = 2.16 for unrodded upper core planes 1 through 6 1.82 for unrodded upper core planes 7 through 8 1.74 for unrodded upper core planes 9 through 11 1.70 for unrodded upper core planes 12 through 13 1.70 for unrodded upper core planes 14 through 18 1.77 for unrodded upper core planes 19 through 31 1.87 for unrodded lower core planes 32 through 43 1.88 for unrodded lower core planes 44 through 48 1.93 for unrodded lower core planes 49 through 50 1.89 for unrodded lower core planes 51 through 53 2.00 for unrodded lower core planes 54 through 55 2.30 for unrodded lower core planes 56 through 61 2.15 for the core planes containing Bank D control rods PFxy = 0.3 where: from 12000 MWD/MTU to 14000 MWD/MTU FxyRTP = 2.08 for unrodded upper core planes 1 through 6 1.82 for unrodded upper core planes 7 through 8 1.74 for unrodded upper core planes 9 through 11 1.72 for unrodded upper core planes 12 through 13 1.72 for unrodded upper core planes 14 through 18 1.88 for unrodded upper core planes 19 through 31 1.95 for unrodded lower core planes 32 through 43 1.87 for unrodded lower core planes 44 through 48 1.90 for unrodded lower core planes 49 through 50 1.85 for unrodded lower core planes 51 through 53 1.92 for unrodded lower core planes 54 through 55 2.18 for unrodded lower core planes 56 through 61 2.15 for the core planes containing Bank D control rods PFxy = 0.3
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 8 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 where: from 14000 MWD/MTU to EOL FxyRTP = 2.03 for unrodded upper core planes 1 through 6 1.82 for unrodded upper core planes 7 through 8 1.75 for unrodded upper core planes 9 through 11 1.75 for unrodded upper core planes 12 through 13 1.77 for unrodded upper core planes 14 through 18 1.96 for unrodded upper core planes 19 through 31 1.97 for unrodded lower core planes 32 through 43 1.87 for unrodded lower core planes 44 through 48 1.88 for unrodded lower core planes 49 through 50 1.82 for unrodded lower core planes 51 through 53 1.86 for unrodded lower core planes 54 through 55 2.06 for unrodded lower core planes 56 through 61 2.15 for the core planes containing Bank D control rods PFxy = 0.3 2.4.4 If the Power Distribution Monitoring System (PDMS) is used for core power distribution surveillance and is OPERABLE, as defined in Technical Specification 3.3.3.14, the uncertainty, UFQ, to be applied to the Heat Flux Hot Channel Factor FQ(z) shall be calculated by the following formula:
UQ U FQ 10. - - - U e 100.0 where:
UQ = Uncertainty for power peaking factor as defined in equation 5-19 of Analytical Method 3.5.
Ue = Engineering uncertainty factor.
= 1.03 Note: UFQ= PDMS Surveillance Report Core Monitor Fxy Uncertainty in %.
2.4.5 If the INCORE movable detectors are used for core power distribution surveillance, the uncertainty, UFQ, to be applied to the Heat Flux Hot Channel Factor FQ(z) shall be calculated by the following formula:
U FQ U qu U e where:
Uqu = Base FQ measurement uncertainty.
= 1.05 Ue = Engineering uncertainty factor.
= 1.03
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 9 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNH (Specification 3.2.3)
FNH = FRTPH [1.0 + PFH (1.0 - P)]
THERMAL POWER where: P = RATED THERMAL POWER 2.5.1 FRTPH = 1.65 2.5.2 PFH = 0.3 2.5.3 If the Power Distribution Monitoring System (PDMS) is used for core power distribution surveillance and is OPERABLE, as defined in Technical Specification 3.3.3.14, the uncertainty, UFH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FNH, shall be the greater of 1.04 or as calculated by the following formula:
U H U FH 1.0 100.0 where: UH = Uncertainty for enthalpy rise hot channel factor as defined in equation 5-19 of Analytical Method 3.5.
2.5.4 If the INCORE movable detectors are used for core power distribution surveillance, the uncertainty, UFH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FNH shall be calculated by the following formula:
U FH U FHm where: UFHm = Base FH measurement uncertainty.
= 1.04 2.6 Boron Concentration (Specification 3.9.1)
A Mode 6 boron concentration, maintained at or above 2055 ppm, in the Reactor Coolant System, the fuel storage pool, the refueling canal, and the refueling cavity ensures the most restrictive of the following reactivity conditions is met:
a) A K-effective (Keff) of 0.95 or less at All Rods In (ARI), Cold Zero Power (CZP) conditions with a 1% k/k uncertainty added.
b) A Keff of 0.99 or less at All Rods Out (ARO), CZP conditions with a 1% k/k uncertainty added.
c) A boron concentration of greater than or equal to 2000 ppm, which includes a 50 ppm conservative allowance for uncertainties.
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 10 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 3.0 ANALYTICAL METHODS The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.
3.1 WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985 (Westinghouse proprietary). Methodology for Specifications listed in 6.9.1.9.a. Approved by Safety Evaluation dated May 28, 1985.
3.2 WCAP-8385, Power Distribution Control and Load Following Procedures - Topical Report, September 1974 (Westinghouse proprietary). Methodology for Specification 3/4.2.1 Axial Flux Difference. Approved by Safety Evaluation dated January 31, 1978.
3.3 WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (Westinghouse proprietary). Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved for Salem by NRC letter dated August 25, 1993.
3.4 WCAP-10266-P-A, Revision 2, The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code, March 1987 (Westinghouse proprietary). Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.
3.5 WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, August 1994 (Westinghouse proprietary). Approved by Safety Evaluation dated February 16, 1994.
3.6 CENPD-397-P-A, Revision 01, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology, May 2000. Approved by Safety Evaluation dated March 20, 2000.
3.7 WCAP-10054-P-A, Addendum 2, Revision 1, Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model, July 1997 (Westinghouse proprietary). Approved by Safety Evaluation dated August 12, 1996.
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 11 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 3.8 WCAP-12472-P-A, Addendum 1-A, BEACON Core Monitoring and Operations Support System, January 2000 (Westinghouse proprietary). Approved by Safety Evaluation dated September 30, 1999.
3.9 WCAP-12472-P-A, Addendum 4, BEACON Core Monitoring and Operation Support System, Addendum 4, September 2012 (Westinghouse proprietary). Approved by Safety Evaluation dated August 9, 2012.
4.0 REFERENCES
- 1. Salem Nuclear Generating Station Unit No. 2, up to Amendment No. 324, Renewed License No. DPR-75, Docket No. 50-311.
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM2 PSEG Nuclear LLC Page 12 ofl4 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 240 220
~ "1 17.5, 228 I i' V V
~
/ I BANKB V 110.8, 228 I
200 I /
180 0, 186 I
/
'2
~ /
/ 100,170~
e 160
'C
~
~ 140
/ '
,I BANK C I ,/
/
Ill Q.
/ V s
~
z 120
/ /
0 E / V fl) 0 100
,/ /
V Q.
~
z / I
/ 1 c( 80 BANKO V
al V
..J 0
et::
.... 60
/ /
z 0 10, 58 /
(.)
40
/
/
/
20 0
129 I
oI I / '
0 20 40 60 80 100 PERCENT OF RATED THERMAL POWER(%)
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLR SALEM 2 PSEG Nuclear LLC Page 13 of 14 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 I(-11,90) I I(11,90) I 80 UNACCEPTABLE I \ UNACCEPTABLE OPERATION OPERATION ACCEPTABLE OPERATION Percent of Rated Thermal Power (%)
60 I
I \ \
I
(-31,50)
I I(31,50) I 40 20 0
-50 -40 -30 -20 -10 0 10 20 30 40 50 Flux Difference (% Delta I)
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)
COLRSALEM2 PSEG Nuclear LLC Page 14 ofl4 Revision 11 SALEM UNIT 2 CYCLE 27 COLR January 2023 FIGURE3 K(z) - NORMALIZED Fo(z) AS A FUNCTION OF CORE HEIGHT 1.2 1.0 g
- .a 0.8 ei::
0 FQ K(Z) Height (FT)
E-<
~
1 2 .40 1.0 0 .0 C) 2 .40 1.0 6 .0
~
IJ.l ii.
0.6 2 .22 0.925 12 .0 N
~
~ 0.4 0
z 0.2 0.0 0 2 4 6 8 12 CORE HEIGHT (FEET)
- This record was final approved on 1/30/2023, 12:23:28 PM. (This statement was added by the PRIME system upon its validation)