LR-N24-0068, Core Operating Limits Report - Cycle 28

Core Operating Limits Report - Cycle 28

ML24295A378
Person / Time
Site:	Salem
Issue date:	10/21/2024
From:	Jennings J Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N24-0068
Download: ML24295A378 (1)

Text

Jason Jennings Director, Site Regulatory Compliance PO Box 236 o PSEG I NUCLEAR Hancocks Bridge, New Jersey 0 8038- 0221

LR-N24-0068 Technical Specification 6.9.1.9

October 21, 2024

U.S. Nuclear R egulatory Commission ATTN: Document Control Desk Washington, D C 20555 -0001

Salem Nuclear Generating Station, Unit 2 Renewed Facility Operating License No. DPR-75 NRC Docket Nos. 50 -311

Subject:

Salem Unit 2 Core Operating Limits Report - Cycle 28

PSEG Nuclear LLC submits the enclosed Core Operating Limits Report (COLR) for Salem Unit 2, Cycle 28 per administrative control 6.9.1.9.

There are no commitments contained in this letter.

Should you have any questions regarding this submittal, please contact Harry Balian at (856) 339 - 2 173.

Sincerely,

Jason Jennings Director, Site Regulatory Compliance PSEG Nuclear LLC

Enclosure:

Core Operating Limits Report (COLR)

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 Corporate Commitment Coordinator, PSEG Nuclear LLC LR-N24-0068

Enclosure

Salem Unit 2

Core Operating Limits Report (COLR)

Cycle 28

COLR SALEM 2 Revision 12 July 2024

Core Operating Limits Report for Salem Unit 2, Cycle 28

Page 1 of 14

• • • This record was final approved on 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 2 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 3 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 4 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 1.0 CORE OPERATING LIMITS REPORT

This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 2 8 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.

TS NRC Approved Section Technical Specifications COLR Parameter COLR Section Methodology (Section 3.0 Number) 3.1.1.3 Moderator Temperature Coefficient MTC 2.1 3.1, 3.6 3.1.3.5 Control Rod Insertion Limits Control Rod Insertion 2.2 3.1, 3.6 Limits 3.2.1 Axial Flux Difference AFD 2.3 3.1, 3.2, 3.6 3.2.2 Heat Flux Hot Channel Factor - FQ(z) 2.4 3.1, 3.3, 3.4, 3.5, 3.6, FQ(z) 3.7, 3.8, 3.9 3.2.3 Nuclear Enthalpy Rise Hot Channel FN H 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 5 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 2.0 OPERATING LIMITS

The cycle-specific parameter limits for the specifications listed in Sec tion 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 6 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024

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]

FQ(z) FQRTPP* K( z) for P > 0.5

FQ(z) FQRTP0.5* K( z) for P 0.5

where: P = THERMAL POWERRATED 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 2.00 for unrodded upper core planes 7 through 8 1.84 for unrodded upper core planes 9 through 11 1.78 for unrodded upper core planes 12 through 13 1.77 for unrodded upper core planes 14 through 18 1.80 for unrodded upper core planes 19 through 31 1.82 for unrodded lower core planes 32 through 43 1.86 for unrodded lower core planes 44 through 48 1.94 for unrodded lower core planes 49 through 50 1.93 for unrodded lower core planes 51 through 53

• • • This record was final approved on 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 7 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024

2.08 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 8000 MWD/MTU to 12000 MWD/MTU

FxyRTP = 2.14 for unrodded upper core planes 1 through 6 1.84 for unrodded upper core planes 7 through 8 1.78 for unrodded upper core planes 9 through 11 1.76 for unrodded upper core planes 12 through 13 1.76 for unrodded upper core planes 14 through 18 1.82 for unrodded upper core planes 19 through 31 1.87 for unrodded lower core planes 32 through 43 1.84 for unrodded lower core planes 44 through 48 1.88 for unrodded lower core planes 49 through 50 1.84 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

where: from 12000 MWD/MTU to 14000 MWD/MTU FxyRTP = 2.05 for unrodded upper core planes 1 through 6 1.84 for unrodded upper core planes 7 through 8 1.78 for unrodded upper core planes 9 through 11 1.76 for unrodded upper core planes 12 through 13 1.78 for unrodded upper core planes 14 through 18 1.93 for unrodded upper core planes 19 through 31 1.93 for unrodded lower core planes 32 through 43 1.84 for unrodded lower core planes 44 through 48 1.87 for unrodded lower core planes 49 through 50 1.82 for unrodded lower core planes 51 through 53 1.88 for unrodded lower core planes 54 through 55 2.10 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 8 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 where: from 14000 MWD/MTU to EOL FxyRTP = 1.97 for unrodded upper core planes 1 through 6 1.78 for unrodded upper core planes 7 through 8 1.73 for unrodded upper core planes 9 through 11 1.79 for unrodded upper core planes 12 through 13 1.81 for unrodded upper core planes 14 through 18 1.99 for unrodded upper core planes 19 through 31 1.96 for unrodded lower core planes 32 through 43 1.85 for unrodded lower core planes 44 through 48 1.86 for unrodded lower core planes 49 through 50 1.80 for unrodded lower core planes 51 through 53 1.82 for unrodded lower core planes 54 through 55 1.98 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:

---Q U U UFQ= + 1 0 100 0.. *e 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 U UFQ qu e= *=

= wher

=

====1.03

• • • This record was final approved on 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 9 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNH (Specification 3.2.3)

FNH = F RTPH [1.0 + PFH (1.0 - P)]

where: P = THERMAL POWERRATED 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, F NH, shall be the greater of 1.04 or as calculated by the following formula:

U F U +H=1 H.0100.0= =

wher= 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 F NH shall be calculated by the following formula:

U F U =HFHm=

=

= = wher== UFHm = Base F H measurement uncertainty.

= 1.04 2.6 Boron Concentration (Specification 3.9.1)

A Mode 6 boron concentration, maintained at or above 2111ppm, 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 10 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 11 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024 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. 330, Renewed License No. DPR-75, Docket No. 50-311.

• • • This record was final approved on 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 12 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024

FIGURE 1

ROD BANK INSERTION LIMITS VS. THERMAL POWER

24 0

22 0..VI / 17.1, 227 V, 70.4, 227

20 0 / I I BANK B V V / V

18 0 ~ I V I ? 0, 186 100, 170

16 0 /

/ V I

14 0 VI / BANK C V ~I /

12 0 /

V / /

10 0 /

V /

80 V /4'1 I BANK D

/ V V V

60 I I V 0, 58

40 /

/ V 20 29, 0 V

0 R/

0 10 20 30 40 50 60 70 80 90 10 0 PERCENT OF RATED THERM AL POWER (%)

• • • This record was final approved on 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 13 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024

FIGURE 2

AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER

100 I I I I (-11,90) (11,90)

80 I \\

UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION AC C EPTABLE OPERATION

60 I \\

\\

(-31,50) ' (31,50)

I I I 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 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)

COLR SALEM 2 PSEG Nuclear LLC Page 14 of 14 Revision 12 SALEM UNIT 2 CYCLE 28 COLR July 2024

FIGURE 3

K(z) - NORMALIZED F Q(z) AS A FUNCTION OF CORE HEIGHT

1.2

1.0

0.8 FQ K(Z) He ight (FT)

2.40 1.0 0.0 2.40 1.0 6.0 0.6 2.22 0.925 12.0

0.4

0.2

0.0 0 2 4 6 8 10 12

CORE HE IGHT (FE E T)

• • • This record was final approved on 07/30/2024 09:57:45. (This statement was added by the PRIME system upon its validation)