LR-N16-0107, Core Operating Limits Report, Cycle 25

Core Operating Limits Report, Cycle 25

ML16187A335
Person / Time
Site:	Salem
Issue date:	07/05/2016
From:	Carr E Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N16-0107 COLR SALEM 1, Rev. 7
Download: ML16187A335 (16)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 JUL 0 5 2016 LR-N16-01 07 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-0001 Salem Generating Station Unit 1 Renewed Facility Operating License DPR-70 NRC Docket No. 50-272 Technical Specification 6.9.1.9

Subject:

Salem Unit 1 Core Operating Limits Report-Cycle 25 In accordance with section 6.9.1.9 of the Salem Unit 1 Technical Specifications, PSEG Nuclear LLC submits the enclosed Core Operating Limits Report (COLR) for Salem Unit 1, Cycle 25.

There are no commitments contained in this letter.

Should you have any questions regarding this submittal, please contact Mr. Michael Phillips at 856-339-1873.

Sincerely, Eric Carr Plant Manager Salem Generating Station map Enclosure

Page2 LR-N16-01 07 cc:

Mr. D. Dorman, USNRC -Administrator-Region 1 Ms. C. Sanders, US NRC -Licensing Project Manager-Salem Mr. P. Finney, USNRC Senior Resident Inspector Mr. P. Mulligan, NJBNE Manager IV Mr. T. Cachaza, Salem Commitment Tracking Coordinator Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. M. Washington, Chief Inspector-Occupational Safety and Health Bureau of Boiler and Pressure Vessel Compliance LR-N16-01 07 Enclosure Salem Unit 1 Core Operating Limits Report (COLR)

Cycle 25

COLRSALEM 1 Revision 7 January 2016 Core Operating Limits Report for Salem Unit 1, Cycle 25 Page 1 of 13

COLRSALEM 1 PSEG Nuclear LLC Page 2 of 13 Revision 7 SALEM UNIT 1 CYCLE 25 COLR January 2016 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 3.1.1.4) 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 FN t.H (Specification 3.2.3) 8 2.6 Boron Concentration (Specification 3.9.1) 9 3.0 Analytical Methods 9

4.0 References 10

COLRSALEM 1 Revision 7 January 2016 Figure Number 1

2 3

PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR LIST OF FIGURES Figure Title Rod Bank Insertion Limits vs. Thermal Power Axial Flux Difference Limits as a Function of Rated Thermal Power K(z)- Normalized FQ(z) as a Function of Core Height Page 3 of 13 Page Number 11 12 13

TS COLRSALEM 1 Revision 7 January 2016 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR 1.0 CORE OPERATING LIMITS REPORT Page 4 ofl3 This Core Operating Limits Report (COLR) for Salem Unit 1 Cycle 25 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.

COLR NRC Approved Section Technical Specifications COLR Parameter Section Methodology (Section 3.0 Number) 3.1.1.4 Moderator Temperature Coefficient MTC 2.1 3.1, 3.6 3.1.3.5 Control Rod Insertion Limits Control Rod Insertion Limits 2.2 3.1, 3.6 3.2.1 Axial Flux Difference AFD 2.3 3.1, 3.2, 3.6 3.2.2 Heat Flux Hot Channel Factor - F Q(Z)

FQ(Z) 2.4 3.1, 3.3, 3.4, 3.5, 3.6, 3.7 3.2.3 Nuclear Enthalpy Rise Hot Channel FN6H 2.5 3.1, 3.5, 3.6, 3.7 Factor-FN L'lH 3.9.1 Boron Concentration Boron Concentration 2.6 3.1, 3.6

COLRSALEM 1 Revision 7 January 2016 2.0 OPERATING LIMITS PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR Page 5 of 13 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 3.1.1.4) 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

COLRSALEM 1 Revision 7 January 2016 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR 2.2 Control Rod Insertion Limits (Specification 3.1.3.5)

Page 6 ofl3 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)

[F xy Methodology]

FQRTP FQ(Z) :::;

p

• K(Z) for P > 0.5 FQRTP FQ(Z) :::;

• K(Z) for P s0.5

0.5 where

THERMAL POWER p

RATED THERMAL POWER 2.4.1 FQRTP

= 2.40 2.4.2 K(Z) is provided in Figure 3.

2.4.3 F:"/

= F:wRTP {1.0 + PFxy (1.0- P)}

where: from BOL to 12000 MWD/MTU F RTP xy

2. 00 for unrodded upper core planes 1 through 6 1.90 for unrodded upper core planes 7 through 8

1. 7 3 for unrodded upper core planes 9 through 18

1. 7 6 for unrodded upper core planes 19 through 31 1.81 for unrodded lower core planes 32 through 43 1.81 for unrodded lower core planes 44 through 53 1.92 for tmrodded lower core planes 54 through 55

2. 00 for unrodded lower core planes 56 through 61 2.13 for the core planes containing Bank D control rods 0.3

COLRSALEM 1 Revision 7 January 2016 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR Page 7 of 13 where: from 12000 MWD/MTU to EOL F RTP

=

xy

2. 00 for unrodded upper core planes 1 through 6 1.86 for unrodded upper core planes 7 thiough 8

1. 78 for unrodded upper core planes 9 through 18

1. 88 for unrodded upper core planes 19 through 31 1.90 for unrodded lower core planes 32 through 43

1. 79 for unrodded lower core planes 44 through 53

1. 79 for unrodded lower core planes 54 through 55 1.94 for unrodded lower core planes 56 through 61 2.13 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:

UFO= (l.O+J *Ue 100.0 where:

UQ =Uncertainty for power peaking factor as defined in equation 5-19 of Analytical Method 3.5.

U e = 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 = uqu *Ue where:

Uqu =Base FQ measurement uncertainty.

= 1.05 Ue = Engineering uncertainty factor.

= 1.03

COLRSALEM 1 Revision 7 January 2016 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR Page 8 of 13 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FN nH (Specification 3.2.3) where:

P THERMAL POWER RATED THERMAL POWER 1.65 2.5.2 PF iJ.J-J

=

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, UFAH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FN AH, shall be the greater of 1.04 or as calculated by the following formula:

um UF!ili

= 1.0 + --

100.0 where:

U AH = 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, UFAH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FN AH shall be calculated by the following formula:

where:

UFt.Hm =Base F11H measurement uncertainty.

= 1.04

COLRSALEM 1 Revision 7 January 2016 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR Page 9 of 13 2.6 Boron Concentration (Specification 3.9.1)

A Mode 6 boron concentration, maintained at or above 2171 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% klk 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.

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 I 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 I 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 I 4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.

COLRSALEM 1 Revision 7 January 2016 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR Page 10 of 13 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-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.

4.0 REFERENCES

1. Salem Nuclear Generating Station Unit No. 1, Amendment No. 310, Renewed License No.

DPR-70, Docket No. 50-272.

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COLRSALEM 1 Revision 7 January 2016 100 80 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR FIGURE2 Page 12 of 13 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER lc-11,90]

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PSEG Nuclear LLC SALEM UNIT 1 CYCLE 25 COLR FIGURE 3 Page 13 of 13 K(Z) -NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 2

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