LR-N17-0166, Core Operating Limits Report - Cycle 26

Core Operating Limits Report - Cycle 26

ML17306B212
Person / Time
Site:	Salem
Issue date:	11/02/2017
From:	Martino P Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N17-0166
Download: ML17306B212 (18)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 0PSEG Nur:lear LLC LR-N17-0166 Technical Specification 6.9.1.9 NOV 02 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-0001

Subject:

Salem Generating Station Unit 1 Renewed Facility Operating License DPR-70 NRC Docket No. 50-272 Salem Unit 1 Core Operating Limits Report-Cycle 26 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 26.

There are no commitments contained in this letter.

Should you have any questions regarding this submittal, please contact Mr. Thomas Cachaza at 856-339-5038.

Sincerely, Patrick Martino Plant Manager Salem Generating Station tjc Enclosure

Page 2 LR-N17-0166 cc:

Mr. D. Dorman, US NRC -Administrator-Region 1 Ms. C. Parker, USNRC -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

LR-N17-0166 Enclosure Salem Unit 1 Core Operating Limits Report (COLR)

Cycle 26

COLR SALEM 1 Revision 8 August 2017 Core Operating Limits Report for Salem Unit 1, Cycle 26 Page 1 ofl3

COLRSALEM 1 PSEG Nuclear LLC Page 2 of 13 Revision 8 SALEM UNIT 1 CYCLE 26 COLR August 2017 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-F0(z) (Specification 3.2.2) 6 2.5 Nuclear Enthalpy Rise Hot Channel Factor FN,.,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 8 August 2017 Figure Number 1

2 3

PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 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)- Nonnalized FQ(z) as a Function of Core Height Page Number 11 12 13

TS COLRSALEM 1 Revision 8 August 2017 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 COLR 1.0 CORE OPERATING LIMITS REPORT Page 4 of13 This Core Operating Limits Report (COLR) for Salem Unit 1 Cycle 26 has been prepared in accordance with the requirements ofTechnical 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 detennine 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-F0(Z)

F0(Z) 2.4 3.1, 3.3, 3.4, 3.5, 3.6, 3.7 3.2.3 Nuclear Enthalpy Rise Hot Channel FNI'IH 2.5 3.1, 3.5, 3.6, 3.7 Factor-FN l'IH 3.9.1 Boron Concentration Boron Concentration 2.6 3.1, 3.6

COLRSALEM 1 Revision 8 August 2017 2.0 OPERATING LIMITS PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 COLR Page 5 of13 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 11k/kfOF.

The EOLIARO/RTP-MTC shall be less negative than or equal to -4.4xl0-4.!1.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.!1k/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 I Revision 8 August 2017 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 COLR 2.2 Control Rod Insertion Limits (Specification 3.1.3.5)

Page 6 of13 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)

s; p

• K(Z) for P > 0.5 FQRTP FQ(Z) :s;

• K(Z) for P 5{0.5

0.5 where

P THERMAL POWER RATED THERMAL POWER 2.4.1 FlTP

=

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

where: from BOL to 10000 MWD/MTU F

RTP =

xy 1.96 for unrodded upper core planes 1 through 6

1. 76 for unrodded upper core planes 7 through 8

1. 70 for unrodded upper core planes 9 through 11 1.69 for unrodded upper core planes 12 through 13

1. 70 for unrodded upper core planes 14 through 18

1. 7 5 for unrodded upper core planes 19 through 31

1. 76 for umodded lower core planes 32 through 43 1.81 for unrodded lower core planes 44 through 48 1.81 for unrodded lower core planes 49 through 50 1.81 for unrodded lower core planes 51 through 53

COLRSALEM 1 Revision 8 August 2017 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 COLR 1.83 for unrodded lower core planes 54 through 55 2.00 for unrodded lower core planes 56 through 61 2.07 for the core planes containing Bank D control rods

0.3 where

from 10000 MWD/MTU to 14000 MWD/MTU F RTP

=

xy 1.96 for unrodded upper core planes 1 through 6

1. 79 for unrodded upper core planes 7 through 8

1. 70 for unrodded upper core planes 9 through 11

1. 70 for unrodded upper core planes 12 through 13

1. 7 4 for unrodded upper core planes 14 through 18 1.87 for unrodded upper core planes 19 through 31 1.90 for unrodded lower core planes 32 through 43 1.80 for unrodded lower core planes 44 through 48

1. 79 for unrodded lower core planes 49 through 50 1.77 for unrodded lower core planes 51 through 53

1. 7 8 for unrodded lower core planes 54 through 55 1.96 for unrodded lower core planes 56 through 61

2. 07 for the core planes containing Bank D control rods PFxy

=

0.3 where

from 14000 MWD/MTU to EOL F RTP

=

'Y 1.96 for unrodded upper core planes 1 through 6

1. 79 for unrodded upper core planes 7 through 8

1. 71 for unrodded upper core planes 9 through 11

1. 71 for unrodded upper core planes 12 through 13

1. 78 for unrodded upper core planes 14 through 18 1.90 for unrodded upper core planes 19 through 31 1.90 for unrodded lower core planes 32 through 43 1.83 for unrodded lower core planes 44 through 48

1. 77 for unrodded lower core planes 49 through 50

1. 73 for unrodded lower core planes 51 through 53

1. 72 for unrodded lower core planes 54 through 55 1.82 for unrodded lower core planes 56 through 61

2. 07 for the core planes containing Bank D control rods 0.3 Page 7 of13

COLRSALEM 1 Revision 8 August 2017 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 COLR Page 8 of 13 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 fonnula:

UFQ =(1.0+ UQ )*Ue 100.0 where:

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

De = Engineering uncertainty factor.

= 1.03 Note: UFQ= PDMS Surveillance Report Core Monitor Fxy Uncertainty in%.

2.4.5 If the IN CORE 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 fonnula:

UFQ = uqll *Ue where:

Uqu =Base FQ measurement uncertainty.

= 1.05 De = Engineering uncertainty factor.

= 1.03 2.5 Nuclear Enthalpy Rise Hot Channel Factor-FN ~H (Specification 3.2.3) where:

P THERMAL POWER RATED THERMAL POWER 1.65 2.5.2 PFLJH 0.3

COLRSALEM 1 Revision 8 August 2017 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 COLR Page 9 of 13 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, FNAH, shall be the greater of 1.04 or as calculated by the following formula:

u

= 1.0+ UAH FAH 100.0 where:

UAH = 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 llH shall be calculated by the following formula:

where: UFilHm = Base F AH measurement uncertainty.

= 1.04 2.6 Boron Concentration (Specification 3.9.1)

A Mode 6 boron concentration, maintained at or above 2133 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.

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.

COLRSALEM 1 Revision 8 August 2017 PSEG Nuclear LLC SALEM UNIT 1 CYCLE 26 COLR Page 10 of 13 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.

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. 318, Renewed License No.

DPR-70, Docket No. 50-272.

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