LR-N19-0067, Core Operating Limits Report - Cycle 27

Core Operating Limits Report - Cycle 27

ML19161A196
Person / Time
Site:	Salem
Issue date:	06/10/2019
From:	Martino P Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N19-0067
Download: ML19161A196 (16)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 J

LR-N 19-0067 Technical Specification 6.9.1.9 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

Subject:

Salem Unit 1 Core Operating Limits Report - Cycle 27 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 27.

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-N 19-0067 cc: Administrator, Region 1 NRG Project Manager-Salem NRG Senior Resident Inspector, Salem Manager, NJBNE Salem Commitment Tracking Coordinator Corporate Commitment Tracking Coordinator

LR-N19-0067 Enclosure Salem Unit 1 Core Operating Limits Report (COLR)

Cycle 27

COLRSAL EM 1 Revision 10 December 2018 Core Operating Limits Report for Salem Unit 1, Cycle 27 Page 1 of 13

COLRSALEM 1 PSEG Nuclear LLC Page2 of13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 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 FNMI (Specification 3.2.3) 8 2.6 Boron Concentration (Specification 3.9.1) 9 3.0 Analytical Methods 9 4.0 References 10

COLRSALEM l PSEG Nuclear LLC Page 3 of 13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 LIST OF FIGURES Figure Figure Title Page Number Number 1 Rod Bank Insertion Limits vs. Thermal Power 11 2 Axial Flux Difference Limits as a Function of Rated Thermal Power 12 3 K( z) - Normalized F 0( z) as a Function of Core Height 13

COLRSALEM 1 PSEG Nuclear LLC Page4 of13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 1 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) 3.1.1.4 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, 3.2.2 FQ(Z) 2.4 Fo(Z) 3.6, 3.7, 3.8 Nuclear Enthalpy Rise Hot Channel FNm 3.1, 3.5, 3.6, 3.2.3 2.5 Factor - FNm 3.7, 3.8 3.9.1 Boron Concentration Boron Concentration 2.6 3.1, 3.6

COLRSALEM 1 PSEG Nuclear LLC Page 5 ofl3 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 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 .1.1.4) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:

The BOL/ARO/HZP-MTC shall be less positive than or equal to O ~k/k/°F.

The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.4xl0-4 ~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 -3.7xl0-4 ~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 PSEG Nuclear LLC Page 6 of 13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 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) ~

• K(Z) for P ::; 0. 5 0.5 THERMAL POWER where: p RATED THERMAL POWER 2.4.2 K(Z) is provided in Figure 3.

where: from BOL to 10000 MWD/MTU F

xy RTP = 2. 03 for umodded upper core planes 1 through 6 1.89 for umodded upper core planes 7 through 8 1.80 for umodded upper core planes 9 through 11

1. 75 for umodded upper core planes 12 through 13

1. 76 for umodded upper core planes 14 through 18 1.80 for umodded upper core planes 19 through 31 1.80 for umodded lower core planes 32 through 43 1.84 for umodded lower core planes 44 through 48 1.90 for umodded lower core planes 49 through 50 1.90 for umodded lower core planes 51 through 53

COLRSALEM 1 PSEG Nuclear LLC Page 7 of 13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 2.00 for unrodded lower core planes 54 through 55

2. 03 for umodded lower core planes 56 through 61

2. 03 for the core planes containing Bank D control rods

0.3 where

from 10000 MWD/MTU to 14000 MWD/MTU FxyRTF = 2. 03 for umodded upper core planes I through 6 1.83 for umodded upper core planes 7 through 8

1. 79 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 1.88 for unrodded upper core planes 19 through 31 1.90 for unrodded lower core planes 32 through 43 1.87 for unrodded lower core planes 44 through 48 1.92 for unrodded lower core planes 49 through 50

1. 84 for unrodded lower core planes 51 through 53 1.90 for unrodded lower core planes 54 through 55 2.03 for unrodded lower core planes 56 through 61

2. 03 for the core planes containing Bank D control rods PFxy = 0.3 where: from 14000 MWD/MTU to EOL FxyRTF = 2. 00 for unrodded upper core planes 1 through 6

1. 78 for umodded upper core planes 7 through 8

1. 73 for unrodded upper core planes 9 through 11

1. 78 for unrodded upper core planes 12 through 13 1.80 for unrodded upper core planes 14 through 18 1.98 for unrodded upper core planes 19 through 31 1.98 for unrodded lower core planes 32 through 43

1. 88 for unrodded lower core planes 44 through 48 1.91 for unrodded lower core planes 49 through 50 1.81 for unrodded lower core planes 51 through 53 1.86 for unrodded lower core planes 54 through 55 1.96 for unrodded lower core planes 56 through 61

2. 03 for the core planes containing Bank D control rods 0.3

COLRSALEM 1 PSEG Nuclear LLC Page 8 of13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 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:

Upa =(1.0+ uQ

- 100.0

)*u.

where:

UQ = Uncertainty for power peaking factor as defined in equation 5-19 of Analytical Method3.5.

u. = 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:

uFQ = uqu *U.

where:

Uqu = Base FQ measurement uncertainty.

= 1.05 U 0 = Engineering uncertainty factor.

= 1.03 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FN&I (Specification 3.2.3)

THERMAL POWER where: P RATED THERMAL POWER 1.65 2.5.2 PFL1H 0.3

COLRSALEM 1 PSEG Nuclear LLC Page 9 of13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 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, UF,rn, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FN,rn, shall be the greater of 1.04 or as calculated by the following formula:

UFAff = 1.0 +UM!--

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 FNAH shall be calculated by the following formula:

where: UFAHm = Base F AH measurement uncertainty.

= 1.04 2.6 Boron Concentration (Specification 3 .9 .1)

A Mode 6 boron concentration, maintained at or above 2050 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) AK-effective (K.,ff) of 0.95 or less at All Rods In (ARI), Cold Zero Power (CZP) conditions with a 1% Lik/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 PSEG Nuclear LLC Page 10 of 13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 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 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 / 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.

3.8 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. 1, up to Amendment No. 324, Renewed License No. DPR-70, Docket No. 50-272.

COLRSALEM 1 PSEG Nuclear LLC Page 11 of 13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 240 220 V II" 116.3, 2251

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COLRSALEM 1 PSEG Nuclear LLC Page 12 of13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 FIGURE2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 k-11,90J 1(11,90)1 80 UNACCEPTABLE I \ UNACCEPTABLE OPERATION OPERATION

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COLRSALEM 1 PSEG Nuclear LLC Page 13 of 13 Revision 10 SALEM UNIT 1 CYCLE 27 COLR December 2018 FIGURE3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 1.2 1.0 FQ K(Z) Height (F1) 2.40 1.0 0.0 2.40 1.0 6.0 2.22 0.925 12.0 0.2 0.0 0 2 4 6 8 10 12 CORE HEIGHT (FEET)