LR-N18-0106, Core Operating Limits Report - Cycle 24

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Core Operating Limits Report - Cycle 24
ML18309A085
Person / Time
Site: Salem PSEG icon.png
Issue date: 11/05/2018
From: Martino P
Public Service Enterprise Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N18-0106
Download: ML18309A085 (17)
v • d • e


Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 0PSEG NuclearLLC NOV 0 S 2018 Technical Specification 6.9.1.9 LR-N18-0106 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 24 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 24.

There are no commitments contained in this letter.

Should you have any questions regarding this submittal, please contact Mr. Harry Balian at (856) 339-2173.

Sincerely, Patrick A. Martino Plant Manager Salem Generating Station Enclosure cc:

USNRC Regional Administrator-Region 1 US NRC NRR Project Manager-Salem US NRC Senior Resident Inspector-Salem NJ Department of Environmental Protection, Bureau of Nuclear Engineering Commitment Coordinator, Salem Generating Station Corporate Commitment Coordinator, PSEG Nuclear, LLC

Page 2 LR-N18-01 06 (The bee list should not be submitted as part of the DCD submittal-remove this page prior to submittal and make the bee distribution accordingly) bee:

President & Chief Nuclear Officer Site Vice President - Salem Plant Manager-Salem Senior Director, Regulatory Operations & Nuclear Oversight Manager-Nuclear Oversight Director-Regulatory Affairs Manager-Licensing Records Management LR-N18-0106 Enclosure Salem Unit2 Core Operating Limits Report (COLR)

Cycle 24

COLRSALEM 2 Revision 8 September 2018 Core Operating Limits Report for Salem Unit 2, Cycle 24 Page 1 of 14

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 20 18 Section Number 1.0 2.0 2.1 2.2 2.3 2.4 2.5 2.6 3.0 4.0 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR TABLE OF CONTENTS Section Title Table of Contents List of Figures Core Operating Limits Report Operating Limits Moderator Temperature Coefficient (Specification 3.1.1.3)

Control Rod Insertion Limits (Specification 3.1.3.5)

Axial Flux Difference (Specification 3.2.1)

Heat Flux Hot Channel Factor-FQ(z) (Specification 3.2.2 )

Page2 ofl4 Page Number 2

3 4

5 5

6 6

6 Nuclear Enthalpy Rise Hot Channel Factor FN MI (Specification 3.2.3) 8 Boron Concentration (Specification 3.9.1) 9 Analytical Methods 9

References 11

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 Figure Number 1

2 3

PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR LIST OF FIGURES Figure Title Rod Bank Insertion Limits vs. Thermal Power Page 3 of 14 Page Number 12 Axial Flux Difference Limits as a Function of Rated Thermal Power 13 K(z)- Normalized Fq(z) as a Function of Core Height 14

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

TS Section 3.1.1.3 3.1.3.5 3.2.1 3.2.2 3.2.3 3.9.1 COLRSALEM2 RevisionS September 2018 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR 1.0 CORE OPERATING LIMITS REPORT Page 4 of 14 This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 24 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.

Technical Specifications Moderator Temperature Coefficient Control Rod Insetiion Limits Axial Flux Difference Heat Flux Hot Channel Factor - F0(z)

Nuclear Enthalpy Rise Hot Channel Factor - FN AH Boron Concentration COLR Parameter MTC Control Rod Insertion Limits AFD Fo(z)

FNAH Boron Concentration NRC Approved COLR Methodology Section

_(_Section 3.0 Number) 2.1 3.1, 3.6 2.2 3.1, 3.6 2.3 3.1, 3.2, 3.6 3.1, 3.3, 3.4, 3.5, 3.6, 3.7, 3.8, 2.4 3.9 2.5 3.1, 3.5, 3.6, 3.8, 3.9 2.6 3.1, 3.6

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 2.0 OPERATING LIMITS PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR Page 5 ofl4 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 Ak/k/°F.

The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.4x10-4 Aklk/°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 Ak/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 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR 2.2 Control Rod Insertion Limits (Specification 3.1.3.5)

Page 6 of 14 2.2.I The control rod banks shall be limited in physical insertion as shown in Figure I.

2.3 Axial Flux Difference (Specification 3.2.1)

[Constant Axial Offset Control (CAOC) Methodology]

2.3.I 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) :::; FQRTP p

FQ(z) :::; FQRTP

0.5 where

P

  • K(z) for P > 0.5
  • K(z) for P ::;0.5 THERMAL POWER RATED THERMAL POWER 2.4.1 FlTP

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

2.4.3 Fx/

= FxyRTP {J.O + PFxy (J.O-P)}

where: from BOL to I 0000 MWD/MTU F xy RTP

= 2. 03 for unrodded upper core planes 1 through 6

2. 00 for unrodded upper core planes 7 through 8
1. 83 for unrodded upper core planes 9 through 11
1. 7 6 for unrodded upper core planes 12 through 13
1. 7 6 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. 85 for unrodded lower core planes 44 through 48 1.92 for unrodded lower core planes 49 through 50 1.89 for unrodded lower core planes 51 through 53
      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR 2.00for unrodded lower core planes 54 through 55

2. 03 for unrodded lower core planes 56 through 61
2. 07 for the core planes containing Bank D control rods PFxy

= 0.3 where: from 10000 MWD/MTU to 14000 MWD/MTU F xy RTP

= 2. 03 for unrodded upper core planes 1 through 6

1. 90 for unrodded upper core planes 7 through 8
1. 80 for unrodded upper core planes 9 through 11
1. 7 6 for unrodded upper core planes 12 through 13 1.76for unrodded uppercore planes 14 through 18
1. 88 for unrodded upper core planes 19 through 31 1.91 for unrodded lower core planes 32 through 43 1.85 for unrodded lower core planes 44 through 48 1.88for unrodded lower core planes 49 through 50 1.83 for unrodded lower core planes 51 through 53 1.89 for unrodded lower core planes 54 through 55 2.03 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 xy RTP

= 2. 03 for unrodded upper core planes 1 through 6

1. 90 for unrodded upper core planes 7 through 8 1.80 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 1.98 for unrodded lower core planes 32 through 43
1. 86 for unrodded lower core planes 44 through 48 1.87 for unrodded lower core planes 49 through 50 1.82for unrodded lower core planes 51 through 53 1.86 for unrodded lower core planes 54 through 55
2. 03 for unrodded lower core planes 56 through 61
2. 0 7 for the core planes containing Bank D control rods PFxy 0.3 Page 7 of 14
      • This record was final approved on 9/14/2018 11:58:43 AM. {This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR Page 8 of14 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:

where:

U0 = Uncertainty for power peaking factor as defined in equation 5-19 of Analytical Method 3.5.

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 F0(z) shall be calculated by the following formula:

UFQ=Vqu*Ve where:

Uqu = Base FQ measurement uncertainty.

= 1.05

u. = Engineering uncertainty factor.

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

P 2.5.1 ji<TP,:j}[

THERMAL POWER RATED THERMAL POWER 1.65 2.5.2 PF &

=

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, Up,m, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FN

,m, shall be the greater of 1.04 or as calculated by the following formula:

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR U F!J.H

= 1.0 +

U !J.H 100.0 Page 9 of 14 where:

UMI = 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, Up,m, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FN MI shall be calculated by the following formula:

where:

UFMim =Base F t.H measurement uncertainty.

= 1.04 2.6 Boron Concentration (Specification 3.9.1)

A Mode 6 boron concentration, maintained at or above 2051 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%

Lik/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

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 20 18 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR Page 10 of14 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.

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.

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018

4.0 REFERENCES

PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR Page 1 1 of14

1. Salem Nuclear Generating Station Unit No. 2, Amendment No. 301, Renewed License No. DPR-75, Docket No. 50-311.
      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR FIGURE 1 Page 12 of14 ROD BANK INSERTION LIMITS VS. THERMAL POWER 228 200 180

'C" I

111G I

14G A.

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0 ii 120 0

II.

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i IQ 100 0

0 C) 80 80 20 0

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10 20 G

a a

n PERCENTDFRATEDlHERMALPOWER(%)

80 f100, 170 l

/

J v v

80 100

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 20 18 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR FIGURE2 Page 13 of 14 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 1-;

ll) 0

§ ll)

K

'"0 ll) 1;l

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J 100 80 60 40 20 0

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UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION

!ACCEPTABLE OPERATION 1/

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ic-31,50) 1 1 (31,50) 1

-40

-30

-20

-10 0

10 20 30 40 Flux Difference (% Delta I) 50

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)

COLRSALEM2 Revision 8 September 2018 1.2 1.0

§:

0.8 0

E--<

u c.'J

)

--< 0.6 I



0 0.4 z

0.2 0.0 0

PSEG Nuclear LLC SALEM UNIT 2 CYCLE 24 COLR FIGURE3 Page 14 of 14 K(Z)- NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 2

FQ K(Z) 2.40 1.0 2.40 1.0 2.22 0.925 4

6 8

CORE HEIGHT (FEET)

Height (FT) 0.0 6.0 12.0 10 12

      • This record was final approved on 9/14/2018 11:58:43 AM. (This statement was added by the PRIME system upon its validation)
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