LR-N17-0101, Core Operating Limits Report, Cycle 23
| ML17136A378 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/16/2017 |
| From: | Grover F Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N17-0101 | |
| Download: ML17136A378 (16) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 0PSEG
,Vurlear Ll.C LR-N17-0101 Technical Specification 6.9.1.9 MAY 16.2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-0001
Subject:
Salem Generating Station Unit 2 Renewed Facility Operating License DPR-75 NRC Docket No. 50-311 Salem Unit 2 Core Operating Limits Report-Cycle 23 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 23.
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, F. Kenneth Gr v Plant Manager Salem Generati tjc Enclosure
Page2 LR-N17-0101 cc:
Mr. D. Dorman, US NRC -Administrator-Region 1 Mr. R. Ennis, 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 Mr. M. Washington, Chief Inspector-Occupational Safety and Health Bureau of Boiler and Pressure Vessel Compliance
LR-N16-01 07 Enclosure Salem Unit 2 Core Operating Limits Report {COLR)
Cycle 23
COLR SALEM 2 Revision 7 March 2017 Core Operating Limits Report for Salem Unit 2, Cycle 23 Page 1 of 13
COLRSALEM2 PSEG Nuclear LLC Revision 7 SALEM UNIT 2 CYCLE 23 COLR Page 2 of13 March 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 3.1.1.3) 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 FNAH (Specification 3.2.3) 8 2.6 Boron Concentration (Specification 3.9.1) 9 3.0 Analytical Methods 9
4.0 References 10
COLRSALEM2 Revision 7 March 2017 Figure Number 1
2 3
PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR LIST OF FIGURES Figure Title Rod Bank Insertion Limits vs. Thermal Power Page 3 of13 Page Number 11 Axial Flux Difference Limits as a Function of Rated Thermal Power 12 K(z)- Normalized F q(z) as a Function of Core Height 13
TS Section 3.1.1.3 3.1.3.5 3.2.1 3.2.2 3.2.3 3.9.1 COLRSALEM2 Revision 7 March 2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR 1.0 CORE OPERATING LIMITS REPORT Page 4 ofl3 This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 23 has been prepared in accordance with the requirements of Teclmical 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 COLR Methodology Technical Specifications COLR Parameter Section (Section 3.0 Number)
Moderator Temperature Coefficient MTC 2.1 3.1, 3.6 Control Rod Insertion Limits Control Rod Insertion Limits 2.2 3.1, 3.6 Axial Flux Difference AFD 2.3 3.1, 3.2, 3.6 3.1, 3.3, 3.4, 3.5, 3.6, 3.7, Heat Flux Hot Channel Factor-F_Q_(z)
Fo(z) 2.4 3.8 Nuclear Enthalpy Rise Hot Channel Factor - FN M-I FN!lH 2.5 3.1, 3.5, 3.6, 3.8 Boron Concentration Boron Concentration 2.6 3.1, 3.6
COLRSALEM2 Revision 7 March2017 2.0 OPERATING LIMITS PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 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 3.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:
The BOLIARO/HZP-MTC shall be less positive than or equal to 0 ?k/k/°F.
The EOLIARO/RTP-MTC shall be less negative than or equal to -4.4x10-4
?k/k/oF.
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
COLRSALEM2 Revision 7 March 2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 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]
FlTP FQ(z) :::;
p
- K(z) for P > 0.5 FQRTP F Q(z) :::;
- K(z) for P :'{ 0. 5
0.5 where
P THERMAL POWER RATED THERMAL POWER 2.4.2 K(z) is provided in Figure 3.
where: from BOL to 12000 MWD/MTU RTF Fxy
=
- 2. 03 for unrodded upper core planes I through 6 I. 83 for unrodded upper core planes 7 through 8 I. 7 4 for unrodded upper core planes 9 through I8 I. 80 for unrodded upper core planes I9 through 3I I.82for unrodded lower core planes 32 through 43 I. 83 for unrodded lower core planes 44 through 53 I.90 for unrodded lower core planes 54 through 55
- 2. 03 for unrodded lower core planes 56 through 6I 2.I3 for core planes containing Bank D control rods PFxy 0.3
COLRSALEM2 Revision 7 March 2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR Page 7 of13 where: from 12000 MWD/MTU to EOL Fx/TP
=
I. 98 for unrodded upper core planes I through 6 I. 80 for unrodded upper core planes 7 through 8 I. 77 for unrodded upper core planes 9 through I8 I. 9 2 for unrodded upper core planes I9 through 3I I.95 for unrodded lower core planes 32 through 43 I. 79 for unrodded lower core planes 44 through 53 I. 84 for unrodded lower core planes 54 through 55 I.95 for unrodded lower core planes 56 through 6I 2.I3 for 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 Teclmical 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
= (1.0 +
U Q J
- U e 100.0 where:
Uq =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, U FQ, to be applied to the Heat Flux Hot Channel Factor Fq(z) shall be calculated by the following formula:
U FQ = Uqu e U, where:
Uqu =Base Fq measurement uncertainty.
= 1.05
- u. = Engineering uncertainty factor.
= 1.03
COLRSALEM2 Revision 7 March 2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNL1H (Specification 3.2.3) where:
P THERMAL POWER RATED THERMAL POWER 1.65 2.5.2 PF t1H
=
0.3 Page 8 of13 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, UFll-h to be applied to the Nuclear Enthalpy Rise Hot Cham1el Factor, FN !lH, shall be the greater of 1.04 or as calculated by the following formula:
utili UF!ili = 1.0+
100.0 where:
ullH = 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 upcertainty, UF!lH, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FN 1\\H shall be calculated by the following formula:
where:
UFI\\Hm =Base F 1\\H measurement uncertainty.
= 1.04
COLRSALEM2 Revision 7 March2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR Page 9 of13 2.6 Boron Concentration (Specification 3.9.1)
A Mode 6 boron concentration, maintained at or above 2222 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 (K.ff) 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.
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.
COLRSALEM2 Revision 7 March2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR Page 10 of13 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 Febmary 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.
4.0 REFERENCES
- 1. Salem Nuclear Generating Station Unit No. 2, Amendment No. 297, Renewed License No. DPR-75, Docket No. 50-311.
COLRSALEM2 Revision 7 March 2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR Page 11 ofl3 240 220 200
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i=
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FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER L
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COLRSALEM2 Revision 7 March2017 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR FIGURE 2 Page 12 ofl3 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER
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COLRSALEM2 Revision 7 March 2017 1.2 1.0 0.8 0
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0 0.4 z
0.2 0.0 0
PSEG Nuclear LLC SALEM UNIT 2 CYCLE 23 COLR FIGURE 3 Page 13 of13 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