LR-N12-0255, Update Core Operating Limits Report - Cycle 22
| ML13002A053 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/15/2012 |
| From: | Fricker C Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N12-0255 | |
| Download: ML13002A053 (16) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-M236 0
lENL V, hclea-r LLC AUG 15 ZO0Z Technical Specification 6.9.1.9 LR-N 12-0255 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-001 Salem Nuclear Generating Station Facility Operating License No. DPR-70 NRC Docket No. 50-272
Subject:
Updated Salem Unit I Core Operating Limits Report - Cycle 22 In accordance with section 6.9.1.9 of the Salem Unit 1 Technical Specifications, PSEG Nuclear LLC submits the updated Core Operating Limits Report (COLR) for Salem Unit 1 Cycle 22 in to this letter.
There are no commitments contained in this letter. Should you have any questions regarding this submittal, please contact Mr. D. Lafleur at (856) 339-1754.
Sincerel s
arl
. Fricker Site ice Presi'dent - Salem Attachments (1) ca %ý ý
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Document Control Desk Page 2 LR-N12-0391 cc:
Mr. W. Dean, Administrator, Region I, NRC Mr. J. Hughey, Project Manager, NRC Mr. D. Schroeder, NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Manager IV, NJBNE Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. T. Cachaza, Salem Commitment Tracking Coordinator Mr. D. Lafleur, Senior Regulatory Assurance Engineer LR-N 12-0391 Core Operating Limits Report for Salem Unit 2, Cycle 20
COLR SALEM 2 Revision 4 August 2012 Core Operating Limits Report for Salem Unit 2, Cycle 20 7~
7 Page 1 of 13
COLR SALEM 2 PSEG Nuclear LLC Page 2 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 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 Enthaply 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
COLR SALEM 2 Revision 4 August 2012 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 20 COLR Page 3 of 13 Figure Number 1
2 3
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 Number 11 12 13
COLR SALEM 2 PSEG Nuclear LLC Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 20 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:
3.1.1.3 Moderator Temperature Coefficient 3.1.3.5 Control Rod Insertion Limits 3.2.1 Axial Flux Difference 3.2.2 Heat Flux Hot Channel Factor - FQ(Z) 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor - FNA 3.9.1 Boron Concentration Page 4 of 13
COLR SALEM 2 PSEG Nuclear LLC Page 5 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 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 3.1.1.3) 2.1.1 The Moderator Temperature Coefficient (MTC) limits are:
The BOL/ARO/HZP-MTC shall be less positive than or equal to 0 Al/k/°F.
The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.2x10"4 Ak/k/°F.
2.1,2 The MTC Surveillance limit is:
The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.5x10 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
COLR SALEM 2 Revision 4 August 2012 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 20 COLR Page 6 of 13 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]
FQJ~TP FQ(Z) <
- K(Z) for P > 0.5 FrP FQ(Z) <_ FQRTO
- K(7)forP.<0.5 THERMAL POWER RATED THERMAL POWER 2.4.1 FQRTP = 2.40 2.4.2 K(Z) is provided in Figure 3.
2.4.3 F
=.,j FRTP[10+PF (1.0 -P)j where: FYRI
= 1.95 for unrodded upper core planes 1 through 6 1.78 1.75 1.82 1.80 1.95 2.13 PFxy = 0.3 for unrodded upper core planes 7 through 11 for unrodded upper core planes 12 through 31 for unrodded lower core planes 32 through 50 for um'odded lower core planes 51 through 55 for unrodded lower core planes 56 through 61 for the core planes containing Bank E) control rods
COLR SALEM 2 PSEG Nuclear LLC Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 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=
1.0+
UQ 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, UFQ, to be applied to the Heat Flux Hot Channel Factor FQ(z) shall be calculated by the following formula:
UFQ = Uqu
- Ue where:
Uqu = Base FQ measurement uncertainty.
= 1.05 U, = Engineering uncertainty factor.
= 1.03 Page 7 of 13
COLR SALEM 2 PSEG Nuclear LLC Page 8 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FN (Specification 3.2.3)
FIVAH =FZHr RTP.0 + pFAH (1. O-_ p)J THERMAL POWER where:
P = RATED THERAfLL POWER 2.5.1 FAHRTP
=
1.65 2.5.2 PFAH
= 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, FNAH, shall be the greater of 1.04 or as calculated by the following formula:
UFAH = 1.0 + U' 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, UFAE, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FNAH shall be calculated by the following formula:
UFAH= UrAHn where:
UFAHn, =Base FAH measurement uncertainty.
-1.04
COLR SALEM 2 PSEG Nuclear LLC Page 9 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 2.6 Boron Concentration (Specification 3.9.1)
A Mode 6 boron concentration, maintained at or above 2183 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 (KIf) of 0.95 or less at All Rods In (AR[), Cold Zero Power (CZP) conditions with a 1% Alkk uncertainty added.
b) A Kefr of 0.99 or less at All Rods Out (ARO), CZP conditions with a 1%
Ak/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.
(W 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 (W proprietary).. Methodology for Specification 3/4.2.1 Axial Flux Difference. Approved by Safety Evaluation dated January 31, 1978.
3.3 WCAP-1 0054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985 (W 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 (W proprietary).
Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986.
COLR SALEM 2 PSEG Nuclear LLC Page 10 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 3.5 WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, August 1994 (W 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 (.W 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 (W proprietary). Approved by Safety Evaluation dated September 30, 1999.
4.0 REFERENCES
- 1. Salem Nuclear Generating Station Unit No. 2, Amendment: No. 284, License No.
DPR-75, Docket No. 50-311.
COLR SALEM 2 PSEG Nuclear LLC Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 0-_
e, i,,
0 J
0 F
0 0.
_-1 z0 240 220 200 180 160 140 120 100 80 60 40 20 0
Page 11 of 13 100 30 40 50 60 70 PERCENT OF RATED THERMAL POWER (%)
COLR SALEM Revision 4 August 2012 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 20 COLR Page 12 of 13 FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 0
H a)
C a) a)I-a) 60 40 20 0 1 1 1 1 1 1 1 1 1 1 1 I1 1 1 1 1 1 1 I
-50
-40
-30
-20
-10 0
10 20 30 40 50 Flux Difference (% Delta I)
COLR SALEM 2 Revision 4 August 2012 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 20 COLR Page 13 of 13 FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 1.2 1.0 0.8
(,.)
- 0.6 0.4 z
0.2 0.0 FQ K(Z)
Height (F-)
2.40 1.0 0.0 2.40 1.0 6.0 2.22 0.925 12.0 4
4 0
2 4
6 CORE HEIGHT (FEET) 8 10 12