LR-N11-0124, Core Operating Limits Report - Cycle 19
| ML11130A070 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/27/2011 |
| From: | Fricker C Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N11-0124 | |
| Download: ML11130A070 (16) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0
PSEG Nuclear LLC APR 2 7 2011 Technical Specification 6.9.1.9 LR-Nl1-0124 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555-001 Salem Nuclear Generating Station Unit 2 Facility Operating License No. DPR-75 NRC Docket No. 50-311
SUBJECT:
CORE OPERATING LIMITS REPORT - CYCLE 19 In accordance with section 6.9.1.9 of the Salem Unit 2 Technical Specifications, PSEG
-Nuclear LLC submits the Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 19 in Attachment 1 to this letter.
There are no regulatory commitments contained in this letter. Should you have any questions regarding this submittal, please contact Mr. E. Villar at (856) 339-5456.
- Sincerel Ca Fricker Site Vice President - Salem Attachments (1) 95-2168 REV. 7/99
Document Control Desk APR 2 7 2011 Page 2 LR-N 11-0124 C
Mr. William Dean, Administrator - Region I Mr. R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector - Salem Mr. P. Mulligan, Manager IV Commitment Coordinator Salem PSEG Corporate Commitment Manager LR-N 11-0124 Core Operating Limits Report For Salem Unit 2, Cycle 19
COLR SALEM 2 Revision 3 January 2011 Core Operating Limits Report for Salem Unit 2, Cycle 19 V
Page 1 of 13
COLR SALEM 2 PSEG Nuclear LLC Revision 3 SALEM UNIT 2 CYCLE 19 COLR Page 2 of 13 January 2011 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
COLR SALEM 2 Revision 3 January 2011 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 19 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 3 SALEM UNIT 2 CYCLE 19 COLR Page 4 of 13 January 2011 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report (COLR) for Salem Unit 2 Cycle 19 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 - FAHN 3.9.1 Boron Concentration
COLR SALEM 2 PSEG Nuclear LLC Revision 3 SALEM UNIT 2 CYCLE 19 COLR Page 5 of 13 January 2011 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 Akik/'F.
The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.2x10 4 Ak/k/0F.
2.1.2 The MTC Surveillance limit is:
The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -3.5x 10-4 Ak/k/0F.
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 PSEG Nuclear LLC Revision 3 SALEM UNIT 2 CYCLE 19 COLR Page 6 of 13 January 2011 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)
EE
- K(Z)forP>0.5 FQRTP FQ(Z) 05
- K(Z)forP*S 0.5 FoZ -0.5 THERMAL POWER where:
P
-RATED THERMAL POWER 2.4.1 FQRTP 2.40 2.4.2 K(Z) is provided in Figure 3.
2.4.3 FL = FyRTP[1.0+PFy(1.0-P)]
where: FxyRTP = 1.95 for unrodded upper core planes 1 through 6 1.73 for unrodded upper core planes 7 through 11 1.72 for unrodded upper core planes 12 through 31 1.82 for unrodded lower core planes 32 through 50
COLR SALEM 2 PSEG Nuclear LLC Revision 3 SALEM UNIT 2 CYCLE 19 COLR Page 7 of 13 January 2011 1.80 for unrodded lower core planes 51 through 55 1.95 for unrodded lower core planes 56 through 61 2.13 for the core planes containing Bank D control rods PFy = 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 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:
UF=k(1.0+ UQ-U
\\
10.0) 0U where:
UQ = Uncertainty for power peaking factor as defined in equation 5-19 of Analytical Method 3.5.
Ue = 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 OUe where:
Uqu = Base FQ measurement uncertainty.
= 1.05 Ue = Engineering uncertainty factor.
=
1.03
COLR SALEM 2 PSEG Nuclear LLC Revision 3 SALEM UNIT 2 CYCLE 19 COLR Page 8 of 13 January 2011 2.5 Nuclear Enthalpy Rise Hot Channel Factor - FNAH (Specification 3.2.3)
FNvAH
=
FAHRTP [1.0 + PFAH (1.0- P)]
THERMAL POWER RATED THERMAL POWER 2.5.1 FIHRp =
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, FAHN, using the PDMS shall be the greater of 1.04 or as calculated by the following formula:
UFA_ = 1.0+ UA 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 FAHN shall be calculated by the following formula:
UFAH = UFAHm where:
UFAHm =Base FAH measurement uncertainty.
= 1.04
COLR SALEM 2 PSEG Nuclear LLC Revision 3 SALEM UNIT 2 CYCLE 19 COLR Page 9 of 13 January 2011 2.6 Boron Concentration (Specification 3.9.1)
A Mode 6 boron concentration, maintained at or above 2184 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 (Kff-) of 0.95 or less at All Rods In (ARI), Cold Zero Power (CZP) conditions with a 1% Ak/k uncertainty added.
b) A Keff 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-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using 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.
COLR SALEM 2 PSEG Nuclear LLC Revision 3 SALEM UNIT 2 CYCLE 19 COLR Page 10 of 13 January 2011 3.4 WCAP-10266-P-A, Rev. 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.
3.5 WCAP-12472-P-A, BEACON - Core Monitoring and Operations Support System, Revision 0 (W proprietary). Approved February 1994.
3.6 CENPD-397-P-A, Rev. 1, Improved Flow Measurement Accuracy Using Crossflow Ultrasonic Flow Measurement Technology, May 2000.
3.7 WCAP-10054-P-A, Addendum 2, Rev. 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.
4.0 REFERENCES
- 1. Salem Nuclear Generating Station Unit No. 2, Amendment No. 277, License No.
DPR-75, Docket No. 50-311.
COLR SALEM 2 Revision 3 January 2011 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 19 COLR Page 11 of 13 FIGURE 1 240 220 200 180 160 g-140 Uf) z0 t-120 UI) 0 z< 100 m
0 I-z 80 0
0 ROD BANK INSERTION LIMITS VS. THERMAL POWER 18.3, 230762 BANKANKI I Ii I I NK CI
! 11 1 1 1 1/7--
BANK C
I-IIII 0
I58'i 10
,!/T7i 60 40 20 0
0 10 20 30 40 50 60 70 PERCENT OF RATED THERMAL POWER (%)
80 90 100
COLR SALEM 2 Revision 3 January 2011 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 19 COLR Page 12 of 13 FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 80 0
cc H-o 5) cc 0
5.)
0 5.)
60 40 20 0
I._._._.
L. LL. L L
-50
-40
-30
-20
-10 0
10 20 Flux Difference (% Delta I) 30 40 50
COLR SALEM 2 Revision 3 January 2011 PSEG Nuclear LLC SALEM UNIT 2 CYCLE 19 COLR Page 13 of 13 FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 1.2 1.0 0
C-)
0.8 0.6 0.4 FQ K(Z)
Height (FT) 2.40 1.0 0.0 2.40 1.0 6.0 2.22 0.925 12.0 0.2 4
4
+
4 4
0.0 L 0
2 4
6 8
CORE HEIGHT (FEET) 10 12