ML12339A006
| ML12339A006 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/29/2012 |
| From: | Fricker C Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML12339A006 (24) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 PSEG NuclearLLC Technical Specification 6.9.1.9 LR-N 12-0391 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-001 Salem Nuclear Generating Station Facility Operating License No. DPR-75 NRC Docket No. 50-311
Subject:
Salem Unit 2 Core Operating Limits Report - Cycle 20 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 20 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.
Sincerely, Carl J. Fricker Site Vice President - Salem Attachments (1)
Aw
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-N12-0391 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 V
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 PSEG Nuclear LLC Page 3 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 LIST OF FIGURES Figure Figure Title Page Number Number I 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 FQ(z) as a Function of Core Height 13
COLR SALEM 2 PSEG Nuclear LLC Page 4 of 13 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 - FNAll 3.9.1 Boron Concentration
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 Ak/1ldF.
4 The EOL/ARO/RTP-MTC shall be less negative than or equal to -4.2x10" 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] 0-4 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 PSEG Nuclear LLC Page 6 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 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]
FQ(Z) --
- K(Z)forP>0.5 FRP F QRrP'0 05
- K(Z)for P 5-<
where: P THERMAL POWER RATED THERMAL POWER 2.4.1 FQl' 2.40 2.4.2 K(Z) is provided in Figure 3.
2.4.3 FpyL = F.RTP [1.O + PFO(1.O- P)I where: FVR'P 1.95 for unrodded upper core planes 1 through 6 1.78 for unrodded upper core planes 7 through 11 1.75 for unrodded upper core planes 12 through 31 1.82 for unrodded lower core planes 32 through 50 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 PFIy = 0.3
COLR SALEM 2 PSEG Nuclear LLC Page 7 of 13 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:
o- 1.00+ .0)U 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:
Um ý Uq,
- U, where:
Uqu = Base FQ measurement uncertainty.
= 1.05 U, = Engineering uncertainty factor.
= 1.03
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 - FNAH (Specification 3.2.3)
F"AIH "= F'dH TP [1. 0 +I PF'dtH (1. O-_ P,).
THERMAL POWER RA TED THERMAL POWER 2.5.1 FAHRTP = 1.65 2.5.2 PFu = 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, UFAFI, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor, FNAu, shall be the greater of 1.04 or as calculated by the following formula:
UA = 1.0 + U&H 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, UFMA, to be applied to the Nuclear Enthalpy Rise Hot Channel Factor FNAH shall be calculated by the following formula:
UFAH - FAFIN where:
UFAHml =Base FAHi 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 (IKc)ff of 0.95 or less at All Rods In (ARI), Cold Zero Power (CZP) conditions with a 1% Akdk uncertainty added.
b) A KFf 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 ( 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 (Wý 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 Chanmel 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 (& 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 COST 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 Page 11 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 FIGURE 1 ROD BANK INSERTION LIMITS VS. THERMAL POWER 240 220 200 180 160 ui 140 di z
0 E 120 0
S00 0ý I-.
z 0
0 80 60 40 20 0
0 10 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER (%)
COLR SALEM 2 PSEG Nuclear LLC Page 12 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 100 80 C
a) 60 0
a)
C 40-C.)
I-a) 20 0 1 1 1 I I I I l I Ii ,Ii i I
-50 -40 -30 -20 -10 0 10 20 30 40 50 Flux Difference (% Delta I)
COLR SALEM 2 PSEG Nuclear LLC Page 13 of 13 Revision 4 SALEM UNIT 2 CYCLE 20 COLR August 2012 FIGURE 3 K(Z) - NORMALIZED FQ(Z) AS A FUNCTION OF CORE HEIGHT 1.2 1.0 0.8 0
6.
- 0.6 0.4 z
0.2 0.0 0 2 4 6 8 10 12 CORE HEIGHT (FEET)
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC 10 CFR Part 50 Appendix E NOV 2 8 2012 10 CFR 50.4(b)(5)
LR-N 12-0366 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Nuclear Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354
Subject:
Salem Hope Creek Emergency Plan Evacuation Time Estimate Pursuant to part 50.4 of Title 10 of the Code of Federal Regulations, PSEG Nuclear LLC
-(PSEG)_hereby-submits the Salem -Hope-Creek-evacuation-time. estimate_(ETE)_analysis._- This-study is submitted in accordance with the requirements of Section IVof Appendix E to 10 CFR 50 to submit an updated study "within 365 days of the later of the date of the availability of the most recent decennial census data from the U.S. Census Bureau or December 23, 2011."
As part of this submittal please find one attachment and one enclosure. The 10 CFR 50.54(q)
Summary Analysis Report, document 2012-35, is attached. Enclosure 1 is the full Salem-Hope Creek ETE study.
There are no regulatory commitments contained in this submittal. The Salem - Hope Creek ETE study submittal is made as required by regulation. This study provides the methods used to derive calculated time for public evacuation. This study is not an implementing procedure or a plan commitment, but provides an important part of the bases for development of protective action recommendations.
LR-N12-0366 N0V 2 8 .']2 Page 2 If you have any questions or require additional information, please contact Michelle Patti, Emergency Services Regulatory Compliance Manager, at 856-339-3691 or Craig Banner, Emergency Preparedness Program Manager, at 856-339-1157.
Respectfully, David Burgin Manager - Emergency Preparedness 0 CFR 50.54(q)
SUMMARY
ANALYSIS REPORT # 2012-35: Evacuation Time Estimate Study, KLD TR-499 Revision 0 - Salem - Hope Creek Evacuation Time Estimate Study w/o enclosure C: W. Dean, Administrator, Region I, NRC J. Hughey, Project Manager - USNRC NRC Senior Resident Inspector - Salem NRC Senior Resident Inspector - Hope Creek P. Mulligan, Manager IV, NJBNE J. Turner, Director DEMA TCachaza Commitment-Tracki nCoord iator--Salem .
-T. . . . .
P. Bonnett - Commitment Tracking Coordinator - Hope Creek L. Marabella - Corporate Commitment Tracking Coordinator
LR-N12-0366 Page 3 BC: Director, Emergency Services Director Regulatory Affairs Manager, Licensing Manager EP Records Management
Attachment LR-N1 2-0366 Page 1 50.54Q I.D. Number: 2012 - 35 50.54Q
Title:
Evacuation Time Estimate Study, KLD TR-499 Revision 0 Description of the change made to the Emergency Plan/Procedures:
The Evacuation Time Estimate Report, prepared by KLD Engineering,P.C. for PSEG, describes the analyses and the results obtained from development of the Evacuation Time Estimates (ETE) for the Salem & Hope Creek Nuclear Generation Stations (SHCNGS), located in Salem, NJ. ETEs are part of the required planning basis and provide PSEG Nuclear along with State and local governments, site-specific information needed for Protective Action decision making. The report was prepared by KLD lAW NUREG/CR-7002 using 2010 Census data.
The Executive Summary from the ETE Report will be maintained as Attachment 6 to the PSEG Emergency Plan replacing the existing 2004 ETE currently in Attachment 6. The complete Evacuation Time Estimate Report, including all associated attachments is maintained as a separate Reference Document internal to the PSEG Document Management System.
EPZ Population Data from the ETE will be used to revise the ERPA Population Table in PSEG Nuclear Emergency Plan Section 1, Introduction. After the ETE is submitted to the NRC for the required 180 day review period, ETE information concerning estimated evacuation times based on time-of-day and weather conditions will be used to update procedure NC.EP-EP.ZZ-0404 - Protective Action Recommendation (PAR) Update. In addition, the States of New Jersey and Delaware will be provided the final ETE Report for their use in updating their emergency plans and/or Protective Action Strategies as they deem necessary.
Description of why the change is editorial (if not editorial, N/A this block):
N/A Description of the licensing basis affected by the change to the Emergency Plan/Procedure (if not affected, omit this element):
" PSEG Emergency Plan o Emergency Plan Section 1 contains information derived from the ETE (Page 1.10 Table 1-1 ERPA Populations) - will be revised per new ETE o Emergency Plan Attachment 6 is the 2004 ETE study performed by KLD Associates - will be revised per new ETE to include only the ETE Report Executive Summary with reference to the complete ETE Report.
" NC.EP-EP.ZZ-0404(Q) Protective Action Recommendations (PARS) Upgrades:
o Attachment 4, Evacuation Time Estimates - General Public Evacuation Times contains the evacuation time estimates for NJ and DE ERPAs for day and evening under normal and adverse conditions - will be revised per new ETE after 180 day NRC review period.
o Attachment 5, PermanentPopulationDistributionBy Compass Direction contains the compass rose with population distributions - will be revised per new ETE after 180 day NRC review period.
Attachment LR-N 12-0366 Page 2 50.54Q I.D. Number: 2012 - 35 50.54Q
Title:
Evacuation Time Estimate Study, KLD TR-499 Revision 0 A description of how the change to the Emergency Plan/Procedures still complies with regulation:
The Evacuation Time Estimate study was performed lAW NUREG/CR-7002 using 2010 Census data and is in accordance with emergency plan planning standards per 10 CFR 50.47 and does not change or contradict program functions or elements.
A description of why the proposed change was not a reduction in the effectiveness of the Emergency Plan/Procedure:
The ETE and resulting changes to the E-Plan and associated procedures are consistent with regulatory guidance provided for development of such studies and the changes have no negative impact on protective actions function as described in the PSEG Emergency Plan and supporting procedures.
LR-N 12-0366 Salem - Hope Creek Evacuation Time Estimate Study (Please see the following document - 498 pages)
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 a PSEG Nuclear LLC 3 A 2012~
Nov 3 0 201210 L CFR 95.9 LR-N12-0338 Document Control Desk Director, Division of Security Operations Office of Nuclear Security and Incident Response U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Salem Nuclear Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354
Subject:
Facility Security Officer
Dear Sir or Madam,
As required per 10 CFR 95.19(b)(1), this letter is to inform you of the change in appointment of the Facility Security Officer at PSEG Nuclear. Please be advised that the Facility Security Officer is Security Director Peter Tocci who received his "L"classified clearance on July 11, 2012.
This submittal does not contain any new regulatory commitments.
If you have any questions please do not hesitate to contact Mrs. Michelle Patti at 856-339-3691.
Sincerely, Paul R. Duke, Jr.
Licensing Manger
LR-N12-0338 Page 2 N0o 3 0 Zell cc:
DCRMS HC Commitment Tracking Coordinator Salem Commitment Tracking Coordinator Corporate Commitment Tracking Coordinator