LR-N07-0261, 10CFR50.46 Report, Informing of Changes in the Application of the Emergency Core Cooling System Evaluation Modes

10CFR50.46 Report, Informing of Changes in the Application of the Emergency Core Cooling System Evaluation Modes

ML072770303
Person / Time
Site:	Hope Creek
Issue date:	09/28/2007
From:	Gaffney M Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N07-0261
Download: ML072770303 (6)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG Nuclear LLC 10CFR50.46 LR-N07-0261 September 28, 2007 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354

Subject:

10CFR50.46 REPORT Pursuant to the requirements of 10CFR50.46, PSEG Nuclear LLC (PSEG) hereby reports changes in the application of the Emergency Core Cooling System (ECCS) evaluation models for the Hope Creek Generating Station. 10 CFR 50.46(a)(3)(ii) requires licensees to report at least annually each change to or error discovered in evaluation models used for calculating ECCS performance and the estimated effect on the limiting ECCS analysis. For significant changes or errors, licensees are required to submit a 30 day report and include a proposed schedule for providing a reanalysis or taking other action necessary to show compliance with 10 CFR 50.46 requirements.

This report satisfies the annual reporting requirement.

For the current operating cycle, the Hope Creek core contains a mixture of Westinghouse SVEA-96+ and GE14 fuel.

If you have any questions regarding this submittal, please contact Mr. Francis D.

Possessky at (856) 339-1160.

Sincerely, Michael aftn a As ur nce Manager- Hope Creek Attachments (2) . 2 P -(

95-2168 REV. 7/99

LR-N07-0261 Page 2 C Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission Mr. R. Ennis, Project Manager - Hope Creek Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. P. Mulligan, Manager IV (Acting)

Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625

Attachment 1 (Pages - 1) 10 CFR 50.46 Report

LR-N07-0261 Attachment 1 Hope Creek Generating Station 10 CFR 50.46 Report Peak Cladding Temperature Rack-up Sheets PLANT NAME: Hope Creek Generating Station ECCS EVALUATION MODEL: SAFER/GESTR-LOCA REPORT REVISION DATE: 9/6/2007 CURRENT OPERATING CYCLE: 14 ANALYSIS OF RECORD Evaluation Model: The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume Ill, SAFER/GESTR Application Methodology, NEDE-23785-1-PA, General Electric Company, Revision 1, October 1984.

Calculations: "SAFE R/G ESTR-LOCA Loss-of-Coolant Accident Analysis for Hope Creek generating Station," NEDC-33153P, Revision 1, GE Nuclear Energy, September 2004.

Fuel: SVEA-96+ and GE14 Limiting Fuel Type: SVEA-96+

Limiting Single Failure: Battery Limiting Break Size and Location: 1.0 Double-Ended Guillotine in a Recirculation Suction Pipe Fuel Type: SVEA-96+ GE14 Reference PCT 1540 OF 1370 OF MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report Dated June 1, 2004 (See note 1) APCT = 0°F APCT = 0°F 10 CFR 50.46 Report Dated September 29, 2006 (See APCT = 0°F APCT = 0°F note 2)

Net PCT 1540 OF 1370 OF B. CURRENT LOCA MODEL ASSESSMENTS None APCT = 0°F APCT = 0°F Total PCT change from current assessments ZAPCT = 0°F ZAPCT = 0°F Cumulative PCT change from current assessments Z APCT I= 00 F Y Y IAPCT I= 0°F Net PCT 1540 OF 1370 OF

Attachment 2 (Pages - 1) 10 CFR 50.46 Report Assessment Notes

LR-N07-0261 Hope Creek Generating Station 10 CFR 50.46 Report Assessment Notes

1. Prior LOCA Model Assessment A 10 CFR 50.46 report for Hope Creek was submitted in June 1, 2004.

Subsequent to this report and with the startup of cycle 13 in November 2004, Hope Creek discharged all GE9B fuel and implemented GE14 fuel. The Referenced LOCA analysis was implemented as analysis of record for GEI4 fuel and SVEA-96+ fuel.

[

Reference:

"SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis for Hope Creek generating Station," NEDC-33153P, Revision 1, GE Nuclear Energy, September 2004.]

2. Prior LOCA Model Assessment In the referenced letter to the NRC, the impact of a GE postulated new heat source applicable to the LOCA event was reported. This heat source is due to recombination of hydrogen and excess oxygen drawn into the vessel from containment during core heatup. The PCT impact for all fuel types was 00 F. The referenced letter also reported the impact of the top peak axial power shape on the small break LOCA. The impact of the top peak axial power shape on the licensing basis PCT was zero degree for both GE 14 fuel and SVEA-96+ fuel for Hope Creek.

[

Reference:

Letter (LR-N06-0401) from Michael Jesse (PSEG) to U.S. NRC, "10 CFR 50.46 Report."]