LR-N07-0145, Supplement to License Amendment Request to Relocate Component Lists for Primary Containment Isolation Valves from Technical Specifications

Supplement to License Amendment Request to Relocate Component Lists for Primary Containment Isolation Valves from Technical Specifications

ML071840164
Person / Time
Site:	Hope Creek
Issue date:	06/21/2007
From:	Barnes G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H06-02, LR-N07-0145
Download: ML071840164 (4)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 0 PSEG NuclearLLC 10 CFR 50.90 LR-N07-0145 LCR H06-02 JUN 2 1 2007 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Facility Operating License No. NPF-57 NRC Docket No. 50-354

Subject:

Supplement to License Amendment Request to Relocate Component Lists for Primary Containment Isolation Valves from Technical Specifications

Reference:

1) Letter from George P. Barnes (PSEG Nuclear LLC) to USNRC, November 15, 2006 In Reference 1, PSEG Nuclear LLC (PSEG) requested an amendment to Facility Operating License NPF-57 and the Technical Specifications (TS) for the Hope Creek Generating Station (HCGS) to remove valve component lists and references to the lists from the Technical Specifications (TS).

PSEG determined that Reference 1 inadvertently omitted the deletion of references to TS Table 3.6.3-1 from TS Actions 3.6.1.2.d and 3.6.1.2.e and from TS restoration Actions 3.6.1.2.d and 3.6.1.2.e. These additional TS changes are consistent with the proposed changes to Limiting Conditions for Operation (LCO) 3.6.1.2.d and 3.6.1.2.e in Reference 1. Attachment 1 to this letter provides a revised marked up TS page that incorporates the additional required changes.

PSEG has determined that the information contained in this letter and attachments does not alter the conclusions reached in the 10CFR50.92 no significant hazards analysis previously submitted.

There are no regulatory commitments contained within this letter.

95-2168 REV. 7/99

LR-N07-0145 LCR H06-02 JUN 2 1 2007 Page 2 Should you have any questions regarding this submittal, please contact Mr. Paul Duke at 856-339-1466.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on __/Z /_/0 _7 (date)

Sincerely, George P. Barnes Site Vice President Hope Creek Generating Station cc: S. Collins, Regional Administrator - NRC Region I R. Ennis, Project Manager - USNRC NRC Senior Resident Inspector - Hope Creek K. Tosch, Manager IV, NJBNE LR-N07-0145 LCR H06-02 Hope Creek Generating Station Facility Operating License NPF-57 Docket No. 50-354 Supplement to License Amendment Request to Relocate Component Lists for Primary Containment Isolation Valves from Technical Specifications Revised Markup of Technical Specification Page TS Page 3/4 6-3

CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION 'ContinuedP ACTION (Continued)

d. The measured combined leakage rate for all containment isolation valves hboundary for the long-term seal of the feedwater lines

_ .th bxceedingd !0 gpm, or

e. The measured combined leakage rate for all nther penetrations and containment isolation valves in hydrostatically tested lines '_ e wwhich penetrate the primary containment exceeding l& gpm, restore: *~ ~ r'ITý

[-- I*G Gi k.-ET"VER.

1*S IT LPZ.- tik0 ( - 0V[

,O 'n L

a. The overall integrated leakage rate(s) (Type A test) to be in accordance with the Primary Containment Leakage Rate Testing Program, and bI

b. T, " ed leakage rate for all penetrations and all valves listed in

c. The leakage rate to less than or equal to 150 scfh per main steam line and less than or equal to 250 scfh combined through all four main steam lines, and

d. The combined leakage rate for all containment isolation valves which form the boundary for the long-term seal of the feedwater i nesen tope B less than or equal to gpm, and t0

e. Lhe combined leakage rae Pro for all other and pen etrate theprimary ctntainment to less than or equal to 10 gpm, prior to increasing reaktor coolant system temperature above 200°F.

SUIRVEILLANCE REQUIREMENTS 4.6.1.2.a The primary containment leakage rates shall be demonstrated in accordance with the Primary Containment Leakage Rate Testing Program for the following:

i. Type A test.

2. Type B and C tests (including air locks).

b. DELETED. accodaneCntaimenwth te PimayLekageRat 'rstig Prgra fo th

c. DELETED.

p Exemption to Appendix "J" of 10 CFR 50.

HOPE CREE " r 3/4 6-3 Amllendment No. 134