LR-N03-0107, TS Bases Change

TS Bases Change

ML030920298
Person / Time
Site:	Hope Creek
Issue date:	03/21/2003
From:	Salamon G Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N03-0107
Download: ML030920298 (5)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bndge, New Jersey 08038-0236 MAR 2 1 2003 @ PSEG NuclearLLC LR-N03-0107 Bases Change H01-002B United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 TECHNICAL SPECIFICATION BASES CHANGE HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NO. NPF-57 DOCKET NO. 50-354 PSEG Nuclear LLC is providing revised Technical Specification (TS) Bases page for Specification 3/4.6.1.2. The revised page was reviewed in accordance with the requirements of 10 CFR 50.59.

License Amendment 134 revised TS 3/4.6.1.2.c to read, "Primary containment leakage rates shall be limited 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 when tested at 5 psig (leakage rate corrected to 1 Pa, 48.1 psig)". Action c was changed to read, "the measured leakage rate exceeding 150 scfh per main steam line or exceeding 250 scfh combined through all four main steam lines" restore "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". TS 3 /4.6.1.2 Bases were revised to read, "If the leakage rate on a main steam line exceeds the requirements of Technical Specification 3.6.1.2.c (150 scfh), the leakage rate for that line will be restored to less than or equal to 25 scfh (when tested at 5 psig and corrected to Pa) prior to plant restart". Pa was added to the Bases in error and should have been 25 psig. This is to ensure that the restore to leakage rate is well below the as found TS limit. contains the revised page for the Hope Creek Technical Specification Bases. Please incorporate this change into the Technical Specification Bases.

A1 95-2168 REV. 7/99

Document Control Desk 2 MAR 2 12003 LR-N03-01 07 BASES CHANGE H01-002B Should you have any questions regarding this submittal, please contact Michael Mosier at 856-339-5434.

Sincerely, 7~t- OA9,6 . ro4mohi)

G. Salamon Manager - Nuclear Safety & Licensing Attachment

Document Control Desk 3 LR-N03-0107 MAR 2 1 2003 BASES CHANGE H01-002B C Mr. H. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. George Wunder Licensing Project Manager - Hope Creek U. S. Nuclear Regulatory Commission Mail Stop 08B1 Washington, DC 20555 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625

Document Control Desk LR-N03-01 07 Attachment I BASES CHANGE H01-002B HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REVISIONS TO THE TECHNICAL SPECIFICATIONS BASES Page B 3/4 6-1

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

In high radiation areas and in areas posted as neutron exposure areas and controlled in a manner similar to high radiation areas, use of administrative means to verify position of valves and blind flanges is acceptable for Surveillance Requirement 4.6.1.1.b since access to these areas is typically restricted in accordance with the requirements of Technical Specification 6.12 and/or plant procedures. In addition, field verification for these components is performed before restarting from each refueling outage. Therefore, the probability of misalignment of these components, once they have been verified to be in the proper position, is low.

3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the design basis LOCA maximum peak containment accident pressure of 48.1 psig, Pa. As an added conservatism, the measured overall integrated leakage rate (Type A test) is further limited to less than or equal to 0.75 La during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves. If the leakage rate on a main steam line exceeds the requirements of Technical Specification 3.6.1.2.c (150 scfh), the leakage rate for that line will be restored to less than or equal to 25 scfh (when tested at 5 psig and corrected to 25 psig) prior to plant restart.

The surveillance testing for measuring leakage rates is consistent with the Primary containment Leakage Rate Testing Program.

3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY and the Primary Containment Leakage Rate Testing Program. Only one closed door in each air lock is required to maintain the integrity of the containment.

3/4.6.1.4 (Deleted)

HOPE CREEK B 3/4 6-1