ML20206Q621

Informs That on 990119 Licensee Provided NRC with Several Revised TS Bases Pages for Plant.Ts Bases Pages B 3/4 6-1 & B 3/4 6-2 Were Revised

ML20206Q621
Person / Time
Site:	Hope Creek
Issue date:	05/14/1999
From:	Richard Ennis NRC (Affiliation Not Assigned)
To:	Keiser H Public Service Enterprise Group
References
TAC-MA5424, NUDOCS 9905190180
Download: ML20206Q621 (5)

Text

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Mr. Harold W. Keiser g,y 14, j999  !

Chief Nuclear Officer & President - 1 Nuclear Business Unit Public Service Electric & Gas Company >

Post Office Box 236 l Hancocks Bridge, NJ 08038 '

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SUBJECT:

HOPE CREEK GENERATING STATION, CHANGES TO TECHNICAL SPECIFICATIONS BASES PAGES (TAC NO. MA5424)

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Dear Mr. Keiser:

, In a letter dated January 19,1999, the Public Service Electric and Gas Company provided to j the U.S. Nuclear Regulatory Commission (NRC) several revised Technical Specification (TS)  ;

Bases pages for the Hope Creek Generating Station (HCGS). The following TS Bases pagas were revised: B 3/4 6-1 and B 3/4 6-2.

The NRC ctaff has reviewed the TS Bases changes and has no objection to the changes. The j enclosed TS pages are being distributed for inclusion in the HCGS TSs. j Sincerely,

, ORIGINAL SIGNED BY:

Richard B. Ennis, Project Manager, Section 2 Project Directorate i I Division of Licensing Project Management  !

Office of Nuclear Reactor Regulation  !

, Docket No. 50 354

Enclosures:

Revised TS Pages B 3/4 6-1  !

l/

cc w/encis: See next ag DISTRIBUTION: kO}

Docket File > JClifford OGC PUBLIC REnnis ACRS i PDI-2 Reading File MO'Brien GMeyer, RGN-l l WBeckner l l

OFFICE PDI-2/PM Pfb PDI-2/SC NAME REnnisN M61 den' JClbrd DATE fO / /99 h/ 9 6/ /99 _

OFFICIAL RECORD COPY ,/ 1 DOCUMENT NAME: LTRBASE1.WPD I 9905390180 990514 34

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PDR ADOCK O y l

m3 Rion p & UNITED STATES g

j NUCLEAR REGULATORY COMMISSION  !

o WASHINGTON. D.C. 20666-0001  !

4 ,o 4***** May 14, 1999 Mr. Harold W. Keiser Chief NuclearOfficer& President-Nuclear Business Unit Public Service Electric & Gas Company Post Office Box 236 Hancocks Bridge, NJ 08038

SUBJECT:

HOPE CREEK GENERATING STATION, CHANGES TO TECHNICAL SPECIFICATIONS BASES PAGES (TAC NO. MA5424)

Dear Mr. Keiser:

In a letter dated January 19,1999, the Public Service Electric and Gas Company provided to the U.S. Nuclear Regulatory Commission (NRC) several revised Technical Specification (TS)

Bases pages for the Hope Creek Generating Station (HCGS). The following TS Bases pages were revised: B 3/4 61 and B 3/4 6-2.

The NRC staff has reviewed the TS Bases changes and has no objection to the changes. The enclosed TS pages are being distributed for inclusion in the HCGS TSs.

Sincerely, Richard B. Ennis, Project Manager, Section 2 Project Directorate i Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-354

Enclosures:

Revised TS Pages B 3/4 6-1 and B 3/4 6-2 cc w/encls: See next page s

I Hope Creek Generating Station cc:

Jeffde J. Keenan, Esquire Manager- Joint Generation Nuclear Business Unit - N21 Atlantic Energy P.O. Box 236 6801 Black Horse Pike Hancocks Bridge, NJ 08038 Egg Harbor Twp., NJ 08234-4130 Hope Creek Resident inspector Richard Hartung U.S. Nuclear Regulatory Commission Electric Service Evaluation Drawer 0509 Board of Regulatory Commissioners Hancocks Bridge, NJ 08038 2 Gateway Center, Tenth Floor Newark, NJ 07102 Mr. Louis Storz Sr. Vice President - Nuclear Operations Lower Alloways Creek Township Nuclear Department c/o Mary O. Henderson, Clerk ,

P.O. Box 236 Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Hancocks Bridge, NJ 08038 '

General Manager - Hope Creek Operations Mr. Elbert Simpson Hope Creek Generating Station Senior Vice President-P.O. Box 236 Nuclear Engineering Hancocks Bridge, NJ 08038 Nuclear Department P.O. Box 236 Director- Licensing Regulation & Fuels Hancocks Bridgo, NJ 08038 Nuclear Business Unit - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Reg;onal Administrator, Region i U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Dr. Jill Lipotl, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415

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3/4.6- CONTAINMENT SYSTEMS j 1

BASES )

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l 3/4.6.1 PRIMARY CONTAINMENT

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l 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmospnere 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 l accident conditions. l 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 i Surveillance Requirement 4.6.1.1.b since access to these areas is typically l restricted in accordance with the requirements of Technical Specification 6.12 l and/or plant procedures. In addition, field verification for these components I 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 prihary 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, P.. As an added conservatism, the measured overall integrated leakage rate (Type A test) is further limited to less than or equal to 0.75 L. 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.

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 CONTAINMEm' 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 -MSIV SEALING SYSTEM 1

Calculated doses resulting from the maximum leakage allowance for the main steamline-isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR 100 guidelines, provided the main steam line system from the iso 1'ation valves up to and including the turbine condenser

+.

1 HOPE CREEK B 3/4 6-1 Revised by letter dated May 14, 1999 I-

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CONTAINMENT SYSTEMS

! PASES

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3/4.6.1.4 MSIV SEALING SYSTEM (Continued) remains intact. Operating experience has indicated that degradation has occasionally occurred in the leak tightness of the MSIV's such that the specified leakage requirements have not always been maintained continuously.

The sealing system will reduce the untreated leakage from the MSIVs when l isolation of the primary system and containment is required. l 3/4.6.1.5 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 48,1 psig in the event of a LOCA. A visual inspection in accordance with the Primary Containment Leakage Rate Testing Program is sufficient.

3/4.6.1.6 DRYWELL AND SUPPRESSION CHAMBER 7NTERNAL PRESSURE The limitations on drywell and suppression chamber internal pressure ensure that the containment peak pressure of 48.1 psig does not exceed the design pressure of 62 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 3 psid. The limit of -0.5 to +1.5 psig for initial positive containment pressure will limit the total pressure to 48.1 psig which is less than the design pressure and is consistent with try safety analysis.

3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature onsures that the containment peak air' temperature does not exceed the design temperature of 340*F during LOCA conditions and is consistent with the safety analysis. The 135*F average temperature is conducive to normal and long term operation.

3/4.6,1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> /365 days limit for the operation of the purge valves and the 6' nitrogen supply valve during plant Operational Conditions 1, 2 and 3 is intended to reduce the probability of a LOCA occurrence during the above l

operational conditions when the applicable combination of the above valves are l open.

Blow-out panels are installed in the CPCS ductwork to provide additional assurance that the FRVs will be capable of performing its safety function subsequent to a LOCA.

HOPE CRECK B 3/4 6-2 Revised W letter dated May 14, 1999