PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236

MAY 1 6 2002 0 PSEG

LR-N02-0192

United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

CORRECTION OF TECHNICAL SPECIFICATION PAGE ISSUED WITH AMENDMENT 245 SALEM GENERATING STATION UNIT NO. I DOCKET NO. 50-272

Ladies and Gentlemen:

On September 19, 2001, the NRC issued Amendment No. 245 for Salem Unit 1. This amendment revised Technical Specification Bases page 3/4 9-4 for the Fuel Handling Area Ventilation System. After subsequent review of the bases page, PSEG Nuclear has determined that the page 3/4 9-4 issued with Amendment No. 245 did not include changes to the same page that were previously approved under Amendment No. 231 issued on June 15, 2000. The changes proposed under Amendment No. 245 did not delete this information and therefore the page issued for Amendment No. 245 should have included these changes.

The attached bases page B 3/4 9-4 contains the information from Amendment No. 231 that was inadvertently omitted during issuance of Amendment No. 245.

If you have any questions regarding this submittal, please contact Brian Thomas at 856-339-2022.

XSincerell

G. Salamon Nuclear Safety and Licensing Manager

95-21-8 ýE" - 99

Document Control Desk 2 MAY 1 6 2002 LR-N020192

C Mr. Hubert J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406

U. S. Nuclear Regulatory Commission ATTN: Mr. Robert Fretz, Licensing Project Manager - Salem Mail Stop 0882 Washington, DC 20555

USNRC Senior Resident Inspector- Salem (X24)

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

REFUELING OPERATIONS BASES

A listing of the active (air/motor operated) valves in the affected flow path to be locked open or disabled.

Note that four filled reactor coolant loops, with at least two steam generators with at least their secondary side water level greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop. This ensures that a single failure does not cause a loss of decay heat removal.

With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 CONTAINMENT PURGE AND PRESSURE-VACUUM RELIEF ISOLATION SYSTEM

The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

3/4.9.10 and 3/4/9/11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL

The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM

The limitations on the fuel handling area ventilation system ensure that all radioactive material released from a dropped irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system is consistent with the assumptions of the accident analyses. Laboratory testing of the carbon adsorber is performed in accordance with ASTM D3803-1989 with an acceptance criteria that is determined by applying a minimum safety factor of 2 to the charcoal filter removal efficiency credited in the design basis dose analysis as specified in Generic Letter 99-02.

The operability of the Fuel Handling Area Ventilation System during movement of irradiated fuel ensures all building exhaust flow is processed through the HEPA/charcoal filter train whenever a Fuel Handling Accident is possible. This will minimize offsite doses following the postulated Fuel Handling Accident.

Amendment No. 245SALEM - UNIT 1 B 3/4 9-4

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