ML20039F656

Forwards Info in Response to Open NUREG-0737 Items II.B.3, post-accident sampling,II.E.4.2,containment Isolation Dependability & Closed Item 36,details of Debris Screen for 6-inch Vent Drywell Line

ML20039F656
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	01/11/1982
From:	James Smith LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.E.4.2, TASK-TM SNRC-657, NUDOCS 8201130193
Download: ML20039F656 (4)

Text

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Mwwww , SHOREHAM NUCLEAR POWER STATION

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• WADING RIVER N.Y.11792 January 11, 1982 SNRC-657

.x, Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation

• U.S. Nuclear Regulatory Commission , 8IMEI'dG Washington, D.C. 20555 .

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Docket No. 50-322 \

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Dear Mr. Denton:

5/ j g Enclosed herewith are sixty (60) copies of information which is either in response to specific NRC open items or confirma-tory relative to past open items which are now closed. The scope of this specific submittal is tabulated on Attachment A.

We trust that the enclosed information is satisfactory. Should you have any questions or require additional information, please do not hesitate to contact this office.

Very truly yours, J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station RWG:mp Enclosure cc: J. Higgins 8201130193'020111 PDR ADDCK 05000322 A PDR l

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~ Attachment A.

I. Items submitted in response to specific NRC open items:

1. Item II.B;3 - Post Accident. Sampling -

Procedures for the Determination'of the Extent of Core Damage

2. . Item II.E.4.2 - Containment Isolation Dependability -

High-radiation isolation signal i

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-II. One item submitted as confirmatory for Open Item 36 (open

-item.36 is now closed)

1. Details of debris screen for 6" vent drywell.line f

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____ ____ ___-m___

m II.B.3 - Procedures for the Determination of the Extent of Core Damage The attached document, " Procedures for the Determination of the Extent of Core Damage under Accident Conditions", RPE 81CCL01, has been prepared for possible use by members of the BWR Owners Group (BWROG). LILCO feels _that utilization of the contents of this report is adequate to address the concept of determining the approximate degree of core damage based on measured fission product concentrations. The information obtained in utilizing these procedures is not considered essential as a detern.inant for potential operator action, although it could be used as guidance. It should be noted that the BWROG Regulatory Guide 1.97 Committee is presently addressing this concern. LILCO plans-to adopt their resolution of this matter as appropriate after it is finalized but not necessarily prior to fuel load.

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l Item II.E.4.2 - Containment Isolation Dependability LILCO commits to providing a high-radiation ~ isolation signal to the 4" and 6" containment vent and purge isolation valves which

-may be open during operational conditions 1, 2 or 3. These valves are designated as follows:

1T24*AOV001A,B 1T24*AOV004A,B 1T46*AOV078A,B 1T46*AOV079A,B This change will be implemented on.a schedule consistent with equipment availability, but is not expected to be completed before commercial operation.

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