ML18219E000

Letter Adequacy of Reactor Pressure Vessel Supports

ML18219E000
Person / Time
Site:	Cook
Issue date:	08/27/1976
From:	Patterson G Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co)
To:	Rusche B Office of Nuclear Reactor Regulation
References
Download: ML18219E000 (8)

Text

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INDIANAl MICHIGAN POWER COMPANY P. O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 Donald C. Cook Nuclear Plant Unit Nos.

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Dear Mr. Rusche:

Mr. Karl Kniel'~s letter of J'uly 30, 1976, received August 4, 1976, requested that we provide the NRC with a schedule for submitting an evaluation of the adequacy of the reactor pressure vessel supports.

This evaluation was to include answers to the Request For Additional Informal.on which was enclosed with Mr. Kniel's letter.

That letter was an outcome of Mr. Kniel'~s earlier letter of November 26, 197$, which informed us of a generic evaluation of reactor pressure-vessel supports.

Since the end of 1975, Indiana Bc Michigan Power Company representatives as well as those from other utilities with operating Westinghouse PWR's have met ]ointly to investigate the several options available to,us to address the adequacy of reactor pressure vessel supports.

The owners group has considered the performance of analyses similar to those suggested in your Request For Additional Information for each individual plant or for several typical designs which would envelope most of the plants in'the owners group.

Discussions with Westinghouse Electric Corporation, as well as with other organizations potentially capable of performing such analyses, have indicated that a significant amount of effort would be required, and in fact such analyses would probably take from one to three years to complete.

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Mr. B. Rusche August 27~ 1976 Plants that are not in operation or not well along in construction can include design features which reduce the amount of any analysis needed to demonstrate acceptable consequences from postulated events.

Operating plants, on the other hand, cannot change structural configurations without severe economic penalties, if at all.

Thus, events which were previously analyzed in accordance with approved techniques and found acceptable are now, because of recently developed methods of

analyses, facing the potential burden of costly state"of-the"art techniques to demonstrate that the consequences remain within acceptable limits.

We have, therefore, concluded that the most prudent course of action in response "to your questions is to propose implementation of an augmented in"service inspection program to preclude a reactor coolant pipe break near the reactor vessel nozzles.

This'program will have a positive impact on plant safety and eliminate the need to perform extensive and lengthy analyses which have minimal impact on the real margin of safety existing in these)plants.

Representatives of the owners group and the Westinghouse Electric Corporation met with members of the Nuclear Regulatory Commission Staff on May 25, 1976 to discuss our efforts to that date including justification of our intention to submit an augmented program and the technical merits of such a program.

Preparation of a report detailing the discussion at that meeting has been completed and is under review.

This document shall be formally transmitted to you in September 1976 as technical justification for selection of the augmented in"service inspection program.

I, Very truly yours, GVP:mam G. V. Pat erson Vice President Sworn to and subscribed to before me this

<7'~ day of August 1976 in New York County, New York Notary Public cc:

see next page KA'I'IIIEIrN gAI(RY'OTARY VU8t!C, State of No Y

No. 41-4GOGi92 Qualified in Queens County CttftIficoto filed in New York Coun t.on>nu'saon tttp>res Natch 30, NP'7

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Nr. B. Rusche Pl 3 August 27> 1976 cc:

G. Charnoff D. D. Comey R. C. Callen P.

W. Steketee R. Walsh R.

S. Hunter R.

W. Jurgensen

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