ML19317F526
| ML19317F526 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 10/31/1977 |
| From: | Roe L TOLEDO EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| 398, NUDOCS 8001150840 | |
| Download: ML19317F526 (2) | |
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Totsoo EDISON Docket No. 50-346 LOWELL E. ROE v.c. pr a.a.
October 31, 1977 14191 259 5242 Serial No. 398
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Director of Nuclear Reactor Regulation
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Attn:
Mr. John F. Stolz, Chief 0 s, Light Water Reactors Branch No. 1 Division of Project Management j
%j U.S. Nuclear Regulatory Commission
\\h Washington, D.C.
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Dear Mr. Stolz:
On September 27, 1977, Science Applications Incorporated submitted to Dr. Charles Trammel, Project Manager, Division of Operating Reactors, Topical Report No. SAI-050-77-PA, "An Analysis of the Probability of Pipe Rupture at Various Locations in the Primary Coolant Loop of a Babcock & Wilcox 177 Fuel Assembly Pressurized Water Reactor--Including the Effects of a Periodic Inspection." The facilities for which this Topical Report and its conclusions are applicable includes the Davis-Besse Nuclear Power Station Unit No. 1.
We believe that this Topical Report alleviates the concerns which prompted the request for information transmitted in Mr. W. R. Butler's letter dated May 21, 1976 regarding the design of reactor pressure vessel support systems at Davis-Besse Nuclear Power Station Unit No. 1.
Our letter to you dated August 25, 1976 proposed a tentative schedule for performing the extensive safety analyses required to respond to Mr. Butler's May 21, 1976 letter,but def erred a con:mitment to performing the analyses pending our evaluation of alternative approaches.
Based on the results of the SAI Topical Report we believe that the probability of an event which would result in loads outlined in Mr. Butler's May 21, 1976 letter is sufficiently small to not warrant inclusion in the design bases of the Davis-Besse Nuclear Power Station Unit No. I supports.
Therefore,we are confident that the Davis-Besse Nuclear Power Station Unit No. I can operate safely with the existing reactor vessel supports, and no further safety analyses of the supports are required.
Yours very truly,
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