ML19210A420
| ML19210A420 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/20/1976 |
| From: | Arnold R METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| GQL-1189, NUDOCS 7910290640 | |
| Download: ML19210A420 (2) | |
Text
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G3 EON oOCKE T NUMDtn (2 M' 50-289 NRC DISTRIBUTION ron PART 50 DOCKET MATERIAL TO:
FROM:
oATc c7 coCUMENT Metropolitan Edison Company 8/20/76 Mr. Robert 'i. Reid Reading, Pa.
oATt ntCrivEo R. C. Arnold 9/23/76
)(.ETTEn ONoronizEo encP INPUT FORM NUMOEn cF COPIES RECEIVED Konio'NAL
)(UNCLAssir Eo One signed CCoPv oESCRIPTicN ENCLOSURE Ltr. re our 6/9/76 ltr. and their 7/9/76 ltr.....concerning Reactor Vessel Supports Analysis.
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PLANT NAME:
Three Mile Island 41 DISTRILUTION FOR REACTOR VESSEL SUPPORT INFO FOR OPERATING REACTORS PER MR. TRAMMELL 7-12-74 SAFETY FOR ACTION /INFORMATION 8/25/76 RJL IGNED AD; Collor
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Directcr cf :Iuclear Reacter Regulation Atta: Mr. Rccert W. Reid, Chief Cperating Reacters 3 ranch #h x
U.S. :iuclear Regulatory Cc==ission w
Washington, D.C.
20555 Lear Sir:
Three Mile Island Nuclear Statica Unit 1 (T:C-1)
Decket :Io. 50-269 Cperating License :Io. DPR-50 Reacter Vessel Suppcrts Analysis Ycur letter of June 9,1976 requested additional infor=ation evaluating the adequacy of the reactor pressure vessel supports under leads not previcusly considered. As indicated in our July 9th resp nse, ve ha.e
=et with other Sabcock & Wilcox :ISSS users in an effort to evaluate the ccncern and discuss pcssible resolutions.
Two paths are being pursued in parallel to 'letermine our required acticn.
The first approach is to undertake a review of verk spenscred by a group of utilities evaluating the absolute and relative pr bability of a postu-lated pipe break between the reacter vessel neccle and the reacter cavity vall.
Science Applicatiens, Inc. (SAI) is submitting their results ir, the form of a tcpical report in late August, 1976.
'"he :IRC staff was intr duced to the study in a presentation involving SAI en July 13, 1976.
The seccnd apprcach is a review of the additional informatien requested in your June 9th letter which requires a ecstly, multi-year analysis fer each plant. Discussions with 3abecek & Wilecx and cther consulting firms, review cf ACES hearings transcripts, and your letter itself indicates uncertainties in the present state of the art analysis to evaluate this particular case.
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Mr. Rctert '4.
Reid, Ch.ef August 20, 197c
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- wf As a result of our preliminary review, it is our belief that the SA:
study substantiates the extremely icw probability of a pipe rupture in the reactor vessel cavity area at IMI-1; therefore, the pcstulated treak represents no significant hazard to the health and safety of the public. In this light, no further analysis is censidered required.
Following your review of the SAI study, we would like to meet with you to discuss this issue and address any further cencerns yor =ight have.
Sincerely,
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R.C.Abncld Vice President ECA:1CL:=ft
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