ML19210A591
| ML19210A591 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/06/1977 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| GQL-1366, SAI-050-77-PA, SAI-50-77-PA, NUDOCS 7910300631 | |
| Download: ML19210A591 (4) | |
Text
u.s. NUCLE AR REGULATORY CC ON OCC KET NUM BE R NRC PoRu 195 50 289 a.vsi
, NRC DISTRIBUTION rom PART 50 DOCKET MATERIAL FROM:
CATE OF QCCUMENT TO:
Hetropolitan Edison CCepany 1C/f/77 Mc. 2. '4. Reid Reading, Pa.
oATE RECEivEo J. G. Herbein 10/13/77 C LE TTE R ONOTORIZEQ PROP INPUTPQ.W NUMBER QP CCP.Es RECEIVED C ORIGIN AL CusCLAssipito One Signed CCcPv OESCRIPTION ENCLOSU RE Consists of info re Reactor Vcssel Supports Adequacy.....
PLANT NAME: Three Mile Island Unit No. 1 RJL IC/13/77 (3-P)
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un a mm METHOPOLil AN EDISON COMPANY POST OFFICE BOX 542 RE ACING. PENNSYLV ANI A 19603 TELEPHONE 215 - 929 2001 October 6, 1977 GQL 1366 Director of Nuclear Reactor Regulation Attn:
Mr. R. W. Reid, Chief Operating Reactors, Btanch No. 4 United States Nuclear Regulatory Commission Washington, D. C.
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Dear Sir:
. U, Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License DPR-50 Docket No. 50-289 Reactor Vessel Supports Adequacy
References:
1.
Letter from R. W. Reid to R. C. Arnold, dated October 15. 1975 2.
Letter from R. C. Arnold to R. W, Reid, GQL 1739, dated November 21, 1975 3.
Letter from R. W. Reid to R. C. Arnold, dated June 9, 1976 4.
Letter from R. C. Arnold to R. W. Reid, GQL 1189, dated August 20, 1976 5.
An Analysis of the Probability of Pipe Rupture at Various Locations in the Primary Coolant Loco of a Babcock and Wilcox 177 Fuel Assembly Pressurized Water Reactor - Including the Effects of a Periodic Inspection, SAI-050-77-PA, September 1976 6.
An Analysis of the Relative Probability of Pipe Ruprare at Various Locations in the Primary Cooling Loop of a Pressurized Water Reactor Including the Effects of a Periodic Inspection, SAI 001-PA, June 1976 References 1 through 5 present the chronology of regulatory requests and response regarding the adequacy of reactor vessel supports assuming an instantaneous double-ended guillotine break inside the reactor cavity. Our last letter to you (Ref. 4) indicated participat!.on of Metropclitan Edison Company in an Owner'r. Group that is investigating this problen.
In August 1976, this Group discussed the analytical options with a brand spectrum of consultants. Our conclusion at that time was, that, given the disparity of opinion between the NRC staff, ACRS, and the analytical consultants regarding 1492 050
Mr. R. W. Reid October 6, 1977 GQL 1366 the appropriateness of the analytical tools, we would defer further analysis of the event until the appropriateness of the available analytical tools are demonstrated.
Instead, we propose for existing plants, such as TMI-1, to determine if such an event is likely enough to represent a risk to the health and safety of the public. We believe that it is possible to perform a quantitative study to determine if indeed the event was probable enough to require further analysis.
If the probaoility of the event is determined to be acceptably low, then further analysis is not warranted. The results of our quantitative analysis have been submitted as a Topical Report by Science Applications, Inc. (Ref. 5).
The report uses similar methodology as a report submitted by the Combustion Engineers User's Group (Ref. 6), but has taken advantage of experience gained in preparation of that report and expands the depth of analysis. The principal features that have been added to the 177 FA report are:
1.
Considering the critical length of through wall defects to vary with the maximum stress level, rather than using a constant conservatively estimated lower bound value.
2.
Considering the material properties, stress levels, and parameters in the detection probabilities and initial defect size distribution to be statistically fixed. Monte Carlo calculations of the distribution of the failure rates then allowed the degree of conservatism of the results to be quantitatively estimated.
3.
Refinements in the analysis of the effects of in-service inspection on the failure probabilities allowed more accurate treatment of the effects of ISI.
4.
Addition of further data to the base of the report.
5.
Factoring in of the effect of ultrasonic inspection on rupture probabilities.
By this letter, we are providing our endorsement of the SAI iteport and its con-clusions. We believe that it fully satisfies your request for information con-cerning the adequacy of Reactor Vessel Supports.
We believe our design and operation of TMI-1, consistent with the SAI Report, provides sufficient assurance that there is no undue risk to public health
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- Mr. K. W. Reid October 6,1977 GQL 1366 and safety and consider this matter to be adequately resolvea. We are interested in discussing the salient features of this report with you at your earliest convenience.
Very truly yours, J. G. Herbein Vice President JGH:JRS:dkf 1492 052