ML19326C244
| ML19326C244 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/23/1977 |
| From: | David Williams ARKANSAS POWER & LIGHT CO. |
| To: | Desiree Davis Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8004220816 | |
| Download: ML19326C244 (4) | |
Text
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NRCPORM 196 U.S. NUCLEAR REGULATODY COMMISSliN DOCK 27 NUMBED y "'SD - 3 r3 n.se s,
NRC DISTRIBUTION PoR PART 50 DOCKET MATERI AL TO:
FROM:
DATE OF DOCUMENT Arkansas Power & Light Co.
11/23/77 Mr. D. K. Davis Little Rock, Arkansas o4TE RECEIVE Daniel H. Williams 11/28/77 ML TrER ONOTORIZED PROP l INPUT PORM NUMBER OF COPIES RECEIVED ORIGIN AL NcLA333 p p g D OCOPy l ggg DESCRIPTION ENCLOSU RE Consists of information concerning the Unit No.1 Reactor Vessel Supports....
PLANT NAME: Arkansas Unit No.1 RJL 11/28/77 (3-P)
DISTRIBUTION FOR REACTOR VESSEL SUPPORT INFO FOR OPERATING REACTORS PER MR. TRAMMELL 7-12-76 A./D EAC SAfr u FOR ACTION /INFORMATION ASSIGNED ADt
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(([gjgggpy H ELPIN G BUILD ARKANSAS ARK ANS AS POWER S LIGHT COMPANY P O. BOX 551 LeT TL E AC' 'K. A AK ANS AS 72203.(5011371-4000 November 23, 1977
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1-117-12
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A Director of Nuclear Reactor Regulation 7 c
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ATTN:
Mr. D. K. Davis, Acting Chief g.
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Operating Reactors Branch #2 C,
i@N U. S. Nuclear Regulatory Commission h"90 m
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Washington, D. C.
20555 N. ' ; -
Subject:
Arkansas Nuclear One-Unit 1
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.. Docket No. 50-313 License No. DPR-51 Reactor Vessel Supports (File: 1510, 2205)
Ref:
A)
D. L. Ziemann to J. D. Phillips,
dated Oct. 17, 1975 B)
W. Cavanaugh to D. L. Ziemann, dated Nov. 21, 1975 C)
D. L. Ziemann to J. D. Phillips, dated June 9, 1976 D)
W. Cavanaugh to D. L. Ziemann, dated July 9, 1976 E)
W. Cavanaugh to D. L. Ziemann, dated Aug. 3, 1976 F) Topical Report by Science Applications, Inc., An Analysis of the Probability of Pipe Rupture at Various Locations in the Primary Coolant Loop of a Babcock 6 Wilcox 177 Fuel Assembly Pressurized Water Reactor-Including the Effects of a Periodic Inspection, Report No.
SAI-050-77-PA, September 1977 G) Topical Report by Science Applications, Inc.,
An Analysis of the Relative Probability of Pipe Rupture at Various Locations in the Primary Cooling Loop of a Pressurized Water Reactor Including the Effects of a Perodic Inspection, Report No. SAI-001-PA, June 1976.
Gentlemen:
This letter supercedes that letter as submitted to you on November 21, 1977, (our register number 1-117-11) concerning the captioned subject, and provides correction in references as discussed with your Mr. M. Connor on November 23, 1977.
775346159 T A x p AYING. INVESTOR OWNE D MEMBER MlOOLE SOUTH UTILITIES SYSTEM
- Mr. D. K. Davis Novcmber 23, 1977 References A) through D) present the chronology of regulatory requests and responses regarding the captioned subject. Reference E) indicated our participation in an Owner's Group (subsequently named the Babcock 6 Wilcox 177 FA Owner's Group - Engineering and Licensing) that was investigating this problem.
In August, 1976, this Group discussed the analytical options with a broad 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 the appropriateness of certain analytical methodologies, we would defer any further analysis using these methodologies for analysis of the postulated event until such time that the validity and appropriateness of the available methodologies are demonstrably annotated.
Instead, we proposed for existing plants, such as Arkansas Nuclear One-Unit 1, to determine if the probability of the postulated event (i.e., double-ended guillotine break inside the reactor vessel cavity) represented sufficient credibility so as to present a hazard to the health and safety of the public.
It was our belief, and this belief is currently maintained, that it is possible to perform a quantitative study to determine if indeed the probability indicated a need to perform further analysis.
The results of our quantitative analysis have been submitted to the NRC staff as a Topical Report by Science Applications, Inc. (Reference F).
This report utilizes similar methodology cs that used in the Combustion Engineering Reactor Owner's Group report (Reference G) as submitted to you. However, Reference F) has taken advantage of the experience gained in the preparation of Reference G) and expands the depth of analysis based on this experience. The principal features which have been additionally incorporated irc 9 Reference F) are:
1.
Considerug the critical length of through wall defects to vary with the maximum stress level, rather than using a constant conservatively estimated lower bound values.
2.
Considering the material properties, stress levels, and parameters in the detection probabilities and initial defect size distribution to be statistically distributed, rather than detenninistically 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 treatments 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.
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Mr..D. K. Davis November 23, 1977 With this letter, we hereby provide our endorsement of the SAI Report (Reference F) and its conclusions, and believe that it fully satisfies your request for. information concerning the adequacy of the ANO-1 Reactor Vessel Supports. We believe our design and operation of ANO-1, consistent with the SAI Report, provides even greater assurance that there is no undue hazard to the health and safety of the public and consider this matter to be adequately resolved. We are available to discuss with you the salient features of this report at your convenience.
Very truly yours,
- f. L Daniel H. Williams Manager, Licensing DHW:RMC:dr l
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