ML19326C240
| ML19326C240 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/21/1977 |
| From: | David Williams ARKANSAS POWER & LIGHT CO. |
| To: | Desiree Davis Office of Nuclear Reactor Regulation |
| References | |
| 1-117-11, NUDOCS 8004220813 | |
| Download: ML19326C240 (4) | |
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NRC romu 196 u.s. NUCLEAA REoULATOnY COMA eSioN ooCKET NUM sE m
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NRC DISTRIBUTION pon PART 50 DOCKET MATERIAL TO:
FROM:
oArm op ooCUMENT Mr. D. K. Davis Arkansas Power & Light Company 11/21/77 Littie Rock, Arkansas o47,,geg,y g o Daniel H. Williams 11/22/77 2C Lartsa ONoronizao amor iNeureoeu Nuusen or copies mEcsivso Womio8NAL fifuncLAssiciso CCopy jgggalP ossCnierioN ENCLoSumE Consists of info. re. the adequacy of the ANO-1 Reactor Vessel Supports..
THIS DOCUMEllT C0t1TAlliS P00R QUALITY PAGES PLANT NAME: Arkansas Unit No.1 (3-P)
RJL 11/22/77 DISTRIBUTION FOR REACTOR VESSEL SUPPORT' INFO FOR OPERATING REACTORS PER MR. TRAMMELL 7-12-76
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H EL PIN G BUILO A R KA NS A S ARK ANS AS POWER & LIGHT COMPANY 880. Box 551 LITTLE AOCK. A AM ANS AS 72203 85011371 + 4000 November 21, 1977 f(
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J o/S Director of Nuclear Reactor Regulation A'ITN: Mr. D. K. Davis, Acting Chief p9
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Washington, D. C.
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Subject:
Arkansas Nuclear One-Unit 1 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 A;g. 3, 1976 F) Topical Report by Science Applications, Inc., An Anslysis of the Probability of Pipe Rupture a*. Various Locations in the Primary Coolant Loop of a Babcock 6 P'ilcox 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 Varioes 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:
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 7732c.vvso T AX P AYING, INVESTOR OWNED MEMBER MlOOLE SOUTH UTILITIES SYSTEM
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Mr. D..K. Davis November 21, 1977 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 nur 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 would not be warranted.
The results of our quantative analysis have been su'bmitted to the NRC staff as a Topical Report by Science Applications, Inc. (Reference 5).
This report utilizes similar methodology as that used in the Combustion Engineering Reactor Owner's Group report (Reference F) as submitted to you. However, Reference-E) hrs taken advantage of the experience gained in the preparation of Reference F) and expands the depth of analysis based on this experience. The principal features which have been additionally incorporated into Reference E) are:
1.
Considering the critical length of through wall defects co vary with the maximum stress level, rather than using a constant conservatively estimated lower bound values.
1 2.
Considering the material properties, stress levels, and parameters in the detection probabilities and initial defect size distribution to be statistically distributed, rather than deterministicaly fixed. Monte Carlo calculations of the i
distribution of the failure rates then allowed the degree of j
conservatism of the results to be quantitatively estimated.
i 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.
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Factoring in of the effect of ultrasonic inspection on rupture probabilities.
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Mr. D. K. Davis November 21, 1977 With this letter, we here f provide our endorsement of the SAI Report (Reference E) and its conclusions, and believe that if 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.
Ve ly you ac Daniel H.
iams Manager, Licensing DHW:RMC:dr m
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