ML20037B582
| ML20037B582 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/13/1974 |
| From: | Stratton W Advisory Committee on Reactor Safeguards |
| To: | Nader R AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20037B577 | List: |
| References | |
| RTR-WASH-1285 NUDOCS 8010210870 | |
| Download: ML20037B582 (3) | |
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<c ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l-UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20545
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Wuch 13, 1974
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I Hr., Ralph Nader 7th Floor 2000 P Street, N. W.
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Washington, D. C. 20036
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Dear Mr. Nader:
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1 The Advisory Committee on Reactor Safeguards wishes to acknoul
_ ~- 9 receipt of your letter of February 28, 1974 which referred to recent ACRS report, " Integrity of Reactor Vessels for Light-W--
.:.; s Power Rea,ctors," (report WASH-1285) and to concerns you have
-!:E regarding some of the points of the report.
As acknowledged in your le'tter, the ACRS has s*tatt: tory resporc ity to advise the Atomic Energy Comission regardi:
the har-proposed or existing reactor facilities. As an air in dischc these responsibilities, we welcome any information which has t bearing on nuclear safety.
In'particular, we invite your de:. '
written technical comments related to these' matters.
Uc issue, in the Federal Register, invitations for presentcti-comments, both written and oral, by the public on matters af.'.
nucicar safety, as well as for public attendance at our mee.ti:-
When this procedure does not suffice for presentation of spee!
technical.information, special arrangements can be cad'c for a nical oral presentation of such inforcation to members of an <
priate ACRS Subcommittee or occasionally to the Cccaittee as,
t Our statutory responsibilities do not include responding to i:
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tories from members of the public directly' and our work load :.
precludes detailed replies to information and inquiries submi!
us.
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Mr. Ralph Natler March 13, 1974
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Your letter mentions four specific " questions" regarding pressure vessel integrity. These relate to:
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.;m (1) the probability of " disruptive failure of presstire vessels, si;a=;;d
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Ehe possible receipt by the ACRS of a report by Sir Alan Ctrell to the Select Committee on Science and Technology
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(3) the status of early pressure vessels,
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=j (4). the performance of pressure vessel designers, manufacturers,
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- A The following comments are offered concerning these questions:
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- =gg Question 1
It was not' part of the purpose of the Committee's report on
!h the integrity of reactor vessels to attempt to, quantify an " acceptable g
level of risk." It was part of the purpose, however, to attempt, as far E
as practical with existing data, to quantify the reliability of reactor vessels. This the Committee did and concluded, as stated in the report, that there is reasonable assurance that "the disruptive failure rate of reactor vessels designed, constructed, and operated in accordance with ASME Code Sections III and XI is less than 1x10-6 per vessel-year" and that "the probability of reactor vessel disruptive failure beyond the capability of Engineered Safety Features is lower than the probability for disruptive failures of all kinds." It would be this lower probabil-
]j ity which is relevant to considerations of accidents exceeding the limits
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of 10 CFR Part 100 normally used in accident analysis.
Question 2: The Committee received.a copy of the five-page report,
" Fracture of Steel' Pressure Vessels" submitted to the Sc1cet Committee' ui=
on Science and Technology by Alan Cottrell, dated 22 Janut.ry,1974, af ter submission of WASH-1285 to Chairman Dixy Lee Ray of the USAEC.
The conclusions in Dr. Cottrell's report are in general agreement with those of t;he ACRS. The ACRS report emphasizes the importance of limit-ing flaw size and avoiding flaw growth by careful control'of environ-mental conditions.
Periodic inspection as required by ASME Code Section m
XI is indicated in the ACRS report to be a major factor in assuring
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Ihrch 13, 1974 Es@
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Mr. Ralph Nader J.:!:::.se anditions of significance to
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vessel reliability by detec. ting flaw cThe' Committee did consider in p --:7 cions of linear clastic fracture mechan ctor pressure vessel steels, t ji{s:.;,Z vessel fracture.
specifically applied to nuclear reac d to the spectrumident conditions considered i extensive Heavy Sec-
((.x possibic degradation mechanisms, an by startup, shutdown, operating and acc l
The Committee believes, based ondirectly relevant to assurance that
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safety reviews.
tion Stec1 Technology studies there is reasonable are acceptably 9E*-
postulated failures tional
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steels used in the United States, of materials properties and operal stic-pla the critical flaw sizes leading to d
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large for the combination i
l loads under relevant clastic or e a
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throughout reactor life.
h were not built to ump.y W:es and those which have been significantly irra With regard to reactor vessels whic 2:E-
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ctrrent codCocmittee is sending a copy of your who may have such a list.
liability depends not I
W: proper codes and standards but also upo The Committee agrees that vessel re in which the are carried out and upon the way
- eport, In Sections 2.7, 6.2, and 6.4 of the r only upon and construction ized these points and made someT design vessels are operated.
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h among others, the Committee emp as
.;;3 adequacy gard.ibility for ensuring the specific recommendations in this reAEC Regulato of detailed adherence to i+
Sincerely yours,
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y, g. Stratton
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