ML19199A366
| ML19199A366 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/07/1979 |
| From: | Carbon M Advisory Committee on Reactor Safeguards |
| To: | Hendrie J NRC COMMISSION (OCM) |
| References | |
| ACRS-0825, ACRS-825, NUDOCS 7904170042 | |
| Download: ML19199A366 (2) | |
Text
{{#Wiki_filter:. g[ ]o, UNITED STATES d- ~ -{ -h yp, p y,e/g- % NUCLEAR REGULATORY COMMISSION . ;.,;N.a n ) - r ADVISORY COMMITTEE ON REACTOR SAFEGUARDS (6bg N. D. C.,2 CSS 5 l g, O MSH1NG 3 0%'2.5 j Anril 7, 1979 gdd ~ Honorable Joseph M, Hendrie ,b. ??. j Chairman O i U.S. Nuclear Regulatory Cocctission 1 Washington, DC 20555 If E C l] SUaJECT: INTERIM REPORT CN RECE';T ACCIDENT AT WE TdREE MILE ISU.{!D 7 ] ~ NUCLEAR STATION UNIT 2 Dear Dr. Hendrie _1 D.: ring its 228th meeting, April 5-7, 1979, the Advisory Comittee on Reactor Safeguards reviewed the circumstances relating to the recent ( accident at the Tnree Mile Island Nuclear Station Unit 2. During this review, the conmtittee had the benefit of discussions with the h3C Staff. Our study of the accident at Taree Mile Island has shown that it is very difficult for a Ph2 plant operator to understand and properly control the course of an accident involving a small break in the reactor coolant system accompanied by other abnormal conditions. The Cocmittee recomends that further analyses be made, as soon as pas-sible, of transients and accidents in FWas that involve initially, or at some time during their course, a small break in the primary system. Tne computer codes used for these analyses should be capable of predict-ing the conditions observed during the accident at Tnree Mile Island, ' including thermal-hydraulic effects and clad and fuel temperatures. ($m 9 The range of break sizes considered should include the smallest that could be deemed significant, and should consider a range of break 1cca-tions. Tne Cocctittee' believes that the analyses recoc= ended above will de.mn-strate, as has the accident at Tnree Mile Island, that additional information regE.rding the status of the system will be needed in order for the plant operator to follcw the course of an accident and thus be~ able to resp 6nd in an appropriate manner. As a minimum, and in the interim, it muld be prudent to consider expeditiously the provision -Ad. p' Rj i cr 79041700 R O \\AdT q
l'. 4 %=...
- f..-
Honorable Joseph M. Hendrie - 2.-- April 7,197L m. 1 c! i,nstrumentation that will provide an unambiguous indication of the i level of fluid in the reactor vessel. Early consideration should be given also to providing remotely controlled means for venting high points in de reactor system, as practical. The foregoing recor:mendations apply to all pressuriced water reactors. The recom:endations in IE Bulletin 79-05A, dated April 5,1979, are be-lieved.to he generally saltable for Babco:k and Wilcox facilities, on an interim basis. However, the Co=nittee believes that the actions I listed in Item 4b. under the heading, " Actions To Be Taken by Licen-sees," may prove to be unduly prescriptive in view of the uncertainties in p::edicting de course of anomalous transients or accidents involving small breaks in the primary systen. With regard to Three Mile Island Unit 2, the Co=nittee believes that decisions should be made expeditiously with regard ~to contingency ceas-h ures which may be prudent concerning contain:nant and reactor cooldown as a backup to the currently planned cooldcWn procedure. The Co=nittee is continuing its review of these and other concerns arising frcm this accident and will provide further advice as it is developed. Sin erely, 1 P.ax W. Carbon Chairman 0 D / ' f 'G(j ] .e' a S .*....3. t s. 4 *<
- '* s;1R&QQ&Wg Wh$QCs'yMG*L&*4 r.Mh*Y-M GwN%i=.?,.,.
~ .i c.t.4T5 ?- ??? Wh$ 2""[ T& GJ-M t' T*Mw%pTK*'-WQ%K Gmww1a m nmmsm_wra-M MsEA 's GTTM'c D <*.'?}}