ML19210A664
| ML19210A664 | |
| Person / Time | |
|---|---|
| Site: | Crane, Arkansas Nuclear  |
| Issue date: | 01/25/1978 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | METROPOLITAN EDISON CO., SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 7910300694 | |
| Download: ML19210A664 (15) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION g
E WASHINGTON, D. C. 20555 e
.o January 25, 1978 h0*
All PWR Licensees (Except for Trojan, North Anna, Indian Point 3.
Beaver Miley and St. Lucie 1)
Gentlemen:
In October of 1975, the NRC staff notified each licensee of an operating PWR facility of a potential safety problem concerning the design of the reactor pressure vessel support system. Those letters requested each licensee to review the design basis for the reactor vessel support system for each of its PWR facilities to determine whether certain transient loads, which were described in the enclosure to the letter, had been appropriately taken into account in the design. Furthermore, these letters indicated that, on the basis of the results of licensees' reviews, a reassessment of the reactor vessel support design for each operating PWR facility may be required.
Licensee responses to that request indicated that these postulated asymmetric loads have not been considered in the design basis for the reactor vessel support system, reactor internals including the fuel, steam generator supports, pump supports, emergency core cooling system (ECCS) lines, reactor coolant system piping, or control rod drives.
Subsequently in June 1976, the NRC staff informed each PWR licensee that a reassessment of the eactor vessel support system design for each of its facilities was required. While the emphasis of these letters was primarily focused on the need to reassess the vessel support design for transient differential pressures in the annular region between the reactor vessel and the cavity shield wall and across the core barrel, we indicated that our generic review may extend to other areas in the nuclear steam supply system (NSSS) ard that further evaluation may be required.
For your information, Enclosure 1 is a summary of the background and current status of our review efforts related to this generic Concern.
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All PWR Licensees January 25, 1978 e
We have now determined that an assessment of the potential for damage to other NSSS component supports (e.g., steam generator and pump supports), the fuel assemblies, control rod drives, and ECCS piping attached to the reactor coolant system due to loadings associated with postulated coolant system piping breaks is required.
Our request for additional infomation transmitted to you in June 1976 has been revised both to clarify our original request and to identify the extension of our concerns to other areas in the NSSS, as identified above. A copy of this revised request for additional infomation is provided as Enclasure 2.
The revised request for additt nal infomation identifies a requirement that your assessment of potential damage to the reactor vessel and other NSSS component supports, reactor vessel, fuel and internals, attached ECCS lines and the control rod drives should include consideratio". of breaks both inside and outside of the reactor pressure vessel cavity. This assessment should be made for postulated breaks in the reactor coolant piping system, (secondary systems are not to be included), including the following Jocations:
a) Reactor vessel hot and cold leg nozzle safe ends b) Pump discharge nozzle c) Crossover leg d) Hot leg at the steam generator (B&W and CE plants only)
A number of licensees, have presented to the NRC staff alternate proposals, other than to conduct a detailed analyses, to resolve this concern. Based upon our review of these proposals, we have concluded that these alternative proposals do not establish an acceptable basis for long tenn operation without a detailed assessment of the risk resulting from these postulated transient loading conditions. We have, however, concluded that the low probability for occurrence of an event which could result in these loads establishes an adequate basis to justify continued operation for a short term period.
The NRC staff will consider an analysis that is applicable to more than one specific plant if it can be adequately demonstrated that such an analysis is either representative or bounding for each plant concerned.
Additional guidance regarding loading combinations (safe shutdown earthquake 1oads,1oss of coolant accident 1oads), will be provided by about March 1, 1978, following the conclusion of staff investigations in this area.
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