ML20140F580
| ML20140F580 | |
| Person / Time | |
|---|---|
| Site: | Monticello, 05000000 |
| Issue date: | 06/06/1977 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Sobon L GENERAL ELECTRIC CO. |
| Shared Package | |
| ML20140F372 | List:
|
| References | |
| FOIA-85-665 NUDOCS 8604010096 | |
| Download: ML20140F580 (2) | |
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JUN 6 1977 Mr. L. J. Sobon, Manager BhR Containment Licensing Mail Code 682 General Electric Company 175 Curtner Avenue San Jose, California 95125
Dear Mr. Sobon:
By letter dated March 3,1977, (L. J. Sobon to R. Tedesco), General Electric addressed two areas of conceni relating to quencher loads on containment structures. They were the identification of the primary parameter associated with the subsequent SRV actuation effect and the need for in-plant testing of leaking SRV's. A qualitative assessment of several parameters, which could potentially influence subsequent activation SRV loads, was provided. Based on this evaluation, General Electric concluded that the SRV line temperature is the key parameter.
Following this rationale, General Electric further concluded that the leaking SRV would result in lower loads on structures than the first actuation without leakage.
Based on our evaluation of the provided information, we find that the conclusions reached have not been justified. In particular, we cannot agree that based upon the available infonnation that the leaking SRV would result in lower loads. Therefore, we continue to believe that future test programs should include a testing phase to investigate the leaking SRV effect. We have reached these conclusions for the following reasons.
1.
All SRV parameters were not included in the qualitative assessment.
For example, the initial in-line air / steam mixture temperature was not included. This parameter, we believe, is rather important; since the Monticello test data (NEDE-21465, Preliminary Report In-Plant Safety / Relief Valve Discharge Load Test - Monticello Plant) indicate that this temperature for the subsequent actuation is consistently higher than the first actuation. This tenperature change should have an effect on the load increase for subsequent actuation, and therefore, cannot be precluded from the evaluation.
2.
There is no known experimental evidence or analytical model to sy port your qualitative assessment that SRV line temperature is the key parameter. The experiments, which have been conducted either for quencher or for ramshead, were not designed to isolate 8604010096 860114 PDR FOIA FIRESTOO5-665 PDR do
Mr. L. J. Sobon
'In fhe effect of specific paraneters on sesequent actuation.
addition, the analytical sedel for a ramshead devim is still the subject of an ongoing effort. Therefore, the analytical model alone camot be used to support this conclusion.
The canclusion that a leaking SRV should result in lower loads 3.
The does not appear to agree with the Monticello test results.
test data indicate that a leaking SRV could result in a load increase ranging frun 134 to 25% in + -ison with those tests without steam leakage.
Since there remains significant uncertainty an whether or not a 1*Mng SRV would increase hydrodynamic loeds at structures, we believe future SRV test p.us. w should include an investigation into this area..
Specifically, provisions within the Caorso test plan should be aMa4 to include leaking SRV effects since this is the next scheduled SRV test y.e a s..
s Robert L. Tedesco, Assistant Director for Plant Systens Di' vision of Systems Safety Office of Nuclear Reactor Regulation Distribution:
Central Files NRR Reading PS Reading CSB Reading G. Lainas n
J. Kudrick C. Grimes J. Guibert T. Su S. Miner
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