NRC Generic Letter 78-09, Multiple-Subsequent Actuations of Safety/Relief Valves Following an Isolation Event

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WASHINGTON D. C. 20555

March 20, 1978

Gentlemen:

RE: MULTIPLE-SUBSEQUENT ACTUATION OF SAFETY/RELIEF VALVES FOLLOWING AN ISOLATION EVENT

In a meeting on October 27, 1977, the General Electric Company (GE) and the Mark I Owners Group provided the staff with the results of an assessment of the effects of multiple-subsequent actuation of safety/relief valves (SRVs)

following an isolation event. This assessment was provided to justify the deferral of this issue until its ultimate resolution as a part of the Mark I Containment Long Term Program. At the conclusion of that meeting, the staff requested that each utility submit a basis for continued operation by November 1, 1977 including a description of any interim corrective measures which may be implemented. The staff further indicated that it may require plant unique assessments to be provided in the near future. A number of the submittals made on November 1, 1977 contained additional information relative to the effects of multiple subsequent SRV actuation.

The assessments that we have received to date have been based on an application of the results of the Monticello SRV discharge (ramshead) tests. During the course of our review of the Monticello test results, we have noted that there are significant variations in the measured structural responses for similar test conditions. As a result, we have concluded that the data base is insufficient to determine the probability distribution for either

(1) the structural responses for similar test conditions Or
(2) the manner by which structural responses for single SRV actuation are to be combined in determining the structural responses to several SRVs discharging simultaneously. Further, in assessing the effects of multiple SRV actuation, the structural responses to single SRV actuation do not combine consistently at various points on the structure, when compared to the responses for the same valves discharging simultaneously.

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We recognize that, at the present time, the Monticello test results provide the best available data for determining the effects of multiple subsequent SRV actuation. However, the application of the Monticello test results involves a considerable amount of subjective judgement. We have, therefore, developed the enclosed criteria, based on our interpretation of the Monticello data, which we believe will provide a "most probable" estimate of the effects of an isolation transient event. In our view, such an estimate is consistent with the philosophy of the Mark I Containment Short Term Program and is acceptable on an interim basis, while the Long Term Program is being conducted.

The enclosed criteria should be used to perform a plant unique assessment of this concern as it relates to Mark I BWR facilities. You are requested to submit this assessment for your facility within 60 days of the receipt of this letter. Since over 100 of these transient events have occurred for which only two events resulted in multiple subsequent SRV actuation, and since no evidence of structural deterioration was found, we conclude that continued operation is acceptable while this assessment is being performed. Your submittal should include a description of the methods used to satisfy these criteria. Where appropriate, plant unique data may be used for this assessment, provided that the test procedures and data are documented.

Sincerely,


Victor Stello, Jr. Director Division of Operating Reactors

Enclosure:

Criteria for the Assessment of Multiple Subsequent SRV Actuations

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Enclosure

CRITERIA FOR THE ASSESSMENT OF MULTIPLE-SUBSEQUENT SRV ACTUATIONS

1. The number of valves which experience subsequent actuation shall be determined from a plant-unique assessment of the transient which reflects the valve groupings and the SRV setpoints in your facility's Technical Specifications. Variations in the SRV setpoints may be accounted for, provided all of the setpoints are distributed in a manner dictated by actual SRV performance testing. Plants with similar SRV discharge arrangements may be grouped for this assessment provided their similarity is demonstrated.

(Although discussions are currently being held between GE and the staff regarding the transient analysis models used to predict the SRV response sequence, we conclude that the current models are acceptable for this interim assessment. The ultimate resolution of this issue in the Long-Term Program will require the use of transient analysis models which resolve staff concerns regarding the current models.)

2. The plant specific variations to the hydrodynamic characteristics of the SRV discharge line configurations shall be accounted for by the use of a correction factor derived from the SRV discharge analytical model. This factor shall be based on average line conditions for those lines predicted to subsequently actuate, as compared to the Monticello "Bay D" discharge conditions. The basis for averaging shall be described and justified.
3. All available peak structural response data for single SRV discharge events, with approximately the same distances between the discharge point and the structures, should be averaged to obtain the expected values of peak structural response at that point as a function of its distance from the discharging SRV. Certain data may be omitted if it can be demonstrated that such data are inconsistent and should not be considered.
4. The effects of a multiple valve discharge event, as determined from the data on individual SRV discharges, shall be determined by taking the SRS of the individual valve effects and increasing this value by 20 percent, except as noted in
(5) below.

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5. For structures excited primarily by the overall movements of the torus (e.g., the suction header, the torus support columns, the ring header, etc.),

the absolute sum of the structural responses to single SRV actuations shall be used to determine the effects of the same valves actuating simultaneously.

6. The consecutive valve actuation factors shall be determined from the Monticello data, or any other available test data, by considering the peak structural responses for an appropriate set of gauges for all consecutive valve actuation tests. For a given set of gauges, the mean plus one standard deviation of all peak structural responses, shall be utilized to compute a set of consecutive actuation factors. These consecutive valve actuation factors shall be averaged to determine one consecutive valve actuation factor which is applicable to the area(s) of the structure of which this set of gauges is appropriate. Certain data may be omitted if it can be demonstrated that such data are inappropriate and should not be considered.
7. If the results of this assessment indicate that the limiting strength ratio for either the torus shell of the torus support system is greater than 0.5, corrective measures should be promptly instituted to reduce the limiting strength ratio for either the torus shell or the torus support system is greater than 0.5, corrective measures are necessary, for your facility your submittal should describe proposed corrective measures, including the associated schedule for their completion.
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