United States Nuclear Regulatory Commission - Protecting People and the Environment

Multiple-subsequent Actuation of Safety/relief Valves Following an Isolation Event (Generic Letter 78-09)

                                UNITED STATES 
                        NUCLEAR REGULATORY COMMISSION 
                           WASHINGTON  D. C. 20555  

                                        March 20, 1978



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.


                                      Victor Stello, Jr. Director
                                      Division of Operating Reactors

Criteria for the Assessment of Multiple Subsequent SRV Actuations



                         CRITERIA FOR THE ASSESSMENT 

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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