Bulletin 79-12: Short Period Scrams at BWR Facilities
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
May 31, 1979
IE Bulletin No. 79-12
SHORT PERIOD SCRAMS AT BWR FACILITIES
Reactor scrams, resulting from periods of less than 5 seconds, have occurred
recently at three BWR facilities. In each case the scram was caused by high
flux detected by the IRM neutron monitors during an approach to critical.
These events are similar in most respects to events which were previously
described by IE Circular 7707 (copy enclosed).
Description of Circumstances:
The following is a brief account of each event.
1. Oyster Creek - On December 14, 1978, the reactor experienced a scram as
control rods were being withdrawn for approach to critical, following
a scram from full power which had occurred about 15 hours earlier. The
moderator temperature was 380 degrees F and the reactor pressure was
190 psig. Because of the high xenon concentration the operators had
not made an accurate estimate of the critical rod pattern. The
operator at the controls was using the SRM count rate, which had
changed only slightly, (425 to 450 cps) to guide the approach. Control
rod 10-43 (first rod in Group 9) was being withdrawn in "notch
override" to notch position 10, when the reactor became critical on an
estimated 2.8 second period. The operator was attempting to reinsert
the rod when the scram occurred. Failure of the "emergency rod in"
switch to maintain contact due to a bent switch stop, apparently
contributed to the problem.
2. Browns Ferry Unit 1 - On January 81, 1979, the reactor experienced a
scram during the initial approach to critical following refueling. The
operator was continuously withdrawing in notch override" the first
control rod in Group 3 (a high worth rod) because the SRM count rate
had led him to believe that the reactor was very subcritical. A short
reactor period, estimated at 5 seconds, was experienced. The operator
was attempting to reinsert control rods when the scram occurred.
IE Bulletin No. 79-12 May 31, 1979
Page 2 of 3
3. Hatch Unit 1 - On January 31, 1979, the reactor experienced a scram
during an approach to critical. Control rod 42-15 (fifth rod in Group
3) was being continuously withdrawn in "notch override" when the scram
occurred, with a period of less then 5 seconds. The temperature was
about 200 degrees F with effectively zero xenon.
As indicated above, these short period trips occurred under a wide variety
of circumstances. They did have several things in common, however. In none
of these cases was an accurate estimate of the critical position made prior
to the approach to critical. In each case a rod was being pulled in a high
worth region. Finally, in each case the operator, believing that the reactor
was very subcritical, was pulling a rod on continuous withdrawal.
Action to be Taken by Licensees:
For all GE BWR power reactor facilities with an operating license:
1. Review and revise, as necessary, your operating procedures to ensure
that an estimate of the critical rod pattern be made prior to each
approach to critical. The method of estimating critical rod patterns
should take into account all important reactivity variables (e.g., core
xenon, moderator temperature, etc.).
2. Where inaccuracies in critical rod pattern estimates are anticipated
due to unusual conditions, such as high xenon, procedures should
require that notch-step withdrawal be used well before the estimated
critical position is reached and all SRM channel indicators are
monitored so as to permit selection of the most significant data.
3. Review and evaluate your control rod withdrawal sequences to assure
that they minimize the notch worth of individual control rods,
especially those withdrawn immediately at the point of criticality.
Your review should ensure that the following related criteria are also
a. Special rod sequences should be considered for peak xenon
b. Provide cautions to the operators on situations which can result
in high notch worth (e,g. first rod in a new group will usually
exhibit high rod worth).
4. Review and evaluate the operability of your "emergency rod in" switch
to perform its function under prolonged severe use.
IE Bulletin No. 79-12 May 31, 1979
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5. Provide a description of how your reactor operator training program
covers the considerations above (i.e., items 1 thru 3).
6. Within 60 calendar days of the date of issue of this Bulletin, report
in writing to the Director of the appropriate NRC Regional Office,
describing your action(s) taken, or to be taken, in response to each of
the above items. A.copy of your report should be sent to the United
States Nuclear Regulatory Commission, Office of Inspection and
Enforcement, Division of Reactor Operations Inspection, Washington,
For all BWR facilities with a construction permit and all other power
reactor facilities with an operating license or construction permit, this
Bulletin is for information only and no written response is required.
Approved by GAO B180225 (R0072); clearance expires 7/31/80. Approval was
given under a blanket clearance specifically for identified generic
1. IE Circular No. 77-07
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