IE Circular No. 81-11, Inadequate Decay Heat Removal During Reactor Shutdown
SSINS No.: 6830
Accession No.:
8011040256
IEC 81-11
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
July 24, 1981
IE Circular No. 81-11: INADEQUATE DECAY HEAT REMOVAL DURING REACTOR SHUTDOWN
Background:
Following several losses of decay heat removal capability at operating
pressurized water reactors (PWRs), IE Bulletin 80-12 "Decay Heat Removal
System Operability" (issued May 1, 1980) requested PWR licensees to take
certain actions intended to reduce the probability of loss of decay heat
removal. All operating PWRs were requested to amend the Technical
Specifications for their facilities with respect to reactor decay heat
removal capability by letter from D. Eisenhut, Division of Licensing, on
June 11, 1980. IE Bulletin 80-12 was issued to boiling water reactor (BWR)
licensees for information with the expectation that the information would be
evaluated for applicability and subsequent action taken as determined
necessary. However, events involving inadequate decay heat removal at
operating BWRs now indicate the need for BWR licensees to provide additional
controls related to decay heat removal.
Description of Circumstances:
1. Brunswick - Temporary Loss of Shutdown Cooling
On December 8, 1980, unplanned heatup of the reactor coolant occurred
at Brunswick Unit 2 when the unit was in cold shutdown (F212F)
with all rods inserted. The heatup occurred while the service water
cooling for the "A" loop of the residual heat removal (RHR) system was
isolated longer than expected for repair of a service water leak.
Shutdown cooling was not lined up to loop "B" (1) because it was
expected that loop "A" would be returned to service before 212F
was reached and (2) because of the length of time required to line up
the "B" loop for operation. During the repair, the recirculation pumps
were off, an RHR pump was running, and the control rod drive pump was
supplying water to the reactor pressure vessel (RPV) while the reactor
water cleanup (CU) system was rejecting water for level control. The
reactor coolant temperature monitored at the CU inlet (from a
recirculation loop) indicated F212F during the repair. The reactor
head vents were reported to be opened during this period, with no
evidence of steaming. However, average coolant temperature at the time
of completion of repair approached 212F with an observed maximum
of 217F. Shutdown cooling was initiated and primary coolant
temperature decreased to a normal temperature within approximately 30
minutes. Primary containment could not be quickly established due to
cables going through the personal access hatch and the torus hatch
being removed.
A similar event occurred at Brunswick Unit 2 on the following day. With
the primary containment and reactor head vents reported open, the
conventional and nuclear service water systems were secured to repair
a conventional service water pump discharge check valve. The primary
coolant
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IEC 81-11
July 24, 1981
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temperature initially was less than 120F. Approximately two hours
after the service water systems were secured, the RHR pumps in the A
loop were secured to reduce coolant heat input from the pumps.
Repairs took longer than anticipated, and when the conventional and
nuclear service water systems were returned to service, the primary
coolant temperature at the vessel bottom head drain was 147F.
Approximately fifteen minutes later shutdown cooling was initiated
using the B loop of the RHR. There were indications of heatup of the
coolant to approximately 212F; however, there was no evidence of
steaming through the open reactor heat vents. Primary coolant
temperature decreased to a normal temperature within approximately
three hours.
2. Dresden Unit 3 - Unplanned Repressurization
On December 20, 1980, the Dresden Nuclear Power Station Unit 3 was in
the cold shutdown condition. Numerous maintenance and modification
outages were in progress which resulted in the shutdown and/or
isolation of all systems which communicate with the reactor vessel, and
which normally provide cooling and recirculation of the primary
coolant. Subsequently, one of three loops of the shutdown cooling
system (SDC) was put in service to maintain reactor water temperature
at approximately 150F. The reactor water level was maintained at
the normal operating level (instead of flooding up) to limit vessel
safe end thermal stresses.
Because the design of the SDC does not allow for throttling of the
cooling water flow to the SDC heat exchangers, it is standard practice
to throttle SDC flow to the recirculation loop to maintain vessel
temperature when in cold shutdown. As the decay heat load decreased the
unit operators reduced SDC flow until insufficient vessel flow existed
to provide mixing of the primary coolant, and accurate temperature
measurements by the recirculation pump and SDC pump suction temperature
instruments. Because the operators monitored only the recirculation
pump and SDC temperatures, a slow heatup and repressurization of the
reactor vessel to 175 psig occurred over a six hour period of time.
Upon discovering the repressurization, SDC flow was increased, and a
second SDC loop was placed in service to expedite the return to cold
shutdown. The indicated recirculation suction temperature rose to
approximately 225F, indicating that the entire vessel contents did
not heat up to the saturation temperature at 175 psig (377F).
During the repressurization event the containment personnel access
doors were open, resulting in violation of the Technical Specification
limiting condition for operation for primary containment integrity. Had
the Technical Specification been revised to conform to current BWR
standard Technical Specifications the LCO's for the High pressure
coolant injection system and isolation condenser systems would also
have been exceeded.
Post event evaluations of the circumstances leading up to the
repressurization, and the chronology of the event itself, establish
that the
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IEC 81-11
July 24, 1981
Page 3 of 4
licensee did not evaluate the potential for adverse effects on plant
safety resulting from procedure changes removing the vessel floodup
requirement, and the effect of removing from service those systems
which normally cool and recirculate the reactor coolant. The potential
for inaccurate response of normally used instrumentation was apparently
not considered by the licensee, and redundant instrumentation which
could have provided warning that the event was in progress was not
utilized by operations personnel.
The licensees of the above facilities have committed to make administrative
and procedural changes to provide personnel additional guidance when
operating in the shutdown cooling mode. Additional information regarding
these events and corrective actions is contained in LERs 2-80-107, 2-80-112
(Brunswick 2), and LER 80-047/01T-0 (Dresden 3).
There have been recent events at other BWRs involving the loss of systems
providing normal decay heat removal, and appropriate action has been taken
by operating personnel to put alternate cooling in service. These events
indicate the need for timely operator response and the need to have backup
systems available.
Recommended Action for Licensees of BWRs with an Operating License:
1. Review your existing procedures and administrative controls that relate
to decay heat removal during reactor shutdown. Analyze these procedures
for adequacy of monitoring and responding to events involving lost or
degraded decay heat removal. Special emphasis should be placed on
conditions involving low core recirculation or cooling flow, or when
maintenance or refueling activities degrade the decay heat removal
capability.
2. Administrative controls should provide the following:
a. Assure that redundant or diverse decay heat removal methods are
available during all modes of plant operation. (Note: When in a
refueling mode with water in the refueling cavity and the head
removed, an acceptable means could include one decay heat removal
train and a readily accessible source of water to replenish any
loss of inventory). (Note: Only one power source needs to be
operable in order to consider the decay heat removal system
operable while in modes 4 and 5).
b. For those cases where single failures or other actions result in
only one decay heat removal train being available, provide an
additional alternate means of decay heat removal or provide an
expeditious means for the restoration of the lost train or method.
c. Implement administrative controls during periods of low flow or no
flow to ensure that the maximum coolant temperature remains below
the saturation temperature. Consideration should be given to
maintaining water level in the reactor vessel sufficiently high to
enable natural circulation at all times.
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IEC 81-11
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d. Require monitoring of the reactor coolant temperature and pressure
at a specified frequency.
3. Any changes needed in the existing procedures or administrative
controls as a result of Items 1 and 2 above should be implemented
within 120 days of the date of this circular.
No written response to this circular is required. If you need additional
information regarding this subject, please contact the appropriate Regional
Office.
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