IE Circular No. 81-11, Inadequate Decay Heat Removal During Reactor Shutdown

                                                          SSINS No.:  6830 
                                                          Accession No.:   
                                                          IEC 81-11        

                               UNITED STATES 
                           WASHINGTON, D.C. 20555 

                               July 24, 1981 



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 

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                                                           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 

                                                           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 

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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                                                           July 24, 1981  
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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 

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