Information Notice No. 93-45: Degradation of Shutdown Cooling System Performance

                                UNITED STATES
                           WASHINGTON, D.C.  20555

                                June 16, 1993

                               SYSTEM PERFORMANCE


All holders of operating licenses or construction permits for nuclear power


The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees to degradation of shutdown cooling system
performance at the Oyster Creek Nuclear Generating Station resulting from
inadequate operating procedures.  It is expected that recipients will review
the information for applicability to their facilities and consider actions, as
appropriate, to avoid similar problems.  However, suggestions contained in
this information notice are not NRC requirements; therefore, no specific
action or written response is required.

Description of Circumstances

On January 23, 1993, the Oyster Creek Nuclear Generating Station was in day 57
of the 14R refueling outage.  The plant was in a normal cold-shutdown
condition with the primary containment open and the reactor coolant system at
about 43 C [110 F].  One reactor recirculation loop was in service, one
recirculation loop was open, and the remaining three recirculation loops were
either idled or isolated.  Two shutdown cooling loops were operating with a
combined flow rate of 11,735 liters per minute [3,100 gpm].  

In order to support the planned activities, the licensee elected to secure
operation of all recirculation pumps and lower reactor water level.  This
configuration would allow completion, in parallel, of recirculation pump
maintenance activities and main steam isolation valve local leak rate testing. 
When the operators lowered the reactor water level to approximately 419 cm  
[165 inches] above the top of active fuel and secured the running
recirculation pump, the shutdown cooling water took the least resistance path
through the open recirculation loop, and the majority of flow bypassed the
core (see Attachment 1).  During the time that the planned activities were in
progress, plant personnel did not realize that shutdown cooling system
performance was degraded, allowing unmonitored heatup of the reactor core.  

On January 25, 1993, after completing the main steam isolation valve local
leak rate testing, an operations engineer discovered that the reactor vessel
metal temperature was about 109 C [228 F] at the mid-vessel point.  Several 


                                                            IN 93-45
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Technical Specifications prerequisites for exceeding 100 C [212 F] were not
met, including the requirement to establish primary containment integrity.  
After discovery, operators took immediate measures to reduce reactor coolant
system water temperature by first maximizing flow in the two operating
shutdown cooling loops and then placing a third shutdown cooling loop in
service.  Additionally, the reactor water level was raised to approximately
508 cm [200 inches] above the top of active fuel and a recirculation pump was

The immediate cause of the event was determined to be plant conditions
established by a temporary procedure change to the shutdown cooling operating
procedure that failed to provide sufficient forced flow through the reactor
core to prevent thermal stratification.  The temporary procedure change had
been implemented to allow shutdown operations with all recirculation pumps
secured and the nominal reactor water level less than the level at which
spillover from the core region to the annulus is assured (approximately 470 cm
[185 inches] above the top of active fuel).  The change had been developed
from a draft engineering evaluation, but did not include a requirement (stated
in the body of the evaluation) to maintain a specified shutdown cooling flow
of 22,712 liters per minute [6,000 gpm].  Instead, the usual shutdown cooling
flow of about 11,735 liters per minute [3,100 gpm] was maintained which, due
to the specified configuration, allowed the inadvertent reactor heatup.

Further details can be found in Licensee Event Report 50-219/93-002 and NRC
Augmented Inspection Team Inspection Report No. 50-219/93-80.


This event at Oyster Creek indicated that the temporary procedure change
represented a significant procedure revision and that the review process did
not identify deficiencies in the procedure before it was implemented.  The
temporary procedure change had been developed from a draft engineering
evaluation which concluded that two shutdown cooling loops in operation would
adequately cool the core.  An additional assumption in the body of the
evaluation was that each loop of shutdown cooling would be operating at its
design flow rate of 11,356 liters per minute [3,000 gpm], for a total flow
rate of 22,712 liters per minute [6,000 gpm].  This flow rate would be
adequate to induce spillover from the core region to the annulus at the
reduced nominal water level.  The temporary procedure change that implemented
the engineering evaluation did not require a minimum flow rate of 22,712
liters per minute [6,000 gpm].  As a result, the operators following the
deficient temporary procedure change placed two shutdown cooling loops in
operation, but did not provide sufficient shutdown cooling water to the
reactor core to maintain reactor vessel temperatures.  In addition, the
temporary procedure change contained no provision or guidance for monitoring
available instruments to ensure that decay heat was being adequately removed
from the reactor core.

As part of the licensee corrective actions, the licensee reviewed the
effectiveness of its safety review process concerning temporary changes and
provided training on the subject of this event to all site personnel who
perform technical and safety reviews.


                                                            IN 93-45
                                                            June 16, 1993
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This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
one of the technical contacts listed below or the appropriate Office of
Nuclear Reactor Regulation (NRR) project manager.

                                             ORIGINAL SIGNED BY

                                    Brian K. Grimes, Director
                                    Division of Operating Reactor Support
                                    Office of Nuclear Reactor Regulation

Technical contacts:  James S. Stewart, RI
                     (215) 337-5240

                     Peter C. Wen, NRR
                     (301) 504-2832
1.  Figure 1, "Oyster Creek Shutdown Cooling System 
      Flow With Open Recirculation Loop"
2.  List of Recently Issued NRC Information Notices  


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