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HomeNRC LibraryDocument CollectionsGeneric CommunicationsInformation Notices1993

# Information Notice No. 93-93: Inadequate Control of Reactor Coolant System Conditions during Shutdown

```
                                 UNITED STATES
                         NUCLEAR REGULATORY COMMISSION
                     OFFICE OF NUCLEAR REACTOR REGULATION
                            WASHINGTON, D.C. 20555

                               December 8, 1993

NRC INFORMATION NOTICE 93-93:  INADEQUATE CONTROL OF REACTOR COOLANT SYSTEM
                               CONDITIONS DURING SHUTDOWN 

Addressees

All holders of operating licenses or construction permits for nuclear power
reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information
notice to alert addressees of recent plant events that involved inadequate
control of the reactor coolant system (RCS) conditions while the plant was
shut down.  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.

Background

During refueling or periods of cold shutdown, technical specifications allow
licensees to remove from service equipment required to be operable during
power operations.  At these times, surveillance tests and maintenance
activities may result in unusual plant configurations not encountered during
power operations.  The events described below demonstrate the importance of
controlling plant activities during shutdown operations to ensure continued
operability of plant equipment, RCS inventory control, and adequate shutdown
cooling. 

Description of Circumstances

Opening of the Residual Heat Removal Pump Suction Relief Valve

On November 21, 1992, the Joseph M. Farley Nuclear Plant had been in cold
shutdown for 27 days.  The "B" and "C" reactor coolant pumps were running, and
the operators were monitoring RCS pressure using the "C" loop wide range
pressure transmitter, which indicated 2,861 kPa [415 psia].  At 6:16 a.m., the
"C" reactor coolant pump was secured for maintenance.  Normal thermal and
hydraulic characteristics caused the system pressure at the loop "C" hot leg
to increase.  A drag pointer recorder indicated that RCS pressure had reached
3,516 kPa [510 psia].  The increased pressure caused the "A" train residual
heat removal relief valve on loop "C" to lift, relieving 6,435 liters 
[1,700 gallons] of RCS inventory to the pressurizer relief tank and causing 

9312030341.

                                                            IN 93-93
                                                            December 8, 1993
                                                            Page 2 of 3

the level in the pressurizer to drop to zero.  The relief valve reseated
approximately 4 minutes later and pressurizer level indication was regained
after another 2 minutes using normal makeup from the refueling water storage
tank.

The Alabama Power Company (the licensee) investigated and found that the
output of pressure transmitter "C" was consistently 172 kPa [25 psi] lower
than the output of pressure transmitter "A", which had been monitored in the
past for this configuration.  This discrepancy indicates that the system
pressure before the event was approximately 3,034 kPa [440 psia] and rose to
approximately 3,206 kPa [465 psia] when the reactor coolant pump "C" was
secured.  The drag pointer recorder was found to be out of calibration by 
338 kPa [49 psi] indicating that the actual pressure attained was 3,178 kPa
[461 psia].  The setpoint for the relief valve that lifted was 3,172 kPa
[460 psia] so the actual liftpoint was within tolerance for the relief valves. 
The licensee later imposed a more restrictive band on allowable system
pressure for this configuration.

Reactor Coolant Spill in Containment

On September 28, 1992, Palo Verde Nuclear Generating Station, Unit 3, had been
in Mode 6 (Refueling) for three days.  Personnel testing the safety injection  
tank 2B outlet motor-operated valve inadvertently released nitrogen from the 
tank to the RCS, causing a large wave in the refueling cavity.  The wave
splashed 5,678 liters [1,500 gallons] of water over the sides of the refueling
cavity, contaminating large areas of the containment building.  The cause of
this event was inadequate venting of the safety injection tank before testing
the outlet valve.

Loss of Shutdown Cooling 

On January 25, 1993, GPU Nuclear Corporation (the licensee for the Oyster
Creek Nuclear Generating Station) experienced a degradation of shutdown
cooling due to a failure to incorporate shutdown cooling flow in accordance
with a licensee engineering evaluation for the current plant recirculation
loop configuration.  This event is discussed fully in NRC Information Notice
93-45, "Degradation of shutdown Cooling System Performance," June 16, 1993.

Discussion

These events were caused by deficiencies in human performance that include:
inadequate development and review of procedures, inadequate monitoring and
trending of plant parameters, and inadequate work practices.  These types of
events are of concern because they indicate that, with the reactor in a
shutdown condition, personnel may have a decreased awareness of the safety
consequences of their actions and may not realize the importance of careful
control of plant activities to ensure continued operability of plant
equipment, RCS inventory control, and adequate shutdown cooling.  Because
plant systems are not in usual operating configurations while the plant is in
a refueling or cold shutdown condition, proper human performance is important .

                                                            IN 93-93
                                                            December 8, 1993
                                                            Page 3 of 3

in maintaining safety and controlling plant parameters such as RCS temperature
and configuration.

This information notice requires no specific action or written response.  If
you have any questions about the information in this notice, please contact
the person listed below or the appropriate Office of Nuclear Reactor
Regulation (NRR) project managers.

                                    /s/'d by BKGrimes

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

Technical contact:  Eric J. Benner, NRR
                    (301) 504-1171

Attachment:  
List of Recently Issued NRC Information Notices
.
```

*Page Last Reviewed/Updated Tuesday, November 12, 2013*

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