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93

Information Notice No. 93­93: 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

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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

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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

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