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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 .
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