ML20054A036
| ML20054A036 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Davis Besse |
| Issue date: | 03/23/1981 |
| From: | ORGANIZATION FOR ECONOMIC COOPERATION & DEVELOPMENT |
| To: | |
| Shared Package | |
| ML19240B432 | List: |
| References | |
| FOIA-81-380 NUDOCS 8204150233 | |
| Download: ML20054A036 (6) | |
Text
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i Loss of Decay Heat Removal Caoability Date and Place - On April 19, 1980, Toledo Edison Company reported the tempo-rary loss of oecay heat removai capability at Davis-Besse Unit I while the plant was in a refueling outage. The Davis-Besse Unit I nuclear plant utilizes a pressurized water reactor, designec by Babcock & Wilcox, and is located in Ottawa County, Ohio. Nature and Probable Consecuences - On April 8, 1980, the plant was taken to cold shutoown for refueling, maintenance, and modifications. On April 19, 1980, the plant experienced a loss of two of four.120 VAC essential instrument buses resulting in a loss of decay heat removal (DHR) capability for about two and one-half hours. At the time of the event, reactor coolant system (RCS) temperature was 90*F, decay heat was being removed by Decay Heat Loop No. 2, the vessel head was l detensioned with bolts in place, the reactor coolant level was slightly below the vessel head flanges, and the manway covers on top of the once-through i steam generators (OTSG) were removed. l Since the plant was in a refueling mode, many systems or components were out l of service for maintenance or testing purposes. In addition, other systems and components were deactivated to preclude their inadvertent activation while in a' refueling mode. Systems and components that were not in service or l deactivated included: Containment Spray System; High Pressure Injection l System; Source Range Channel 2; Decay Heat Loop No.1; Station Battery IP and IN; Emergency Diesel-Generator No. 1; 4.16 KV Essential Switchgear Bus C1; and l l 13.8 KV Switchgear Bus A (this bus was energized but not aligned). l The event was due to the tripping of a non-safeguards feeder breaker in 13.8 KV Switchgear Bus B (probably due to mechanical vibration or bumping of the i
2 3 breaker by construction workers who were working in the area). Because of the extensive maintenance and testing activities being conducted at the time, Channels 1 and 3 of the Reactor Protection System (RPS) and Safety Features Actuation System (SFAS) were being energized from only one electrical source, the source emanating from the tripped breaker. Since the SFAS logic used at Davis-Besse is a two-out of-four input scheme in which the loss (or actuation) of any two input signals results in the actuation of all four output channels (i.e., Channels 1 and 3, and Channels 2 and 4), the loss of power to Channels 1 and 3 bistables also resulted in actuation of SFAS Channels 2 and 4. The actuaticn of SFAS Channels 2 and 4, in turn, affected Decay Heat Loop.No. 2, the operating loop. Since the initiating event was a loss of power, all five levels of SFAS and the SFAS interlock channel on the Decay Heat Isolation Valve DH-12 actuated on the loss of 120 VAC power to Channels 1 and 3 (i.e., Level 1 - High Radiation; Level 2 - High Pressure Injection; Level 3 - Low Pressure Injection; Level 4 - Containment Spray; and Level 5 - ECCS Recirculation Mode). Loss of power in the interlock circuit resulted in loss of decay heat pump suction from RCS hotleg No. 2 when containment isolation valve DH-12 closed. Actuation of SFAS Level 3 aligned the Decay Heat Pump No. 2 suction 'to the Borated Water Storage Tank (BWST) in the low pressure injection mode. Actuation of SFAS Level 5 represents a low level in the BWST; therefore, upon its actuation, ECCS opera-tion was automatically transferred from the Injection Mode to the Recirculation Mode. Transfer to the Recirculation Mode closed the supply valve from the,BWST and l opened the valve to the dry Containment Emergency Sump. During the opening and closing of these two valves (60 to 90 seconds), approximately 3,500 gallons i of water from che BWST were injected into the RCS via the Decay Heat Pump No. 2 and approximately 1,500 gallons backflowed into the Emergency Sump by I gravity. The operator manually stopped Decay Heat Pump No. 2 approximately 2 minutes into the event to stop the injection of water into the RCS to' pre. vent pump damage due to loss of suction. .The above sequence of even'ts resulted in the loss of decay heat removal capability for approximately two and one-half hours, the time to properly check out and realign the electrical systems and realign and vent air from l No. 2 Decay Heat Loop. Decay Heat Loop No. I was drained at the time in preparation for maintenance and not available as an alternate decay heat i l removal system. l During the time of the event, the reactor coolant temperature increased from 90 F to about 170 F (the Technical Specification definition for refueling ' mode is an average temperature of <140 F); however, the final temperature reachied I was still considerably below that which could adversely affect the heat transfer characteristics of the fuel such that fuel damage could result. There were no offsite releases of radioactivity, and there were no overexposures or injuries to personnel associated with the event.
3 The loss of power also caused communication problems in addition to the loss of decay heat removal capability. The Gaitronics System (internal communica-tions system) was without power for about 33 minutes. This complicated com-munications between the control room personnel and personnel in other parts of the plant, which may have contributed to the delay in restoring decay heat removal capability. There have been other incidents involving the decay heat removal systems at Davis-Besse Unit I during the refueling outage which began on April 8, 1980. On April 18, 1980, while the reactor was still in cold shutdown prior to entering the refueling mode - an operator discovered that the water level in the RCS had dropped significantly. The loss of reactor coolant inventory was from an open valve which had been opened to facilitate draining of the out-of-service Decay Heat Loop 1 and from a partially open valve (due to the valve's remote operator being out of adjustment) in the operable Decay Heat Loop 2. A few minutes into the event, the running decay heat pump was tripped because of concern for possible loss of suction; the pump was restarted about 29 minutes later. This was a violation of the Tec'hnical Specifications which require that while in Mode 5 (cold shutdown), at least one reactor coolant loop must be in operation with an associated reactor coolant pump or a decay heat pump operating. On May 28, 1980, Decay Heat Isolation valve DH-11 was inadvertently tripped closed resulting in loss of decay heat removal for about two minutes until the valve was reopened and the pump restarted. An Instrumentatipn and Control (I&C) mechanic was preparing to test an NRC required modification using the same pressure instrument used to activate DH-11 interlock circuit. Due to a test procedural inadequacy, the valve interlock circuit actuated when a test connection caused the pressure input to spike high. On May 31, 1980, the control room operator stopped the decay heat pump when the flow meter indication for the decay heat loop dropped off scale The pump was returned to service when it was discovered that an I&C mechanic had taken the flow meter out of service for surveillance testing without informing the control room. On June 14, 1980, an inadvertent SFAS Levels 1, 2, and 3 actuation resulted in a loss of decay heat removal tiow for about two minutes. An I&C mechanic was I in the process of restoring containment pressure inputs to SFAS following an Integrated Leak Rate Test. Due to a procedural inadequacy, the SFAS actuated aligning the operating decay heat pump to the BWST and injecting water into the RCS and refueling canal. The BWST level dropped to the low level limit, l actuating SFAS Level 5, closing the BWST isolation valves. This caused a loss of suction to the decay heat pump. On July 10,1980, at 1050 hours, the decay heat removal flow was inadvertently reduced to about 2000 gpm for approximately 51 seconds on restoring power to the flow control valve. Due to a procedural error, power was lost to the flow control valve when SFAS Channel 2 was de energized for maintenance work on the 120-VAC essential Bus Y-2. When power was restored to the control valve, the
4 s valve controller caused the decay heat flow to momentarily de' crease below the minimum required flow of 2800 gpm. On July 24, 1980, at 0935 hours, a blown fuse caused the Decay' Heat Isolation valve DH-12 to close resulting in loss of decay heat removal for about 50 minutes until manual bypass valves were open. The blown fuse was caused by an electrician pulling wires (in the cabinet containing the isolation valve control wires) for an NRC-required design change. Also on July 24, 1980, at 2232 hours, Decay Heat Isolation valve DH-11 was inadvertently tripped closed resulting in loss of decay heat removal for about two minutes until the valve was reopened and the pump restarted. Inadequate job planning for restoration of the system from a modification led an I&C mechanic to perform steps out of sequence. On August 8,1980, Decay Heat Isolation valve DH-11 was inadvertently tripped closed resulting in loss of decay heat removal for about three minutes until the valve was reopened and the pump restarted. A bistable in the valve circuit was removed during maintenance causing the valve to trip. In the planned maintenance, the function of the bistable had been improperly designated. On August 13, 1980, Decay Heat Isolation valve DH-li was inadvertently tripped closed resulting in loss of decay heat removal for about five minutes until the valve was reopened and the pump restarted. An I&C mechanic failed to fully defeat the automatic trip on DH-11 prior to performing modification work on an SFAS channel. Cause or Causes - Two factors identified as major contributors to the events are (1) extensive maintenance activities which led to a loss of the DHR capability, and (2) inadequate procedures and/or administrative controls which, if corrected, could have precluoed the events or at least ameliorated their effects. For example, for the rather extended loss of DHR capability on April 19, 1980, the high' pressure injection and containment spray pumps had been deactivated to preclude their inadvertent actuation during refueling. If SFAS Level 5 had also been bypassed or deactivated, or if the emergency sump isolation valves had been closed and their breakers opened, this event would have resulted in a minor interruption of decay heat flow. It is also believed that the event could have been avoided, or ameliorated, if the maintenance activities had been less extensive during the outage, or at least better coordinated. If l activities had been restricted so that two SFAS channels would not be lost by a single event (e.g., serving Channels 1 and 3 from separate sources), the loss of DHR capability would not have occurred. In addition, if a backup DHR system would have been readily availa'ble, consequences of the loss of the operating DHR loop would have been lessened. actions Taken to Prevent Recurrence i Licensee - As a result of the April 19, 1980 event, to increase the reliability of oecay heat removal during the refueling outage, the licensee (1) closed and electrically disabled the isolation valves to the containment emergency sump, 1
1 5 (2) kept second decay heat loop in standby until the refueling canal was filled, and (3) reviewed future electrical distribution system maintenance, modification, and testing to provide maximum diversity to the 120-VAC instru-ment power buses. Appropriate operating procedures were modified. Corrective actions to the internal communication system problem will.be included in the licensee's response to NRC's Inspection and Enforcement Circular No. 80-09 (Problems with Plant Internal Communications Systems) recommendations issued to all holders of a power reactor operating license or construction permit on April 28, 1980. To date, the licensee has made several changes including the placement of the Red Phone on uninterruptible power, installing internal antennas in the auxiliary building for better radio communication and other telephone changes. However, the power s,ource for the Gaitronics System has yet to be modified. Long-term corrective actions were taken by the licensee in accordance with the NRC's Inspection and Enforcement Bulletin No. 80-12: 1. Additional revisions to EP 1202'.32, Loss of DHR Frequency Procedure, to include alternate methods to those previously listed to supply water to the reactor core and reference to appropriate procedures for monitoring core temperatures using the incore thermocouples. 2. Additional guidance was provided for venting the DHR if air is drawn into the system. 3. Five procedures were revised to insure power is removed for Emergency Sump Isolation valves DH-9A and 98 in Modes 5 and 6. 4. Instrument AC System Procedure SP 1107.09 was revised to allow the 120-VAC instrument power inverter to be supplied from the DC Bus when normal AC feed is rot evailable. This will minimize the possible loss of power to two instrument channels at one time. 5. A special procedure was written to require, whenever possible, that. redundant decay heat' system not be intentionally removed from service in Modes 4, 5, or 6 unless at least one steam generator is available for decay heat removal, the refueling canal is filled, or the decay heat pump can be restored to service or a gravity flow path to the RCS can be established within four hours. The special order,also covers expediting the restoration of redundant or diverse methods if component failure causes loss of alternate decay heat removalimethods. ~~ ~
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