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Information Notice No. 84-35: BWR Post-scram Drywell Pressurization
INS No.: 6835 84-35 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 April 23, 1984 Information Notice No. 84-35: BWR POST-SCRAM DRYWELL PRESSURIZATION Addressees: All boiling water power reactor facilities holding an operating license (OL) or construction permit (CP). Purpose: This information notice is provided as a notification of events that resulted in drywell pressure increases following a reactor scram and the subsequent unavailability of systems that could be used to reduce drywell pressure. We expect recipients of this notice will review the information for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem occurring at their facilities. However, suggestions contained in this information notice do not constitute NRC requirements and, therefore, no specific action or response is required. Description of Circumstances: Edwin I. Hatch Plant Unit 2 In August 1982, Hatch Nuclear Plant Unit 2 sustained a reactor scram and group 1 isolation from full power conditions when an inboard main steam isolation valve (MSIV) failed closed. After resetting the scram, the opera- tors controlled vessel level and pressure with the high pressure coolant injection (HPCI) system and the safety relief valves (SRVs). On opening the "A" SRV a second time the operators noticed that drywell pressure was in- creasing rapidly. When the drywell pressure reached 2.0 psig a second reactor scram and loss of coolant accident (LOCA) signal were initiated. The scram signal reopened the scram inlet and outlet valves while the LOCA signal initiated the emergency core cooling system and shut off the drywell coolers. Drywell pressure continued to rise and peaked at about 4.0 psig. Some time after the second scram signal was received, it was discovered that the scram discharge volume (SDV) drain line isolation valve had not fully closed. With the sustained reactor scram signal, the reactor water continued to flow out of the open scram valves, through the SDV drain, and into the reactor building equipment drain (RBED) sump where the hot pressurized reactor water flashed to steam. The steam then flowed back through open drains into the reactor building causing a high area temperature trip of the reactor core isolation cooling (RCIC) system and elevated air temperatures in other parts of the reactor building. 8404160142 . IN 84-35 April 23, 1984 Page 2 of 3 The only immediate way of isolating the flow of reactor water through the scram valves was by resetting the reactor scram signal, which would close the scram valves. This could not be readily accomplished because the 2.0 psig drywell pressure scram signal was still present. The operator unsuccessfully attempted to reduce the drywell pressure with the ventilation, system and decided that personnel would have to restore the drywell coolers to operation in order to rapidly reduce the drywell pressure. To accomplish this, an electrical technician had to be sent to the equipment cabinets to bypass the trip signal for the drywell coolers since no control room trip override switch was provided on Unit 2. (Such a switch does exist at Hatch Unit 1). During this event it took nearly one hour and 45 minutes from the time the coolers tripped to the time they could be restored. It took nearly another hour before the coolers reduced the drywell pressure below 2 psig, at which time the scram could finally be reset and the steam release to the reactor building terminated. Quad Cities Station Unit 2 In June 1982, the Unit 2 reactor scrammed from approximately 95 percent power. The reactor scrammed from a low water level signal following the trip of a reactor feed pump due to the loss of one of two major 4 kV buses. Prior to the trip of the turbine generator, and the subsequent loss of the other major 4 kV bus, the operator ran the vessel level up to the high level trip point of the feedwater pumps. The operator then controlled reactor pressure by intermittently opening main steam relief valves. Approximately 30 minutes after the scram the drywell pressure exceeded 2 psig and a second scram sig- nal was received along with an emergency core cooling system (ECCS) initia- tion signal and the trip of the drywell coolers and the reactor building closed cooling water (RBCCW) pumps. Both 4 kV buses were returned to service 40 minutes after the trip and sup- pression pool cooling was established 50 minutes after the trip. The trip signal to the drywell coolers and RBCCW pumps was bypassed about l hour and 35 minutes after the trip and the drywell pressure was reduced to below 2 psig. Once the drywell pressure was reduced with the drywell coolers, the second scram signal could be reset. Discussion: During the Hatch event it is believed that the high drywell pressure was caused by the "A" SRV tailpipe vacuum breaker sticking in the open or par- tially open position (see Information Notice No. 83-26, "Failure of Safe- ty/Relief Valve Discharge Line Vacuum Breakers"). The high drywell pressure experienced during the Quad Cities event was apparently caused by leaky gas- kets on the flanged elbows in the relief valve discharge lines. Such bypasses or leakage pathways in relief valve piping may reduce the effectiveness of the pressure suppression system and are of themselves significant events. This Information Notice is focussed on the subsequent difficulties in making equipment available to reduce drywell pressure to normal following leakage . IN 84-35 April 23, 1984 Page 3 of 3 from SRV piping. During both events, systems normally used to reduce drywell pressure were tripped by a high drywell pressure condition and could not be readily reset. Considerable time was spent before drywell pressure was re- duced because electrical jumpers had to be installed in order to bypass the trip signals. At Hatch Unit 2, no design changes were made to the drywell chiller trip logic to retain drywell cooling in the presence of a LOCA signal. Instead, the licensee's corrective action included training of site personnel on bypassing signals in general along with operator training on the functions of systems that would allow signals to be bypassed, if needed, on an emergency basis. At Quad Cities Unit 2 the high drywell pressure trip logic has since been modified to allow for continuous operation of the drywell coolers and the RBCCW pumps when normal power is available to the emergency buses. A more detailed description of the event and corrective actions taken after the event at Hatch Unit 2 is discussed in Power Reactor Events, Volume 5, No. 4, Nuclear Regulatory Commission, January, 1984. Because of the potential seriousness of this type of event, licensees may wish to consider design changes to the high drywell pressure logic to prevent tripping the drywell coolers when offsite power is available, and/or provide convenient override arrangements (e.g., switches) to permit rapid restarting of drywell coolers when a high drywell pressure condition still exists. Before making such design changes, appropriate analyses should be performed in consultation with NRC. No written response to this notice is required. If you have any questions regarding this matter, please contact the Regional Administrator of the appropriate NRC Regional Office or this office. Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical Contact: Eric W. Weiss, IE (301) 492-4973 Attachment: List of Recently Issued IE Information Notices
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