Headquarters Daily report MAY 24, 1994 *************************************************************************** REPORT NEGATIVE NO INPUT ATTACHED INPUT RECEIVED RECEIVED HEADQUARTERS X REGION I X REGION II X REGION III X REGION IV X *************************************************************************** PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION I MAY 24, 1994 Licensee/Facility: Notification: Gpu Nuclear Corp. MR Number: 1-94-0060 Oyster Creek 1 Date: 05/24/94 Forked River,New Jersey SRI/PC Dockets: 50-219 BWR/GE-2 Subject: Technical Specification Required Shutdown Reportable Event Number: N/A Discussion: On May 22, 1994, Oyster Creek initiated a Technical Specification (TS) required shutdown following the failure of a stop-check valve in the control rod drive (CRD) system. There are two CRD pumps, each of which is provided with a Velan 2" stop- check discharge valve. A single suction flowpath from the condensate storage tank branches into the two parallel CRD pumps, then the two lines again form a common line on the discharge of the pumps. Prior to the shutdown, the plant was operating at 100% power with the "B" CRD pump in service. At about 10:00 a.m. on 5/22, the operators started a CRD pump operability surveillance test. The "A" CRD pump was placed in service, and then the "B" pump was secured. At that time, the operators received a CRD low charging water pressure alarm and noted that all CRD flowrates (e.g., charging water, drive water, cooling water) indicated zero flow. The "B" pump was immediately restarted, and CRD system parameters returned to normal. The operators again removed the "B" pump from service about 15 minutes later. The response was the same (i.e., loss of CRD flow). This time, however, the operators did not immediately restart the "B" pump, and that pump subsequently began rotating backwards. The "B" stop-check valve had apparently failed to seat, and the CRD system flow configuration was such that the discharge flow from the "A" pump was returning to the common suction piping via the "B" pump. As a consequence, CRD flow was not being provided to the reactor and hydraulic control units. Operators unsuccessfully attempted to stop the reverse flow situation by manually closing the "B" stop-check discharge valve. Then they closed the "B" suction valve, however, the suction relief valve lifted (set to lift at 250 psig). Subsequently, the "A" CRD pump was stopped for about one minute in order to terminate flow. The operators then initiated two start attempts for the "B" CRD pump; each time the pump failed to start due to a low suction pressure trip. The low suction pressure trip was a result of over-ranging the pressure switch when the "B" suction valve was closed, pressurizing the suction piping. The "A" CRD pump was then restarted. After additional mechanical agitation to the "B" stop-check discharge valve, CRD flows were regained, however, they were below required (Flows reported in 50.72). The plant was removed from service, and cold Shutdown was reached at 6:30 a.m. on 5/23. As of 5/24, the licensee inspected both CRD stop-check valves, and identified that excessive wear appears to have been the cause for the leakage; no foreign materials or damage from possible foreign material was evident. The seating surface for the "A" stop-check valve was lapped successfully. However, the wear to the "B" valve is more severe, and lapping the components may not be successful. The licensee is pursuing plans to either replace the valve (cannot locate a spare valve) or modify the "B" train to provide a similar stop/check arrangement if lapping is not successful. The "B" CRD pump was run to determine whether any pump damage occurred during the event, due to either foreign material or from reverse rotation; no abnormal conditions were identified. The licensee successfully recalibrated the pressure switch. Outage duration was initially expected to be two days. However, the CRD valve repairs could potentially extend the duration by several days. Regional Action: The resident inspectors are onsite reviewing the licensee's activities. Contact: John Rogge (610)337-5146 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III MAY 24, 1994 Licensee/Facility: Notification: Commonwealth Edison Co. MR Number: 3-94-0099 Lasalle 1 2 Date: 05/24/94 Marseilles,Illinois Dockets: 50-373,50-374 BWR/GE-5,BWR/GE-5 Subject: EDO VISIT TO LASALLE STATION Reportable Event Number: N/A Discussion: On May 24, 1994, Mr. James Taylor, Executive Director for Operations; Mr. James Milhoan, Deputy Executive Director for Nuclear Reactor Regulation, Regional Operations and Research; Mr. Frank Miraglia, Deputy Director of Nuclear Reactor Regulation; Mr. John Martin, Regional Administrator Region III; and other NRR and RIII staff will be visiting LaSalle Station. The purpose of the meeting is to discuss various concerns with plant management. In addition, Mr. Taylor will be meeting with Mr. Ken Strahm, Vice President of PWR Operations; and Mr. Steve Perry, Vice President of BWR Operations. Regional Action: Information only. Contact: H.B. CLAYTON (708)829-9602 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III MAY 24, 1994 Licensee/Facility: Notification: Toledo Edison Co. MR Number: 3-94-0100 Davis Besse 1 Date: 05/23/94 Oak Harbor,Ohio RI PC Dockets: 50-346 PWR/B&W-R-LP Subject: PORV INADVERTENT OPENING DURING MAINTENANCE ACTIVITIES Reportable Event Number: N/A Discussion: On 5/19/94 at 5:55 p.m. EDT, with the reactor at 100 percent power, I&C technicians were in process of replacing a reactor trip module (RTM) in RPS channel 1 when the power operated relief valve (PORV) inadvertently opened for approximately 5 seconds. Reactor coolant system pressure decreased about 60 psi before the PORV reclosed. The licensee subsequently determined that during checkout and testing of the newly installed RTM, an electrical spike occurred within the RTM followed by a trip of the DC supply breaker. The spike caused generation of an erroneous high RCS pressure signal to the PORV, initiating its opening. Upon trip of the supply breaker, the high pressure signal was lost and the PORV then reclosed. Additional plant effects were minimal, in part, because preceding the RTM replacement, the plant's integrated control system (ICS) was in "track" due to the rod control and reactor demand stations having been placed in "Hand." The trip of the DC supply breaker caused a loss of non-nuclear instrumentation (NNI) fed through RPS channel 1, which resulted in the loop 1 RCS pressure failing low and causing the energization of the pressurizer heaters. NNI inputs were switched to RPS channel 2 and reenergized. Pressurizer heaters were placed in "Hand" and controlled manually. The replacement RTM was removed to the I&C shop and is undergoing further inspection. Repair of the original RTM was subsequently completed, the RTM returned to service, and RPS channel 1 placed back into operation. Licensee review into the cause of the electrical spike and breaker trip is still ongoing at this time. Regional Action: The resident inspector reviewed the plant's response and continues monitoring of the licensee's ongoing evaluation to determine root cause of the electrical spike. Contact: R.D. LANKBURY (708)829-9631