Headquarters Daily report JULY 06, 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 JULY 6, 1994 Licensee/Facility: Notification: Public Service Electric & Gas Co. MR Number: 1-94-0075 Salem 2 Date: 07/05/94 Hancocks Bridge,New Jersey SRI PC Dockets: 50-311 PWR/W-4-LP Subject: UNISOLABLE FLANGE LEAK Reportable Event Number: 27478 Discussion: At 1:45 p.m. on July 2, Salem Unit 2 experienced a 20 gallon per minute (gpm), unisolable leak from a flange on a 1 inch pipe connected to No. 22 reactor coolant pump seal package. The pipe and flange assembly were previously used to measure differential pressure across the pump seal to support the original start-up testing of the unit. The assembly has been abandoned in-place since that time. At the time the leak occurred, Unit 2 was in Mode 3, at 2235 psig and 531 degrees F. The licensee was attempting to mitigate a potential leak through the flange, as evidenced by boric acid crystals noted in the area during the outage in December 1993. Leak Repair Incorporated technicians had completed installation (under the supervision of the licensee) of a flange clamp and initiated injection of sealant material when the leak occured. The injection of material had progressed to the point of filling the void within the flange volume, with an increase in pressure to about 50 psig at the time the leak increased. All personnel evacuated the containment immediately. No personnel injuries occured. Minor personnel contamination due to noble gas was noted on some individuals. The shift entered TS 3.4.7.2.b, for unidentified leakage > 1 gpm, and initiated actions to place the plant in Mode 5. The plant reached Mode 5 on July 3. With reactor pressure at about 15 psig, the leak had slowed to about 1 gpm. The licensee determined that there was no damage attributed to the leak and initiated action to repair the leaking assembly. On July 4, leak repair technicians re-adjusted the flange clamp and noted that 3 of the 4 stud bolts on the flange were loose. The bolts were replaced and re-torqued and the leak was stopped. A freeze seal was successfully applied and the flange assembly was disassembled and replaced with a blank flange under the licensee's modification process. Region I inspectors were on-site during the evolution to review and witness the repair activities. The licensee completed the work activity at about 10:00 p.m. on July 4, 1994. Regional Action: Residents and regional inspectors are continuing to followup on the licensee's activity and cause analysis for the leak. Contact: John White (610)337-5114 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION I JULY 6, 1994 Licensee/Facility: Notification: Gpu Nuclear Corp. MR Number: 1-94-0076 Oyster Creek 1 Date: 07/06/94 Forked River,New Jersey SRI PC Dockets: 50-219 BWR/GE-2 Subject: Technical Specification Required Shutdown and Unusual Event Reportable Event Number: 27474 Discussion: During performance of monthly surveillance testing of the containment spray/emergency service water (CS/ESW) system No. 2, plant operators noted a rapid increase in heat exchanger differential pressure on the ESW side. Pressure had increased to approximately 75 psid in about five minutes(normally 10 psid). The surveillance was started on July 5, 1994, at 10:25 a.m. and secured at 10:45 a.m. CS/ESW system No. 2 was declared inoperable at 10:50 a.m. CS/ESW system No. 1 was started at 11:50 a.m. to verify operability of the redundant system. CS/ESW system No. 1 was secured at 11:52 when ESW differential pressure reached 55 psid. CS/ESW system No. 1 was declared inoperable at 11:53 a.m. Both CS/ESW systems maintained approximately 2300 to 2400 gpm ESW flow during testing, normal flow is 3400 to 3600 gpm. With both CS/ESW systems inoperable the licensee is required to be in cold shutdown within the next 30 hours. An unusual event was declared at 12:05 p.m. because both CS/ESW systems were inoperable, requiring a plant shutdown, and would not be returned to service within the next eight hours. The licensee commenced a plant shutdown at 11:53 a.m. on July 5, 1994, at 30 megawatts electric per hour. Concurrent with the shutdown the licensee opened the CS/ESW system 2 heat exchangers and drained them to allow cleaning. The plugging was caused by a layer mussels and silt approximately 1 inch thick on the inlet tube sheet. There was also partial blockage of the third pass tube sheet. The heat exchanger was cleaned and temporarily closed for additional flushing at 9:22 p.m. on July 5, 1994. Results of the flushing indicated that differential pressure was within limits and further cleaning was not required. The heat exchangers were fully closed and the CS/ESW system No. 2 successfully passed the surveillance test and was declared operable at 2:25 a.m. on July 6, 1994, and the unusual event was terminated. The plant will be returned to full power on July 6, 1994. Regional Action: Routine Resident Followup, Regional Review of Design Contact: John Rogge (610)337-5146 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III JULY 6, 1994 Licensee/Facility: Notification: Commonwealth Edison Co. MR Number: 3-94-0122 La Salle 1 Date: 07/06/94 Marseilles,Illinois SRI VIA LAN Dockets: 50-373 BWR/GE-5 Subject: UNIT 1 REACTOR SCRAM Reportable Event Number: N/A Discussion: On July 5, 1994, a circuit card failed in the Unit 1 Bailey reactor water level control system, causing a reduced flow demand to the Lovejoy feedwater turbine control system. Turbine driven reactor feedpump (TDRFP) "B" was in three element automatic control and the motor driven reactor feedpump was in manual control. TDRFP "A" was off and unavailable as power ascension testing for the Lovejoy modification (completed in the recently concluded refueling outage) had not yet been conducted for that pump. The failure caused a drop in reactor water level which the reactor operator tried to mitigate by taking the TDRFP "B" controller to manual and raising flow demand. The location of the failure rendered this action ineffective and the unit scrammed on low reactor water level. A replacement circuit card was not available and the corresponding card for TDRFP "A" was utilized. The licensee considered operating TDRFP "B" preferable as it had undergone the majority of its modification testing. During the scram, a fuse in the back of the full core display blew causing one row of scram lights and local power range monitor downscale lights not to work. This was the first of these particular fuses known to have blown at LaSalle. The fuse was replaced. In addition, the reactor manual control system (RMCS) tripped. A branch amplifier card was found to be malfunctioning and was replaced. This did not appear to be related to the scram as sporadic RMCS trips had been occurring for a couple of days. This failure did not affect the scram function. An electro-hydraulic control pump unexpectedly tripped and the standby pump automatically started. This pump is powered from a bus that fast transfers from the unit auxiliary transformer to the system auxiliary transformer. The licensee suspected it may have been dropped during the transfer and was conducting additional testing of the breaker and motor. Following the repairs and replacements noted above, the unit was restarted. Regional Action: The SRI responded to the site and the resident inspectors will continue to follow the licensee's investigation. Contact: H.B. CLAYTON (708)829-9602