Headquarters Daily report APRIL 20, 1995 *************************************************************************** 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 - HEADQUARTERS APRIL 20, 1995 MR Number: H-95-0092 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: CONTROL ROOM AIR CONDITIONING INOPERABLE FOLLOWING A LARGE BREAK ACCIDENT The NRR/AEOD Events Assessment Panel on April 18, 1995, classified the Fort Calhoun discovery that a large break accident could disable the control room air conditioning and cause certain safety equipment to become overheated, as a Significant Event for the Performance Indicator Program. At Fort Calhoun, the component cooling water (CCW) system transfers heat to the raw water (RW) system from the containment coolers as well as from the air conditioning units in the control room and from other components. However, the capability of the containment coolers to remove the containment heat produced by large-break accidents inside the containment greatly exceeds the capability of the CCW system to reject the heat to the RW system at the normally low CCW temperatures needed to support the control room air conditioners. Therefore, a large-break accident will cause the CCW temperature to rise rapidly. The licensee calculated that, under design-basis accident conditions, the CCW temperature could reach 106F within 3 minutes and the maximum temperature could reach 186F. The control room air conditioning units would automatically shut down at a CCW temperature of 106F. In addition, the air conditioning unit condensers are equipped with rupture discs that could blow out at a CCW supply temperature as low as 130F releasing the freon refrigerant into the control room. Without air conditioning control room temperature could increase to levels that could hinder operator activities and cause design temperatures of safety-related equipment in the control cabinets to be exceeded. In addition, the CCW system pipe hangers are only designed to withstand temperatures of up to 160F. During the recently completed 1995 refueling outage, the licensee provided air-cooled freon condensers for the control room air conditioners which are independent of the CCW system. The licensee's analysis indicates that all four CCW to RW heat exchangers are necessary to ensure that the CCW temperature will not exceed the 160F limit for the piping supports during an accident, when the raw water temperature equals or exceeds 70F. A modification provides automatic actuation of the RW inlet and outlet for all four heat exchangers under these conditions. The loss of operability of any of the four heat exchangers requires entry into a limiting condition of operation. The licensee has submitted a technical specification amendment request to incorporate the appropriate new limiting conditions for operation in the Fort Calhoun Technical Specifications. An information notice discussing this event has been written and is in concurrence. CONTACT: Donald C. Kirkpatrick, NRR/DOPS/OECB (301) 415-1849 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION APRIL 20, 1995 MR Number: H-95-0093 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: POTENTIAL FOR UNFILTERED RELEASE FOLLOWING LOSS OF COOLANT ACCIDENT The NRR/AEOD Events Assessment Panel on April 18, 1995, classified an unfiltered release path in the enclosure building filtration system (EBFS) at Millstone, Unit 2, as a Significant Event for the NRC Performance Indicator Program based on programmatic weakness. On December 6, 1994, with the plant defueled, the licensee identified a release path from the Enclosure Building that would allow a direct discharge to atmosphere bypassing the charcoal filters following a Loss of Coolant Accident (LOCA). Specifically, a hydrogen analyzer cabinet and sample hood exhaust fan were found to take a suction on the enclosure building and discharge approximately 1000 cfm out the Unit 2 Main Exhaust stack. This flow path has HEPA filters but lacks charcoal filters. The purpose of the non- safety related exhaust fan is to maintain a negative pressure on the sample hood to prevent technicians from being exposed to gas while obtaining routine chemistry samples; however, the fan has no automatic shut off feature and there are no isolation dampers in the line to prevent a release during an event that would actuate the EBFS. The licensee's radiological assessment staff performed an evaluation to determine the effects of this condition. Their analysis used the design basis accident assuming substantial core damage with subsequent release of appreciable quantities of fission products as identified in 10 CFR 100.11. The licensee calculated site boundary thyroid dose to be 670 rem, which is significantly greater then 10 CFR 100.11 limit of 300 rem. This configuration has existed since initial plant construction and startup. On February 9, 1995, after further investigation, the licensee identified another potential design deficiency in the enclosure building purge system such that, coincident with a single failure, a release path from the Enclosure Building would allow an unfiltered discharge to atmosphere following a LOCA if Enclosure Building purging operations were being performed. The licensee's radiological assessment staff concluded that the calculated site boundary thyroid dose would also exceed 10 CFR 100.11 limits if the release went undetected. The licensee has included these design deficiencies as restart issues. The NRC has issued enforcement discretion, EA 95-004, on February 16, 1995, relating to NRC Inspection Report No. 50-336/94-34. The Office of Enforcement categorized the violation as Severity Level III with enforcement discretion granted. CONTACT: Stephen S. Koenick, NRR/DOPS/OECB (301) 415-2841 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION APRIL 20, 1995 MR Number: H-95-0094 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: POTENTIAL PRESSURE LOCKING OF CONTAINMENT SUMP RECIRCULATION VALVES The NRR/AEOD Events Assessment Panel on April 18, 1995, classified the potential pressure locking of both containment sump recirculation valves at Millstone, Unit 2, as a Significant Event for the Performance Indicator Program. The significant event classification was based on the risk significance of the event. On January 26, 1995, Millstone Unit 2 determined that both containment sump recirculation gate valves may experience pressure locking during a design-basis loss-of-coolant accident (LOCA) and fail in the closed position. The failure of these valves would make the water source for the emergency core cooling system (ECCS) and containment spray unavailable during the recirculation phase of the LOCA. This condition was discovered by the licensee after re-evaluation of all Unit 2 valves to address weaknesses in their previous evaluation criteria identified by the NRC Motor-Operated Valve Inspection at Millstone Unit 1 in March 1994. Millstone Unit 2 has two containment sump recirculation valves (one in each of two parallel paths). The valves are normally-closed parallel-wedge gate valves, with the containment sump side of the valve dry and exposed to the containment. Water from the refuelling water storage tank may leak past the pumpside gate valve disc and fill the valve bonnet with water. During a LOCA, the sump side of each valve would be exposed to reactor coolant with temperatures as high as 289 degrees F for some time until the valve is required to open. Heating of the valve would heat water trapped in the valve bonnet. A pressure increase of 33 psi per degree F is predicted. The licensee has determined that a bonnet pressure of 150 psig (requiring only about a 5 degrees F temperature rise) could prevent the valve from opening. Information Notice 95-14, "Susceptibility of Containment Sump Recirculation Valves to Pressure Locking," was issued on February 28, 1995, to alert licensees to this issue. In addition, Temporary Instruction 2515/129, "Pressure Locking of PWR Containment Sump Recirculation Gate Valves," was issued on March 17, 1995, to assess the short term actions by licensees with containment sump recirculation valves potentially susceptible to pressure locking. Information from this issue will also be accounted for in a proposed generic letter on pressure locking and thermal binding. The Probabilistic Safety Assessment Branch performed a risk assessment of this event. Assuming that the recirculation function is not recoverable and that it is essential for mitigation of medium and large LOCAs, it is estimated that the inoperability of the containment sump recirculation valves has the potential to increase the core damage frequency to the 10-3/year range. If recovery is possible, the estimate would be reduced. Opportunities for recovery might include manually opening a containment sump valve, or obtaining additional borated water from Millstone Unit 3. However, in the absence of procedures for these actions, they are not likely to have enough probability of success to significantly change the estimate. In addition, the containment temperature and pressure would increase due to the loss of containment sprays which could result in a challenge to containment integrity and potential containment failure. The loss of containment spray will also reduce the fission product scrubbing inside containment and would increase release to the environment through containment leakage, even if containment integrity is maintained. CONTACT: Eric J. Benner, NRR/DOPS/OECB (301) 415-1171 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION APRIL 20, 1995 MR Number: H-95-0095 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: LACK OF AN ADEQUATE PREVENTIVE MAINTENANCE PROGRAM FOR CERTAIN SOLID STATE PROTECTION SYSTEM COMPONENTS The NRR/AEOD Events Assessment Panel meeting on April 18, 1995, classified the Salem licensee's lack of an adequate preventive maintenance (PM) program for the solid state protection system (SSPS) logic matrix power supplies as a significant event for the performance indicator program. The panel considers the absence of such a program a significant programmatic weakness. The Salem and Diablo Canyon licensees reported a condition that could result in the failure of one or both trains of the SSPS during a seismic event or a main steamline break in the turbine building. Information notice 95-10 was issued by the NRC to inform other licensees of the findings of the Diablo Canyon and Salem licensees. Both licensee's attempted to implement a design change that they felt would eliminate the design vulnerability. The Diablo Canyon licensee was successful in making the design change. The Salem licensee encountered numerous problems in attempting the design change. As part of the Salem licensee's design change procedure, the front panel breaker on one of the 15V dc power supplies SSPS logic matrix power supplies was manually tripped. There are two 15V dc power supplies in each train. The power supplies are operated in parallel, separated by diodes, and each power supply is sized for 100 percent of the load requirements. When the circuit breaker for one of the 15V dc supplies was manually tripped, the circuit breaker for the second 15V dc supply tripped open. This was not the expected response. As a result of both supplies being tripped, all 15V dc loads were de-energized. The unexpected opening of the 15-V output breaker prompted an investigation by the licensee. Several power supply anomalies were discovered during the investigation. All of the anomalies could be directly attributable to the licensee's lack of an adequate maintenance program for the class 1E power supplies. Several potential causes were identified for the power supply anomalies. Voltage regulators in several power supplies were degraded. Installed capacitors were degraded, possibly due to aging. Dirt and metal filings were found on some power supplies. A wire from the rear of a 15V power supply was found to have shorted against a transistor metal heat sink. Several power supplies were more than 15 years old. The as-found condition of the SSPS power supplies revealed what the staff considers to be significant programmatic weaknesses in the licensees maintenance of the class 1E system. Although considerable redundancy exists in the SSPS, this does not relieve the licensee from the responsibility of having an adequate PM program to ensure operability of all safety related systems, structures and components. CONTACT: E.N. Fields, NRR/DOPS/OECB (301) 415-1173 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III APRIL 20, 1995 Licensee/Facility: Notification: Wisconsin Electric Power Co. MR Number: 3-95-0069 Point Beach 1 Date: 04/19/95 Two Rivers,Wisconsin RI VIA PC Dockets: 50-266 PWR/W-2-LP Subject: UNIT 1 STARTUP FOLLOWING REFUELING SHUTDOWN Reportable Event Number: N/A Discussion: Unit 1 was synchronized to the grid at 05:50 (CDT) on April 17 following refueling outage 22. This 37-day outage included installation of a new emergency diesel generator (EDG) and removal from service of an existing EDG. The station now has three operable EDGs. One EDG is dedicated "A" train for both units and the other two are dedicated one per unit for "B" train with manual cross tie capability. Eddy current inspection on both steam generators during the outage resulted in plugging one tube in Steam Generator "B" and none in Steam Generator "A." Also feedwater J-tubes in each Steam Generator were inspected with no deficiencies noted. Additional major activities completed during the outage included molded case circuit breaker replacement, boric acid concentration reduction in the boric acid storage tanks, refurbishment of reactor coolant pump "B" seal assembly, replacement of the "A" residual heat removal (RHR) pump rotating assembly, and high pressure turbine refurbishment. Regional Action: Routine resident inspector followup. Contact: M. J. FARBER (708)829-9605 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III APRIL 20, 1995 Licensee/Facility: Notification: Iowa Electric Light & Power Co. MR Number: 3-95-0070 Duane Arnold 1 Date: 04/20/95 Palo,Iowa SRI VIA LAN Dockets: 50-331 BWR/GE-4 Subject: STARTUP AFTER REFUELING OUTAGE (RFO) 13 Reportable Event Number: N/A Discussion: On April 17, 1995, the licensee initiated a reactor startup after RFO-13. The reactor was critical at 5:21 p.m. (CDT). On April 19 at 4:25 p.m., the main turbine generator was connected to the grid officially ending the outage. The planned 53-day outage, which began on February 24, 1995, was completed 1 day behind schedule due to emergent work. In addition to refueling the reactor, major activities included reactor vessel core shroud inspection, chemical decontamination of the reactor recirculation system, suppression pool and emergency core cooling system strainer inspection, replacement of both recirculation pump seal packages, and overhaul of the "B" low pressure turbine and the main generator. The next RFO is scheduled to begin in September 1996. Earlier on April 17, the reactor was taken critical at 6:50 a.m. As part of the startup testing, the licensee determined that the shutdown margin (SDM) may have been less than the technical specification (TS) required minimum and less than the predicted value. The startup was halted and the reactor was shutdown while the data was reevaluated by the licensee and General Electric (GE). Detailed evaluation of the data determined that the SDM was actually above the TS minimum. During the second startup, thelicensee and GE both verified that SDM was above the TS minimum. Regional Action: The resident inspectors will continue to observe power ascension activities and followup on the cause for the calculation error for the predicted value of SDM. Contact: J. A. HOPKINS (319)851-5111 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV APRIL 20, 1995 Licensee/Facility: Notification: Texas Utilities Electric Co. MR Number: 4-95-0053 Comanche Peak 1 Date: 04/20/95 Glen Rose,Texas Senior Resident Inspector Dockets: 50-445 PWR/W-4-LP Subject: STARTUP FROM REFUELING OUTAGE AND SUBSEQUENT FORCED OUTAGE Reportable Event Number: N/A Discussion: At 1:06 a.m. CDT on April 18, 1995, Unit 1 synchronized to the grid, thus ending the fourth refueling outage. At 5:10 a.m., the unit re-entered Mode 2 after the operators were unable to meet feedwater temperature requirements to open Feedwater Isolation Valve 4. The licensee's initial inspection revealed that a 3-inch manual isolation valve disk, on the feedwater isolation valve bypass line, separated from its stem. The licensee completed repairs on the valve and reentered Mode 1 at 1:09 p.m. on April 19. The licensee synchronized to the grid at 2:15 p.m. and increased power to 28 percent, where they will perform core flux mapping prior to further power ascension. The isolation valve was a 3-inch manually operated Borg-Warner solid wedge gate valve. The maintenance activity was performed in Mode 2 using single valve isolation. The licensee stated that this type of manual isolation valve has had similar previous problems. No maintenance activities had been performed on the valve during the refueling outage. The fourth refueling outage began on March 4. In addition to normal refueling maintenance activities, the licensee performed a 5-year containment spray nozzle and eductor test; performed steam generator eddy current testing and sludge lancing; and replaced one reactor coolant pump, one main transformer, and four containment spray pump impellers. During the outage, the licensee installed temporary diesels as part of a shut down risk enhancement, while performing maintenance on the emergency diesels. Regional Action: The residents will continue to follow the licensee actions. Contact: Harry Freeman (817)897-1500 David N. Graves (817)860-8141 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION IV APRIL 20, 1995 Licensee/Facility: Notification: Southern California Edison & San MR Number: 4-95-0054 Diego Gas & Electric Co. Date: 04/20/95 San Onofre 2 Resident Inspector San Clemente,California Dockets: 50-361 PWR/CE Subject: PRIMARY-TO-SECONDARY LEAKAGE Reportable Event Number: N/A Discussion: On April 16, 1995, with Unit 2 in Mode 3 during a refueling outage, Southern California Edison observed an increase in activity detected by the blowdown radiation monitor for Steam Generator 2EO88. The licensee sampled the steam generator and estimated that a 20 gallon per day primary-to-secondary leak might exist. On April 17, Unit 2 transitioned to Mode 2 for the performance of low power physics testing. During physics testing, the licensee took additional radiochemistry samples and confirmed that a 20 gallon per day primary-to-secondary leak existed in Steam Generator 2EO88. Unit 2 was returned to Mode 3 on April 20, after completion of low power physics testing. The licensee intends to return Unit 2 to Mode 5, reduce reactor coolant system level to midloop, perform a secondary side steam generator pressure test to find the source of the leak, and conduct appropriate repairs. San Onofre Unit 3 is operating at full power. Regional Action: The resident inspectors will monitor the licensee's actions to determine the source of the primary-to-secondary leakage and to effect repairs. Contact: D. Acker (510)975-0315 J. Sloan (714)492-2641