Headquarters Daily report SEPTEMBER 08, 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 SEPTEMBER 8, 1994 MR Number: H-94-0079 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: Potential for Damage to Reactor Coolant System Pressure Boundary The NRR/AEOD Events Assessment Panel on September 6, 1994, classified a plant condition found during scheduled surveillance, that involved multiple failures of the safety/relief valves (SRVs) at Millstone Unit 1 to operate at the intended pressure, as a Significant Event for the NRC Performance Indicator Program. Millstone utilizes a total of six SRVs that are a combination safety and relief valve. Classification was based on degradation of important safety equipment that had the potential to degrade the primary coolant pressure boundary (failure of SRVs to open) and reactor operation potentially outside the limits of the technical specifications (all SRVs are to be operable). It is infeasible to quantitatively determine the risk significance of this event. Unknown parameters to which the quantification process is very sensitive include: 1) the pressure at which the coolant pressure boundary would fail, 2) the location of the boundary failure, and 3) the probability that the two valves that did not open at the maximum test pressure would not open at a higher pressure sufficiently low to prevent boundary failure. On March 29, 1994, subsequent to a refueling shutdown and testing of the SRVs that were removed for refurbishing, the licensee stated that of the six SRVs, two did not open up to the maximum test pressure of about 1250 psig (desired set pressure is 1125 psig), and the lift pressure of each of the other four SRVs exceeded the technical specification tolerance of one percent. The average overpressure lift of the four valves was greater than six percent. It is not known when the set pressures "drifted", but the valves had not been required to operate during the last operating cycle or since they were last refurbished and installed in June 1991. An analysis was performed to determine the potential effect of the higher than expected safety/relief valve lift pressures. The licensee concluded that the safety limit of 1375 psig for the reactor coolant system would not have been exceeded. However, the calculated maximum pressure was in excess of 1350 psig assuming 1) the "worst case" anticipated transient which has all steam isolation valves closing, 2) that the four operable valves lifted at the as found lift pressure, and 3) that the two valves that failed to open were assumed not to operate. Since the peak pressure occurs almost immediately after closure of the steam isolation valves, no consideration was given for potential operator action to utilize the manually initiated lift feature of the safety/relief valves. The cause of the "drifting" was attributed to oxide bonding of the seat and disk in the pilot valve of the SRVs. SRV set point drift is a problem that has been experienced previously at Millstone Unit 1 and at other facilities with two stage valves manufactured by Target Rock. As a result of a long term effort to correct this problem, and at the recommendation of the Boiling Water Reactor Owner's Group, the licensee replaced three of the pilot valves with valves having a new disk material - a platinum stellate alloy. This material is intended to reduce excess oxygen by the recombination of oxygen and hydrogen that collects in the pilot valve as a result of the radiolytical breakdown of the water. Operating experience with pilot valves with this new material is to be evaluated; laboratory experience has been positive. CONTACT: J. Carter, NRR/DORS/OEAB (301) 504-1153 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION SEPTEMBER 8, 1994 MR Number: H-94-0080 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: Inadvertent Pump Down of Reactor Vessel The NRR/AEOD Events Assessment Panel on June 14, 1994, classified the inadvertent pumping of reactor coolant inventory to the drywell at Millstone Unit 1 as a Significant Event for the NRC Performance Indicator Program. A test procedure led to connecting the operating shutdown cooling (SDC) system to an "open" containment spray system. Classification was based upon programmatic deficiencies and a quantitative risk significance determination was not feasible. On April 10, 1994, in preparation for a reactor restart following a refueling outage, the licensee was performing a surveillance test of the Low Pressure Coolant Injection (LPCI)/Containment Cooling System logic that utilized a recently revised and approved procedure. The logic test, which had been performed many times in the past, had been recently revised to incorporate actuation testing of valves and relays. The licensee's established procedure for review and approval was followed. Prior to giving approval to perform the test, the licensee recognized that a large number of systems and components had been returned to operable status in preparation for reactor restart and that the decay heat was low. One part of the procedure required verifying that 3 pairs of series isolation valves would operate. When the one pair of containment isolation valves to the containment sprays was opened, a flow path for reactor coolant being pumped in the operating SDC system was established via LPCI train A and train B and the containment spray system to the drywell. An operable drywell sump hi-level alarm alerted the operators to a problem which was associated with the testing. The isolation valves were closed within about four minutes of the initial loss of inventory, but approximately 12,000 gallons were lost which decreased the water level in the reactor vessel from 85 inches to 15 inches. A further decrease in vessel water level would have actuated a low level alarm and automatic isolation of the SDC system, thereby stopping the loss of inventory. The loss of reactor coolant was caused by a deficient procedure. Two procedures had been recently combined, primarily by the Instrument/Controls staff, and processed as required. Those reviewing and assessing the revised procedure did not recognize the significance of the changes nor were the operators trained to use the new procedure. Thus, the procedure resulted in connecting two trains of LPCI which provided the flow path for the coolant loss. In addition, some operable equipment was placed in a status that would prevent automatic initiation of corrective action. A limited amount of equipment that could function to alert the staff or mitigate the loss of reactor vessel level is required to be "operable" during reactor shutdown. Following the event, the licensee reassessed all such combined procedures and made any necessary corrections. No other significant unrecognized system interactions were found. NRC is preparing an information notice to alert licensees to the importance of comprehensive procedure review and equipment control. Contact: J. Carter, NRR/DORS/OEAB (301) 504-1153 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION SEPTEMBER 8, 1994 MR Number: H-94-0081 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: Failure to Take Effective Corrective Action for Breaker and Damper Performance Problems The NRR/AEOD Events Assessment Panel on September 6, 1994, classified the failure to take effective corrective action for breaker and damper performance problems at LaSalle, Units 1 and 2, as a Significant Event for the NRC Performance Indicator Program. The identified violation was categorized at Severity Level III and notification of significant enforcement action (EN 94-034) was issued on March 30, 1994, involving proposed civil penalty of $75,000, which was paid by the licensee on May 13, 1994. The equipment problems themselves were not significant, but the licensee's ineffective corrective action and the recurrence of failures are due to a significant programmatic deficiency. A quantitative risk significance determination of the programmatic deficiency was not feasible. On March 3, 1989, a 10 CFR Part 21 report was filed with the NRC concerning contaminated grease in the operating mechanism of two HK series circuit breakers. The contaminated grease resulted in the breakers failing to close on demand and perform their safety function. In response to this report, the breaker manufacturer issued a letter to all nuclear facilities where these breakers were known to be in service. Over three years later, several communications were exchanged between the vendor, a contracted engineering firm and the licensee. However the licensee concluded that no immediate actions were necessary and that the procedure revision and its associated visual inspection were adequate to address the issue. In spite of that, the licensee experienced repetitive failure of the same type of breakers in October 1992 and in October 1993 caused by the hardened grease. On November 29, 1993, the B RPS bus tripped and it caused both Unit 1 and 2 Secondary Containment Isolation Reactor Building Ventilation (VR) systems to trip. However, the VR supply damper did not fully close as expected. Before this event, the VR dampers had a history of repetitive failures and several vendor recommendations and proposals had been issued to the licensee responding to the failures. A vendor's letter dated September 20, 1989, stated that the damper blades were rubbing on the bottom inside surface of the body and this friction force opposed the spring closing force and stopped the blades just short of complete closure. The vendor recommended addition of hangers from the hinge area to the blade. In a letter dated March 17, 1992, the vendor further recommended that the dampers' actuators be replaced with larger ones to provide an actuator with an integral spring. The increased return spring force would overcome the dampers friction force. In spite of these, the licensee failed to identify and resolve the problem effectively. Contact: T. Yamada, NRR/DORS/OEAB (301) 504-1170 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION I SEPTEMBER 8, 1994 Licensee/Facility: Notification: Public Service Electric & Gas Co. MR Number: 1-94-0101 Salem 1 2 Date: 09/08/94 Hancocks Bridge,New Jersey Dockets: 50-272,50-311 PWR/W-4-LP,PWR/W-4-LP Subject: Project Engineer PC Reportable Event Number: N/A Discussion: Public Service Electric & Gas Co. (PSE&G) Chairman James Ferland announced on September 7th that Leon R. Eliason, currently president of Northern States Power Company (NSP), Minneapolis, Minn., has been elected Chief Nuclear Officer for PSE&G, and president of a newly structured business unit. Eliason's appointment is effective October 1, 1994 and he will report directly to Ferland. Eliason will replace Steven E. Miltenberger as Chief Nuclear Officer. The newly formed Nuclear Business Unit will encompass all operational and support activities for both Salem units, as well as, Hope Creek. Mr Eliason joined NSP in 1965 and has held increasingly responsible positions, such as, Monticello Plant Manager, General Manager-Nuclear Plants, and Vice President-Nuclear Generation. NSP operates Monticello Nuclear Station (535 MWe-GE BWR) and Prarie Island Nuclear Station (two 503 MWe-WEC PWRs). Regional Action: This morning report was provided for informational purposes. Contact: John White (610)337-5114 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION I SEPTEMBER 8, 1994 Licensee/Facility: Notification: Connecticut Yankee Atomic Power Co. MR Number: 1-94-0102 Haddam Neck 1 Date: 09/08/94 Hartford,Connecticut RI PC Dockets: 50-213 PWR/W-4-LP Subject: SITE MANAGEMENT CHANGE Reportable Event Number: N/A Discussion: On September 7, Connecticut Yankee Atomic Power Company (CYAPCo) announced the termination of Mr. T. McDonald as Maintenance Manager. The job action was the result of a company audit concerning internal allegations. The replacement maintenance manager is expected to be announced in the next couple of weeks. Regional Action: Regional and Resident Inspector monitoring the performance effects of the recent management change. Contact: William Raymond (203)267-2571 Stephen Barr (610)337-5173 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III SEPTEMBER 8, 1994 Licensee/Facility: Notification: Northern States Power Co. MR Number: 3-94-0158 Prairie Island 1 2 Date: 09/08/94 Welch,Minnesota RI PC Dockets: 50-282,50-306 PWR/W-2-LP,PWR/W-2-LP Subject: MANAGEMENT REORGANIZATION Reportable Event Number: N/A Discussion: On September 7, 1994, the licensee announced the following management changes, effective immediately. Leon Eliason has resigned his position as President, Northern States Power Company (NSP) - Generation, to accept the position of Chief Operating Officer for New Jersey Public Service Company. Mr. Eliason was replaced by Doug Antony, the former Vice President - Nuclear. Ed Watzl, the former Site Manager - Prairie Island Nuclear Generating Plant (PINGP), was promoted to Vice President - Nuclear. Mike Wadley, Plant Manager - PINGP, will act as site manager in addition to plant manager pending NSP's review of the status of the site manager position. Regional Action: Information only. Contact: W. KROPP (708)829-9633 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III SEPTEMBER 8, 1994 Licensee/Facility: Notification: Alt And Witzig Engineering, Inc. MR Number: 3-94-0159 Alt And Witzig Engineering, Inc. Date: 09/06/94 Indianapolis,Indiana TELEPHONE CALL FROM THE LICENSEE License No: 030-14041 Subject: STOLEN AND RECOVERED MOISTURE/DENSITY GAUGE Reportable Event Number: N/A Discussion: At approximately 4:30 p.m. on September 6, 1994, Alt & Witzig Engineering, Inc's Radiation Safety Officer (RSO) reported the theft of a Campbell Pacific, Model CPN, moisture/density gauge from the back of an open bed pickup truck located at a temporary jobsite of the licensee in Florence, Kentucky. The missing gauge contained a nominal 10 millicurie (370 MBq) cesium-137 sealed source and a nominal 40 millicurie (1.85 GBq) americium-241 sealed source. The theft occurred while the gauge operator was making a telephone call from a construction trailer. The operator had not secured the gauge from unauthorized removal. At approximately 8:00 a.m. on September 7, 1994, the RSO reported that the gauge had been found behind a portable toilet at the jobsite. The RSO believed that a construction worker had removed the gauge from the truck on Tuesday afternoon and returned the gauge to the jobsite Wednesday morning. The RSO stated that there was no visible damage to the gauge and the source rod was locked in its shielded position. The RSO stated the gauge will be wipe tested for leakage and kept out of service until he receives the results of the wipe test. Since the source rod was locked in its shielded position when the gauge was found, the device did not pose a significant radiation hazard to workers or members of the general public while it was missing. Regional Action: Region III notified NMSS, the Commonwealth of Kentucky, and the Region II State Agreements Officer of the incident. Contact: W.P. REICHOLD (708)829-9839 B.J. HOLT (708)829-9836 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION III SEPTEMBER 8, 1994 Licensee/Facility: Notification: Detroit Edison Co. MR Number: 3-94-0160 Fermi 2 Date: 09/07/94 Newport,Michigan FROM RESIDENT SITE Dockets: 50-341 BWR/GE-4 Subject: SPILL OF 1000 GALLONS OF CONDENSATE WATER DUE TO SEAT LEAKAGE Reportable Event Number: N/A Discussion: On 9/7/95 while pumping the torus to main condenser hotwell approximately 1000 gals of slightly contaminated water spilled into the condensate pump bay. The licensee isolated the area and pumped out the water. No equipment damage or radioactive release occurred. Based on preliminary investigation the leak occurred due to excessive seat leakage past a condensate pump isolation valve. Regional Action: Resident inspector was on site and is following up on the event. Contact: M.P. PHILLIPS (708)829-9637