Headquarters Daily Report OCTOBER 23, 1995 *************************************************************************** REPORT NEGATIVE NO INPUT ATTACHED INPUT RECEIVED RECEIVED HEADQUARTERS û REGION I û REGION II û REGION III û REGION IV û PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS OCTOBER 23, 1995 MR Number: H-95-0129 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS Subject: PROGRAMMATIC WEAKNESSES IN PROBLEM IDENTIFICATION AND RESOLUTION AT SALEM, UNITS 1 AND 2 The NRR/AEOD/RES Events Assessment Panel on September 19, 1995, classified the recent performance problems at Salem culminating in the June 7, 1995, Unit 2 shutdown as a Significant Event. The significant event determination was based on programmatic weaknesses in the licensees ability to identify and correct conditions adverse to quality. Salem, Unit 2, commenced a Technical Specification required shutdown after declaring both trains of residual heat removal (RHR) inoperable on June 7, 1995. The RHR trains were declared inoperable based on the failure of their minimum flow valves to open on low flow as required to prevent pump failure. The deficiency was first noted on the 22 RHR pump on January 26, 1995, and on the redundant 21 RHR pump on February 9, 1995. However, the licensee did not determine the cause of the failures or initiate corrective action until June 7, 1995. Salem, Unit 1, declared emergency diesel generators (EDGs) 1B and 1C inoperable on June 7, 1995, while EDG 1A was inoperable due to maintenance, resulting in all three EDGs being declared inoperable. Unit 1 was in cold shutdown at the time. EDGs 1B and 1C were declared inoperable after review of a 1B EDG surveillance failure of June 1 revealed that the jacket water instrument lines were susceptible to vibration induced fatigue failure. The licensee had failed to correct the vibration problem despite several previous failures of the instrument lines. These issues and numerous others, including untimely assessment of degraded switchgear ventilation equipment in Unit 1 and untimely and inadequate resolution of identified nonconservatisms in the setpoint for low temperature overpressure transient conditions, were the subject of escalated enforcement resulting in a proposed civil penalty of $600,000. The licensee has received a confirmatory action letter and has agreed not to restart the units without first obtaining NRC agreement. Technical Contact: John Tappert, NRR/DRPM/PECB (301) 415-1167 _ REGION I MORNING REPORT PAGE 2 OCTOBER 23, 1995 Licensee/Facility: Notification: Maine Yankee Atomic Power Co. MR Number: 1-95-0133 Maine Yankee 1 Date: 10/23/95 Wiscasset,Maine SRI PC Dockets: 50-309 PWR/CE Subject: FAILURE OF BONNET BOLTING DUE TO BORON CORROSION Discussion: On October 20,1995, Maine Yankee engineering personnel determined that two manual loop isolation high pressure safety injection valves (HSI) had experienced bonnet bolting failure due to boron corrosion. Seven of eight bonnet bolts for HSI-26 broke when station maintenance personnel attempted to remove the bonnet bolting in order to facilitate replacement. Additionally, another bonnet bolt broke for HSI-16 during the removal process. The initial assessment by Maine Yankee determined that the valve design allowed for the collection of boron at the bolting, the bonnet bolting was carbon steel, and is not in the boron corrosion monitoring program since the bolts are not part of the pressure boundary. The purpose of the bolting is to hold the valve bonnet in place and allow normal system pressure to provide sealing. The valves are Anchor-Darling Company eight inch pressure seal type. The above repairs were being performed in response to finding all eight bonnet bolts for HSI-36 broken in April of 1995. The cause for this bonnet bolting failure was determined to be hydrogen embrittlement due to improper heat treatment during the manufacturing process. The failed bolting was replaced with bolting of the proper material. As a result of the bolting failure, the licensee identified a situation during plant shutdown where the potential exists for a reactor cavity drain down. The leakage occurs when the bonnet to body seating surface separates as the valve is positioned closed. While these valves are not isolable from the reactor coolant system they remain functional as isolation valves when closed. The licensee continues to evaluate the significance of this failure mode. In addition to the other repairs, a seal weld is being installed between the body and bonnet to eliminate the leak path. Maine Yankee has notified other nuclear facilities of the problem. Maine Yankee has expanded their review of safety related valves with this design and is also expanding the scope of the boron corrosion monitoring program to include this type valve. The inspection and repair will be complete prior to plant startup. Regional Action: Routine Resident Followup Contact: John Rogge (610)337-5146 William Olsen (207)882-7519 _ REGION I MORNING REPORT PAGE 3 OCTOBER 23, 1995 Licensee/Facility: Notification: Public Service Electric & Gas Co. MR Number: 1-95-0134 Hope Creek 1 Date: 10/23/95 Hancocks Bridge,New Jersey SRI PC Dockets: 50-354 BWR/GE-4 Subject: FAILURE OF THE REACTOR MANUAL CONTROL SYSTEM Discussion: On October 20, 1995 at about 9:00 p.m., operators found that the full core display and four rod display indications were not operating properly. Hope Creek was operating at about 94 percent power with all rods fully withdrawn (position 48) and a power coastdown in progress. At the time, operators were performing a routine rod exercise surveillance test. When the control room operator selected the first rod (a peripheral rod, number 02-19) and attempted to insert, he observed no indication of movement, either by position indication on the full core display or the four rod display and no indications of insert, withdrawal or settle lights changing state on the rod select module as would be expected. He re-attempted to insert the rod again with no indication of movement. The operator selected another peripheral rod, 10-11, and attempted an insert, again with no indication of movement. A third attempt on rod 02-19 was made without position indication changing. The operator then informed the Work Control Nuclear Shift Supervisor (NSS) (an on-shift SRO, but not the SRO who has control room command and control responsibility) about the problems with rod movement. Operators attempted to move rod 42-03, the last rod selected prior to the test; and again, no change to indicated position was observed. The Work Control NSS then reviewed the NSSS computer control rod display screen and found that: rod 02-19 was at step 42; rod 10-11 was at step 46; and rod 42-03 was at step 46. The "duty" SRO was informed of the failed reactor manual control system indication; abnormal operating procedures were entered; and following direction of the reactor engineer the three rods were then fully withdrawn relying upon the NSSS computer for position verification. Troubleshooting efforts identified poor contact on two card connectors (Source Select and Display Time cards) as causing the failure of the full core and four rod displays. When this was repaired, two hydraulic control unit accumulator alarms occurred, indicating that the loss of the display also masked the accumulator alarm function. Operators restored the two accumulators to normal operating conditions. Operators have commenced a temporary logging of the control rod select function on an hourly basis to ensure that the display has not locked up and to verify the accumulator trouble alarms are functioning. The licensee has commenced a multi-discipline team review of this event to identify operator performance and equipment issues. REGION I MORNING REPORT PAGE 4 OCTOBER 23, 1995 MR Number: 1-95-0134 (cont.) Regional Action: The resident inspectors are following licensee actions. Contact: Robert Summers (609)935-3850 Larry Nicholson (610)337-5128 _ REGION I MORNING REPORT PAGE 4 OCTOBER 23, 1995 Licensee/Facility: Notification: New York Power Authority MR Number: 1-95-0135 Indian Point 3 Date: 10/23/95 Buchanan,New York SRI PC Dockets: 50-286 PWR/W-4-LP Subject: INDIAN POINT 3 SITE MANAGEMENT CHANGES Discussion: Effective October 23, 1995, Indian Point 3 has implemented the following management changes. Marc Pearson has replaced Nick Eggemeyer as the Operations manager. The position of General Manager of Operations (GMO) has been deleted. The Operations manager will now report directly to the Site Executive Officer. The former GMO will lead a special task force to upgrade operations procedures. The two other departments which had also previously reported to the GMO will now report to the General Manager Support Services. These two departments are the Performance and Reliability department and the Radiological and Environmental Services department. John Kelly has been assigned as the new General Manager Support Services. Mr. Kelly had previously been the Director of Regulatory Affairs and Special Projects. The former General Manager Support Services, Jim Comiotes, was recently reassigned to the position of Tactical Assessment Coordinator. Regional Action: None Contact: Curtis Cowgill (610)337-5233 _ REGION III MORNING REPORT PAGE 5 OCTOBER 23, 1995 Licensee/Facility: Notification: Toledo Edison Co. MR Number: 3-95-0158 Davis Besse 1 Date: 10/23/95 Oak Harbor,Ohio SRI VIA PC. Dockets: 50-346 PWR/B&W-R-LP Subject: SITE MANAGEMENT CHANGES Discussion: Effective 10/16/95, Mr. Sushil Jain, Davis-Besse (DB) Engineering and Services Director has been reassigned to the licensee's corporate office as the Engineering Services Director. As such he will no longer have any responsibilities for activities at the Davis-Besse site. A permanent replacement for Mr. Jain has not yet been announced. In the interim, Mr. James Michaelis, currently the DB Nuclear Support Manager will be the acting DB Engineering and Services Director as well. Also, Mr. William T. O'Connor, Manager of DB Regulatory Affairs has announced his resignation from the company effective 10/31/95. Mr. O'Connor has been hired by Detroit Edison to be the Manager, Assessment at the Fermi Nuclear Station. A replacement for Mr. O'Connor has yet to be named. Regional Action: None. Contact: R.D. LANKSBURY (708)829-9631 _ REGION IV MORNING REPORT PAGE 6 OCTOBER 23, 1995 Licensee/Facility: Notification: Southern California Edison & San MR Number: 4-95-0130 Diego Gas & Electric Co. Date: 10/20/95 San Onofre 2 SRI San Clemente,California Dockets: 50-361 PWR/CE Subject: FAILURE OF TRIP-THROTTLE VALVE ON TURBINE-DRIVEN AUXILIARY FEEDWATER PUMP Discussion: On October 18, 1995, the licensee performed an annual overspeed trip test of the Unit 2 turbine-driven auxiliary feedwater pump. Following the trip, the licensee was unable to drive the motor actuator for the trip-throttle valve fully closed to enable the trip mechanism to be reset. The as-found condition of the Limitorque actuator was that the torque switch had opened, which was unexpected, and the torque switch bypass limit switch had opened. The valve had closed most of the way before the failure. Inspection of the actuator revealed that excess grease (Mobil 28 lithium-based grease) from the stem had become dirty and worked its way onto the surfaces of the bronze sliding nut, which was not supposed to be lubricated. The dirty grease prevented the sliding nut from moving as necessary for the valve to relatch. Additionally, the grease had deteriorated somewhat in the high-temperature environment. The licensee cleaned the actuator components and lubricated appropriate parts. Postmaintenance testing has not yet been completed. The NRC's Headquarters-based Integrated Performance Assessment Process (IPAP) team was onsite and observed the failure from both the control room and the valve location. The Senior Resident Inspector was also onsite and responded to the failure. The licensee's preliminary review of records indicated that the valve and actuator had been replaced with new components in early 1991 and that the sliding nut had never been inspected for cleanliness. The sister valve in Unit 3 was replaced during the 1995 refueling outage. The Senior Resident Inspector visually inspected the Unit 3 valve and determined that no degraded condition was apparent. Regional Action: The IPAP team and the resident inspectors will review the licensee's followup actions. Contact: Dennis F. Kirsch (510)975-0290 Jim Sloan (714)492-2641 _ REGION IV MORNING REPORT PAGE 7 OCTOBER 23, 1995 Licensee/Facility: Notification: Nebraska Public Power District MR Number: 4-95-0131 Cooper 1 Date: 10/23/95 Brownville,Nebraska Licensee Dockets: 50-298 BWR/GE-4 Subject: LICENSING BASIS CONCERNS WITH UTILIZATION OF SPENT FUEL POOL Discussion: On October 20, 1995, the NRC staff discussed concerns with CNS management regarding the licensee's utilization of the spent fuel pool and the spent fuel pool cooling system. The licensee's Updated Safety Analysis Report (USAR) indicated that the normal utilization of the spent fuel pool and cooling system involved servicing a heat load associated with a discharge of approximately 160 fuel bundles (less than one-third core) and maintaining fuel pool temperature at 125 degrees Fahrenheit. The USAR referenced the ability of the system to accommodate an emergency heat load associated with an unscheduled discharge of the entire core just prior to a refueling outage. The USAR wording indicated that, for this case, the spent fuel pool cooling system would be cross-tied with the residual heat removal system to maintain pool temperature below 150 degrees Fahrenheit. The licensee has in the past performed full core offloads to facilitate refueling operations and on October 20 was in the process of performing a full core offload to support its current refueling outage which began on October 15, 1995. A 10-year inservice inspection of the reactor vessel is within the scope of the current refueling outage. The licensee halted offload of fuel assemblies at 160 fuel bundles upon learning of the NRC concerns. The licensee researched its licensing basis and found that the USAR was in error. It was found that the licensing basis for the spent fuel pool and cooling system had been modified coincident with a 1977 licensing amendment associated with reracking of the fuel pool. At that time, it was defined that the fuel pool and cooling system would normally service the heat load associated with a full core offload and that the residual heat removal assist mode of operation would be utilized to maintain pool temperature below 150 degrees Fahrenheit. However, this normal heat load was defined to be associated with a core that was discharged to the pool 13 days after shutdown. The licensee's current schedule would have had the core offload completed in less than 12 days. The worst case spent fuel pool heat load was defined as a full core discharge of an end-of-cycle core coincident with a safe shutdown earthquake which would disable the spent fuel pool cooling system. In this scenario, it was determined that sufficient time was available to provide make-up water to the pool from one of four defined sources prior to the onset of bulk boiling. The licensee performed a 10 CFR 50.59 analysis to change the description of the utilization of the spent fuel pool and cooling system as described in the USAR and modified its refueling procedures to be consistent with the revised USAR. The licensee resumed offloading the core at 9:28 a.m REGION IV MORNING REPORT PAGE 8 OCTOBER 23, 1995 MR Number: 4-95-0131 (cont.) (CDT) on October 23, 1995. The licensee has instituted administrative controls to ensure that core offloading does not occur at a rate that would have the core discharged prior to 13 days after shutdown (currently in its 9th day). Regional Action: Inspectors independently verified that only 160 bundles had been offloaded prior to the suspension of operations on October 20, 1995. Inspectors will verify that offload rate is consistent with the analysis. An inspection is planned to assess past licensee performance. The 10 CFR 50.59 analysis is being reviewed concurrently by the Region and the Office of Nuclear Reactor Regulation (NRR). NRR has the lead for the technical review. Contact: T. Reis (817)860-8185 _