Headquarters Daily report APRIL 28, 1994 *************************************************************************** REPORT NEGATIVE NO INPUT ATTACHED INPUT RECEIVED RECEIVED HEADQUARTERS X REGION I REGION II REGION III REGION IV *************************************************************************** PRIORITY ATTENTION REQUIRED MORNING REPORT - HEADQUARTERS APRIL 28, 1994 MR Number: H-94-0042 NRR DAILY REPORT ITEM GENERIC COMMUNICATIONS BRANCH/EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT OFFICE OF NUCLEAR REACTOR REGULATION Subject: N/A NRC Information Notice 93-53, Supplement 1, "Effect of Hurrican Andrew on Turkey Point Nuclear Generating Station and Lessons Learned," will be issued April 29, 1994. The NRC is issuing this information notice supplement to inform addressees of further lessons learned as a result of the investigations undertaken to assess the effects of Hurricane Andrew on the Turkey Point Nuclear Generating Station (Turkey Point); to expand the scope of the lessons learned to other external events, as appropriate; and to discuss existing regulatory guidance for various external events. Technical contacts: Hans Ashar, NRR (301) 504-2851 John T. Chen, RES (301) 492-3919 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION APRIL 28, 1994 MR Number: H-94-0043 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: POTENTIALLY UNAVAILABILITY OF ALL CORE COOLING CAPABILITY FROM FAILURE OF EMERGENCY LOAD SEQUENCER The NRR/AEOD Events Assessment Panel on April 26, 1994, classified a design deficiency in the Emergency Load Sequencing System at Beaver Valley, Unit 2, as a Significant Event. Classification was based on the potential risk from the unavailability of both trains of core cooling system. On November 4, 1993, during a surveillance test, the Emergency Load Sequencer (ELS) failed to automatically load safety related equipment to the emergency bus. The same surveillance conducted on the redundant train, with additional diagnostic equipment, found the same problem and identified the root cause as a system "lock up" resulting from a failed microprocessor based relay. The problem was repeatable once in three attempts. Both trains of ELS were susceptible for this "lock up". In the event of an accident, without off site power, the ELS failure could prevent the automatic initiation of core cooling. Operator actions may not be sufficiently timely for all design bases accidents. These deficiencies occurred after licensee's modification to replace unreliable electro-magnetic timing relays with micro-processor based relays. The licensee's dedication process was inadequate to identify the relay's vulnerability to voltage spikes generated by operation of other electro-mechanical relays in the same system. The licensee corrected the problem by reducing the magnitude of the voltage spikes with surge suppressor diodes. An AIT inspection examined the event and the licensee's corrective actions. The risk associated with this event is estimated to be a CDF increment in the mid 10-5 to low 10-4 range. The CCDP for the total duration of two fuel cycles would be in the low 10-4 range. This calculation assumes random sequencer failure affecting both safety trains due to the deficiency, requiring operator action away from control room for recovering service water, and power operation with 70% capacity factor. Contact: Thomas Koshy, NRR/DORS/EAB (301) 504-1176 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION APRIL 28, 1994 MR Number: H-94-0044 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: LOSS OF OFFSITE POWER AND FAILED MAIN STEAM ISOLATION VALVE The NRR/AEOD Events Assessment Panel on April 26, 1994, classified the loss of offsite power and failed main steam isolation valve at McGuire, Unit 2 as a Significant Event for the NRC Performance Indicator Program. On December 27, 1993, with McGuire, Unit 2, operating at 100% power, the 525kV bus line "2B" was lost due to a failed insulator. An expected turbine runback failed to initiate, causing breakers for bus line "2A" to open on overcurrent protection, which resulted in a loss of offsite power (LOOP) to the unit. The reactor tripped and emergency diesel generators (EDGs) successfully started and loaded. The plant experienced an excessive cooldown due to full auxiliary feedwater (AFW) flow and unisolated secondary steam loads. Safety injection initiated on low pressurizer pressure approximately 7 minutes into the event, and low steam line pressure actuated a main steam isolation valve (MSIV) closure signal approximately 1 minute later. MSIV "B" failed to close fully and attempts to manually close the valve were unsuccessful, which resulted in steam generator (SG) "B" reaching a dryout condition. Differential pressure (P) across the tubes of SG "B" reached 1981 psid before the operators were able to reduce the P to less than 1600 psid (as suggested by Westinghouse guidelines) using pressurizer power operated relief valves (PORVs). Offsite power was restored approximately 1.5 hours into the event and the vital buses were realigned to offsite power approximately 2.5 hours into the event. An Augmented Inspection Team (AIT) was dispatched to the site to review the event. The AIT noted weaknesses in the licensee's electrical system design, MSIV maintenance, procedural adherence, document control, and corrective actions from a previous LOOP. The AIT findings are documented in Inspection Report 50-369/93-33 and 50-370/93-33. An Information Notice has been drafted to inform licensees of MSIV binding which occurred at McGuire due to the licensee setting valve clearances and testing the valve while cold, versus normal operating temperature as recommended by the valve vendor. A risk evaluation performed by the Probabilistic Safety Assessment Branch estimates the total CCDP for this event to be 2 x 10-4. The event was modelled in the ASP program as a plant-centered LOOP with the pressurizer PORVs challenged (PORVs automatically cycled to control increasing reactor coolant system pressure subsequent to safety injection, and PORVs manually cycled to reduce P across SG "B" tubes). The Standby Shutdown Facility at McGuire was included in the analysis and its total system failure probability was assumed to be 0.34, as defined in NUREG/CR-4674. The contribution to the total CCDP from the sequence listed above, is estimated to be 1.1 x 10-4. Because P across the SG "B" tubes was increased during this event, the potential for inducing a SG tube rupture was evaluated. The conditional probability of inducing a SG tube rupture was calculated using an equation given in NUREG-0844 which considers the actual peak P seen across the SG tubes during an event. The contribution to the total CCDP from transient induced SG tube rupture sequences is estimated to be 9.4 x 10-5. Contact: Eric Benner, NRR/DORS/EAB (301) 504-1171 PRIORITY ATTENTION REQUIRED MORNING REPORT - REGION APRIL 28, 1994 MR Number: H-94-0045 NRR DAILY REPORT ITEM SIGNIFICANT EVENTS EVENTS ASSESSMENT BRANCH DIVISION OF OPERATING REACTOR SUPPORT Subject: SEPARATION OF VALVE STEM FROM MOTOR OPERATOR The NRR/AEOD Events Assessment Panel on April 26, 1994, classified the inoperability of a recirculation spray pump suction valve at Beaver Valley Unit 2, due to separation of the valve stem from the motor operator, as a Significant Event for the NRC Performance Indicator Program. On October 17, 1993, the licensee was attempting to perform a full-flow surveillance test on one of the recirculation spray pumps. The pump did not pass any flow when it was started. Although the pump suction valve remained closed, control room indication of valve position showed the valve to be open because the position indication was derived from motor operation, not from stem rotational position itself. The licensee found that the valve, which had been in the closed position at the start of the test, had not opened because the valve stem had become uncoupled from the motor operator. The licensee inspected the three other valves in similar service and found two were starting to decouple. During normal plant operation, the four valves of concern are in the closed position, isolating the recirculation spray system pumps from the containment sump. In the event of a loss-of-coolant accident, they remain closed until about ten minutes after a containment isolation phase B signal is received. The four trains are then activated and the pumps circulate water from the sump, through heat exchangers, to the containment spray headers. When safety injection has lowered the level in the refueling water storage tank to a pre-determined low level, two of the four trains are switched from the spray recirculation mode to the safety injection cold leg recirculation mode, providing water directly to the reactor coolant system via the low-head safety injection system as well as to the suction of the high-head safety injection pumps. If the four suction valves do not open, neither containment spray nor coolant injection would operate as designed to mitigate the accident. Based on the potential common-cause failure of all four trains, the Events Assessment Panel classified the event as a Significant Event. The event will be evaluated for risk insight by Probabilistic Safety Assessment Branch. The valves are Model 1400 12-inch butterfly valves manufactured by Henry Pratt Company. The motor operator is a Limitorque Model HBC-1. The stem is coupled to the operator drive gear through a cylindrical spline adaptor. The bottom of the spline adaptor rests on a mounting plate through which the valve stem passes. The hole in the mounting plate is designed to be slightly larger than the valve stem but smaller than the spline adaptor. In the case of the four valves at Beaver Valley Unit 2, the mounting plate holes were sufficiently over-size that the spline adaptor could drop down through the hole far enough that the adaptor would no longer be engaged with the motor drive gear. The licensee made temporary repairs by installing a split-cylinder clamp on the valve stem to provide a seating surface for the spline adaptor. Permanent repairs, in the form of new mounting plates with proper hole size, are intended to be installed at the next refueling outage. The four valves noted above are the only safety-related valves affected at either of the Beaver Valley reactors. The licensee has provided notification of this situation in accordance with the requirements of 10 CFR Part 21. Contact: Robert Benedict, NRR/DORS/EAB (301) 504-1157