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| Entered date | Event description | |
|---|---|---|
| ENS 43139 | 2 February 2007 21:05:00 | At 1350 hours on February 2, 2007, with the plant at 100% power, it was determined (that) the low pressure injection (LPI) system net positive suction head calculation does not account for the additional flow through the failed LPI pump recirculation line during certain accident scenarios. The additional flow is upstream of the flow element used by control room operators to throttle system flow to maintain net positive suction head flow requirements. The additional flow could result in net positive suction head below required design limits. The system design is not affected in events where both LPI trains perform as designed. Emergency operating procedures direct control room operators to open the LPI system discharge flow cross-connect line isolation valves, if accessible, following a LPI pump failure. Operators are then directed to throttle system flow through the operable LPI pump to maintain proceduralized values. These values are designed to provide sufficient design flow and maintain pump NPSH. During a simulator training scenario, operators identified when the discharge cross-connect line isolation valves were opened, the idle Building Spray train indicated flow. Follow-up investigation identified the increased flow was due to back flow through the failed LPI pump minimum flow recirculation line. This additional flow is upstream of the flow element used by operators to maintain adequate net positive suction head for the operable LPI pump. The additional flow could result in not meeting NPSH design requirements. The licensee entered the 72 hour Technical Specification limiting condition for operation (LCO) for one inoperable LPI train. The licensee is revising calculations and emergency operating procedures to account for the additional flow. This condition is reportable in accordance with 10CFR 50.72(b)(3)(ii) and (b)(3)(v) as an unanalyzed condition, and a condition that could have prevented the fulfillment of the safety function of the LPI system to mitigate the consequences of an accident, respectively. The NRC Resident Inspector was notified of this event by the licensee.
The purpose of this report is to retract the ENS report made on February 2, 2007 at 2105 hours ( ENS #43139) under 10CFR50.72(b)(3)(ii) and (b)(3)(v) as an unanalyzed condition, and a condition that could have prevented the fulfillment of the safety function of the Low Pressure Injection (LPI) system to mitigate the consequences of an accident, respectively. The initial report was made when it was determined that the LPI system net positive suction head (NPSH) calculation does not account for the additional flow through the LPI pump recirculation line during certain accident scenarios. The additional flow could result in net positive suction head below required design limits. Due to this condition, it was not certain if the LPI system could have met its design basis requirements. The licensee entered the 72 hour Technical Specification limiting condition for operation (LCO) for one inoperable LPI train. The LCO was exited on February 3, 2007 at 9:25PM following implementation of a procedure change that accounted for the additional flow and ensured that adequate NPSH was maintained. A subsequent engineering evaluation has determined that sufficient LPI pump NPSH would have been available to perform its design basis function prior to the procedure change. The engineering evaluation shows that the LPI pumps remained capable of performing their design basis functions based on the following three independent assessments: 1) the LPI pumps would have operated well beyond their mission time without significant cavitation damage at the available NPSH 2) proceduralized operator actions would have throttled flow to restore required NPSH if signs of cavitation occurred 3) an evaluation using realistic Reactor Building pressures showed that sufficient NPSH would exist. The licensee notified the NRC Resident Inspector. Notified R1DO (Hott) |
| ENS 43050 | 13 December 2006 21:00:00 | While operating at 100 % power the unit experienced an automatic Reactor trip from Reactor Protection System actuation at 17:48 on 12/13/06. At the time of the trip there was a grid disturbance followed by the unit tripping. The cause of the trip is under investigation. All plant systems functioned as designed. The unit is being maintained in Hot Shutdown conditions using applicable plant procedures. All control rods inserted fully after the reactor trip. S/G safety valves lifted and reset per design during the transient. The S/Gs are being supplied with normal feedwater and decay heat is being removed to the main condenser via the steam dump valves. The electric plant is in a normal shutdown lineup and the EDGs are available. The licensee notified the NRC Resident Inspector, PEMA, the Counties of Dauphin, Cumberland, Lancaster, York, and Lebanon. |
| ENS 42527 | 25 April 2006 18:04:00 | TMI Issue Report # 482679 identified an issue while performing reviews of fire abnormal operating procedures to assure compliance with the Fire Hazards Analysis Report (FHAR), in that a control logic error was identified in the circuitry elementary drawing for the isolation valves DH-V-6A and DH-V-6B between the Borated Water Storage Tank (BWST) and the Reactor Building (RB) sump. Plant circuitry was verified to be wired as per the elementary drawing. This circuitry design was to prevent a hot short, due to a fire, from opening the valve. However, the identified control logic error could allow a spurious opening to occur on DH-V-6A or DH-V-6B due to a fire. The FHAR credits these valves as being protected from spuriously opening due to a fire in AB-FZ-5 (Auxiliary Building 281' general area). If this protection is not provided, then spurious opening could result in draining the BWST inventory to the RB sump. This hot short condition would result in the depletion of the BWST inventory and loss of the High Pressure Injection (HPI) makeup capability, resulting in an unanalyzed condition that significantly degrades plant safety. This condition is reportable under 10 CFR 50.72(b)(3)(ii)(B), 'Any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety.' A 60-day LER is also required under 10 CFR 50.73(a)(2)(ii)(B) for the same degraded condition. An hourly fire-watch has been established in the affected fire zone in the 281' elevation Auxiliary Building as an interim compensatory measure. Additionally, the control circuitry at the 1A and 1B ES MCCs will be modified to prevent the RB sump isolation valves DH-V-6A and DH-V-6B from spuriously opening due to a hot short. The licensee notified the NRC Resident Inspector. |
| ENS 42416 | 14 March 2006 13:42:00 | A licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been put on administrative hold, and the individual has been escorted offsite. Contact the Headquarters Operations Officer for additional details. The licensee has informed the NRC Resident Inspector. * * * * UPDATE ON 3/14/06 AT 1534 FROM S. BRANTLEY TO P. SNYDER * * * * The licensee will be issuing a press release on this event. The licensee has informed the NRC Resident Inspector. |
| ENS 42413 | 13 March 2006 15:35:00 | The Technical Support Center ventilation system was discovered to be non-functional at 1110 on March 13, 2006. The cause of the ventilation problem was determined to be damaged drive belts on the supply fan. The belts were replaced and the ventilation system supply fan was returned to service at 1253 on March 13, 2006. This condition is considered a Loss of Offsite Response Capability and is therefore reportable under 10CFR50.72(b)(3)(xiii). A corrective action request has been generated to follow remedial action. The licensee notified the NRC Resident Inspector. |