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| Entered date | Event description | |
|---|---|---|
| ENS 49043 | 16 May 2013 16:46:00 | At approximately 1934 EDT on May 15, 2013, maintenance personnel entered the primary containment personnel air lock to determine the cause of the inability to attain test pressure during a type B leak rate of the airlock. The inner door seal was found degraded and partially rolled from its required position allowing air from inside the airlock to enter the primary containment. During the limited time the outer airlock door was opened for access into the airlock concurrent with the degraded seal on the inner door, a condition existed that could have prevented the fulfillment of the safety function of the structure to control the release of radioactive material. The inner door seal was subsequently replaced and the leak rate of the personnel air lock completed with satisfactory results. During the seal replacement activity, the outer airlock door remained closed to provide the barrier against the release of radioactive material should it be required. This is a notification per 10 CFR 50.72(b)(3)(v) for a condition that could have prevented the control of the release of radioactive material. It is recognized that this notification (was) not within eight hours of the event. The condition has been entered into the station's corrective action program. The licensee will notify the New York State Public Service Commission and the NRC Resident Inspector.
This notification is being made to retract Event Notification #49043, which reported a condition that could have prevented the control of the release of radioactive material when the primary containment airlock inner door seal was found degraded concurrent with the outer door being open. Further analysis by engineering of the actual conditions which were recorded when the airlock Type B leak rate test was being performed and calibration checks of the leak rate monitor instrumentation used for the test, has determined that the leakage through the degraded inner airlock door seal, when combined with the Appendix J As-Left Minimum Pathway Type B and C leak rates, remains below the Technical Specification Primary Containment As-Left Minimum Pathway Leakage Limit of 0.6 La. Therefore, for the period in which the inner airlock door seal was degraded and the outer airlock door was open, the primary containment function to control the release of radioactive material was maintained and the initial notification per 10 CFR 50.72(b)(3)(v) is being retracted. The licensee has notified the NRC Resident Inspector and the New York State Public Service Commission. R1DO (Ferdas) notified. |
| ENS 48883 | 3 April 2013 15:45:00 | A licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's unescorted access to the plant has been terminated. The licensee informed the NRC Resident Inspector. |
| ENS 49084 | 31 May 2013 15:07:00 | This 60-day telephone notification is being made per the reporting requirements specified in 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to describe an invalid actuation signal affecting containment isolation valves in more than one system. On April 2, 2013, Nine Mile Point 2 (NMP2) received a Division 2 reactor building area high ambient temperature isolation signal when lifting a lead for trip unit E31-N638B while performing surveillance N2-IPS-LDS-Q010, Reactor Building General Area Temperature Instrumentation Channel Functional Test. The isolation signal provided a closure signal to two Reactor Core Isolation Cooling System (RCIC) valves, and three Residual Heat Removal (RHR) system containment isolation valves. As a result of the isolation signal one of the RCIC containment isolation valves, 2ICS*MOV128 closed. The other four valves were already in their normal closed position. The RHR system valves are associated with the RHR Shutdown Cooling System and second RCIC isolation valve is used to warmup and place the RCIC system in standby following an isolated condition. All affected isolation valves responded as designed. As a result of 2ICS*MOV128 closing the RCIC system was declared inoperable. Technical Specification 3.5.3, RCIC System, Condition A was entered. Action A.1 required verifying the High Pressure Core Spray System (HPCS) was operable immediately. Action A.2 requires restoring RCIC to operable within 14 days. After the instrumentation system was restored to normal, the RCIC system was subsequently restored to available later that day at 1205 (EDT) and operable at 1500 (EDT). The actuation signal was not valid because it resulted from maintenance activities when leads were lifted, and the trip unit had not been bypassed as required by the procedure. There were no isolation logic signals in response to actual plant conditions or parameters. This event was entered into the corrective action system as Condition Report (CR) 2013-002461. There were no actual safety consequences or impact on the health and safety of the public as a result of this event. The licensee notified the NRC Resident Inspector and the State. |
| ENS 44923 | 21 March 2009 07:54:00 | At 0121 on Saturday, March 21, 2009, Nine Mile Point Unit One was manually scrammed from approximately 18% rated power in preparation for Refueling Outage 20. The plant scram was a planned evolution. Following the manual scram insertion at 0121, the High Pressure Coolant Injection (HPCI) System automatically initiated on low Reactor Pressure Vessel (RPV) level. At 0123, RPV level was restored above the HPCI System low level actuation setpoint, and the HPCI System initiation signal was reset. At Nine Mile Point Unit One the HPCI system is a mode of operation of the feedwater and condensate system. It is not an Emergency Core Cooling System (ECCS). A HPCI System actuation signal on low RPV level is normally received following a reactor scram, due to level shrink. 10 CFR 50.72(b)(3)(iv)(A) requires reporting within 8 hours when a valid actuation of the feedwater coolant injection system occurs. The event has been entered into the corrective action program. There are no other adverse impacts to the station based on this event. All control rods fully inserted. Electrical lineup is normal and there are no safety equipment out of service. The licensee has notified the NRC Resident Inspector
Event notification 44923 is being retracted after further evaluation because it has been determined that the High Pressure Coolant Injection (HPCI) system initiation was part of a pre-planned sequence. 10 CFR 50.72(b)(3) requires that any event or condition that results in valid actuation of any of the systems listed in this section, including a feedwater injection system, must be reported within 8 hours. Except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The HPCI actuation that occurred after the manual scram from 18% Rated Thermal Power (RTP) was part of a pre-planned sequence in accordance with an operating procedure on plant shutdown. The HPCI initiation was also discussed during Just-in-Time training for the operating crew prior to performance of the soft shutdown. At Nine Mile Point Unit 1, the HPCI system is a mode of operation of the feedwater and condensate systems. The HPCI system is not an Emergency Core Cooling System (ECCS). A HPCI system actuation signal on low reactor water level is normally received following a reactor scram, due to reactor water level shrink. The licensee will notify the NRC Resident Inspector. Notified R1DO (Holody). |
| ENS 44206 | 13 May 2008 08:59:00 | At 0806 EDT, a loss of offsite power occurred when the 115 KV line #4 was lost combined with the unavailability of 115 KV line #1 which was out of service at the time for planned maintenance. Both EDGs started and loaded on the safety busses as expected. The licensee declared an Unusual Event at 0825 EDT per EAL 6.1.1 based on loss of offsite power for greater than 15 minutes. The Unit 1 recirc pump #13 tripped as a result of the event and spent fuel pool cooling was also lost. The licensee remained at power and there was no impact on Nine Mile Unit 2. Power for 115 KV line #1 was restored at 0845 EDT. Cause of the loss of Line #4 is still under investigation. The reactor is stable and the safety busses continue to be power by the EDGs. Restoration of a normal electrical power lineup and termination of the UE is still under review. The licensee notified the NRC Resident Inspector, the State, and local authorities.
The licensee terminated the Unusual Event at 10:22 EDT based on restoration of offsite power to safety bus 103 from line #1. Bus 102 still remains energized by the EDG at this time. Investigation into the loss of line #4 continues. The loss of line #4 has been attributed to a Fitzpatrick protective relay 87B actuation for the B phase differential current. Investigation continues. The NRC Resident has been notified by the licensee. NRR EO (Thorp), Rg 1 (Dentel), IRD (McDermott), R1DO (Krohn), DHS (Dwight) and FEMA (McKentry) notified.
This condition is also reportable under 10 CFR 50.72(b)(3)(v)(D) 'Any event or condition that at the time of discovery could have prevented fulfillment of a safety function of systems that are needed to mitigate the consequences of an accident.' Specifically, both offsite power 115 KV lines were not available. The second paragraph which applies is 10 CFR 50.72(b)(3)(iv)(B) 'Any event or condition that results in the valid actuation of the systems listed in paragraph (b)(3)(iv)(B).' Specifically, the auto start of the Emergency Diesel Generators on a valid loss of offsite power signal. The cause of the loss of line 4 is still being investigated. Offsite power has been restored to onsite buses including PB 102, 103 and 101. The licensee will notify the NRC Resident Inspector of this update. Notified R1DO (Krohn). |