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| Entered date | Event description | |
|---|---|---|
| ENS 46879 | 24 May 2011 17:56:00 | This 8-hour notification is being made per the reporting requirements specified by 10 CFR 50.72(b)(3)xiii. At 0946 CDT on May 24, 2011, the TVA Corporate Operations Duty Specialist notified the Nuclear Power Group Emergency Duty Officer that the fifteen minute communication test for Lawrence County indicated a loss of communications with the Lawrence County siren activation system. At 0958 CDT, Lawrence County Alabama Emergency Management Agency (EMA) staff reported that a transformer had de-energized at 0918 CDT and that both the primary siren activation point (Lawrence County EMA office) and the backup activation point (Moulton, Alabama Fire Department) were without power. The EMA offices normally have backup power supplied by a diesel generator. However, the generator had failed during EMA operations associated with the April 27, 2011 tornadoes. The State of Alabama had supplied a generator for temporary use but this generator required manual actions for making connections in order for it to be placed into service. At 1003 CDT, electrical power was restored and the primary and backup siren activation points were returned to service. Browns Ferry has 100 ANS sirens and 32 are located in Lawrence County. All of the 32 Lawrence County sirens were affected. The duration of the condition was estimated to be approximately 45 minutes (0918 CDT through 1003 CDT). (Note that the polling system that provides the out of service times polls the system every 15 minutes so the timeframe by necessity is an approximation.) Both primary and backup activation systems were operable at 1003 CDT. A silent test was performed from the Lawrence County EMA office at 1159 CDT with all 32 Lawrence County sirens satisfactorily responding. The licensee has notified the State of Alabama, the Lawrence County EMA and the NRC Resident Inspector of this report. |
| ENS 46847 | 13 May 2011 01:53:00 | At 1825 CDT on 05/12/2011, with Browns Ferry Nuclear Plant Unit 3 in Mode 4, Browns Ferry Nuclear Plant PCIS (Primary Containment Isolation System) relay maintenance activities for a Group 1 PCIS relay inadvertently interrupted the neutral for a Group 2 PCIS relay and a Group 2 PCIS isolation occurred resulting in a loss of Shutdown Cooling. Relays were reset and Shutdown Cooling was restored at 1905 CDT. Moderator temperature prior to the event was 112.5 degrees F and the highest moderator temperature recorded during the loss of Shutdown Cooling was 122 degrees F. This condition is reportable under 10CFR50.72(b)(3)(v) - Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: B.) Remove residual heat. This is also reportable as a 60 day written report IAW 10CFR 50.73(a)(2)(v). The NRC Resident Inspector has been notified of this event. This event was entered into the licensee's Corrective Action Program as SR# 368205". |
| ENS 46880 | 24 May 2011 18:03:00 | This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv) and 10 CFR 50.73(a)(1) to describe an invalid RPS (SCRAM) actuation. On March 30, 2011, at 1443 hours Central Daylight Time (CDT), during a refueling outage, Browns Ferry Unit 2 received an invalid Common Accident Signal (CAS) as a result of maintenance activities. The CAS caused a full Unit 2 Reactor SCRAM and associated system initiations. The CAS was initiated due to invalid indications on both Channels A and B of low-low-low reactor water level, which did not exist; therefore, the actuation was invalid. The affected equipment responded as designed. All four Unit 1/2 Emergency Diesel Generators auto started and all four Unit 3 Emergency Diesel Generators auto started. Unit 2 received a full Reactor SCRAM and Core Spray Pumps A, B, C, and D auto started and injected into the reactor. Unit 2 Division I Residual Heat Removal (RHR) System was in Shutdown Cooling with only the C pump in service. The 'A' RHR pump auto started and Shutdown Cooling flow increased, as expected. Unit 2 Division II RHR System had been tagged out for maintenance and did not respond. High Pressure Coolant Injection and Reactor Core Isolation Cooling received auto initiation signals; however, their steam isolation valves were tagged closed and the systems did not start. The Inboard Main Steam Isolation Valves (MSIVs) isolated as a result of the CAS signal. The outboard MSIVs had been previously closed and tagged for refueling outage purposes. This event was entered in the Corrective Action Program as Service Request (SR) 346544, which generated Problem Evaluation Report (PER) 346568. There were no safety consequences or impact on the health and safety of the public as a result of these events. The NRC Senior Resident Inspector has been notified. The shutdown reactor water level transmitters share a common variable leg. When maintenance unrelated to the transmitters was performed, the variable leg was lost causing the low-low-low reactor water level SCRAM signal to be generated. |
| ENS 46881 | 24 May 2011 18:09:00 | This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv) and 10 CFR 50.73(a)(1) to describe an invalid actuation of multiple secondary containment isolation signals affecting more than one system. On March 26, 2011, at 2005 hours Central Daylight Time (CDT), following planned maintenance activities, an unplanned actuation of secondary containment isolation signals affecting containment isolation valves in more than one system occurred and resulted in a partial Primary Containment Isolation System (PCIS) (Group 6, Secondary Containment) isolation and receipt of a Unit 2 Reactor Building Ventilation Abnormal Alarm. Unit 2 was in Mode 4, in a refueling outage, and at 0 percent power (0 MWT). Units 1 and 3 were both in Mode 1 and operating at approximately 100 percent power. Plant Conditions, which initiate PCIS (Group 6, Secondary Containment) actuations, are Low Reactor Vessel Water Level, High Drywell Pressure, High Reactor Zone Exhaust Radiation, or High Refueling Floor Exhaust Radiation. At the time of the event, these conditions did not exist; therefore, the partial actuations were invalid. For this occurrence, equipment normally affected by a complete Group 6 isolation responded as follows. Trains 'B' and 'C' of Standby Gas Treatment (SGT) System started while Train 'A' did not. Train 'B' of the Control Room Emergency Ventilation System (CREVS) started while Train 'A' did not. Secondary containment (Unit 1, 2, and 3 Reactor and Refuel Zones) normal ventilation isolated - fans stopped and dampers closed. Suppression Chamber Exhaust Inboard Isolation Valve 2-FSV-64-32 and Drywell Exhaust Inboard Isolation Valve 2-FCV-64-29 did not close. Because only a partial (B part of the logic) PCIS isolation was relayed, only the affected Group 6 equipment received the isolation demand. All equipment that received the isolation demand responded and performed as designed. Control Room personnel commenced actions of the applicable Alarm Response Procedure and the Group 6 Ventilation System Isolation Abnormal Operations Procedure. Train 'A' of the SGT System was later started. This event was entered in the Corrective Action Program as Service Request (SR) 344609 and Problem Evaluation Report (PER) 344680. There were no safety consequences or impact on the health and safety of the public as a result of these events. The NRC Resident Inspector has been notified of this event. Although Group 6 PCIS testing occurred earlier that day, the licensee could not determine the signal initiator and considered it spurious. |
| ENS 46179 | 16 August 2010 18:02:00 | Browns Ferry Nuclear Plant inadvertently exceeded a limitation of its National Pollutant Discharge Elimination System (NPDES) Permit for Tennessee River temperature conditions which was reported to the Alabama Department of Environmental Management (ADEM) at 1340 Central Daylight Time, on August 16, 2010. As such, this is being reported in accordance with 10 CFR 50.72(b)(2)(xi) as 'Any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made.' The NPDES Permit requires that when the 24 hour ambient average river temperature exceeded 90 degrees Fahrenheit, the 24 hour average downstream river temperature may equal but not exceed the upstream value. The 24 hour average upstream temperature was found to be 90.4 (rounded to 90) degrees Fahrenheit and the downstream temperature was found to be 90.6 (rounded to 91) degrees Fahrenheit. Browns Ferry had been maintaining compliance with the NPDES permit. The departure is thought to be due to a summer storm cooling off the upstream temperature. With the time delay between upstream temperature changes and the ensuing rate at which the downstream temperature changes are observed, the resultant 24 hour average downstream river temperature exceeded the upstream value. Duration of this condition was found to be only several hours. The three Browns Ferry units, which had already been derated to approximately 50 percent rated thermal power in order to maintain compliance with the NPDES Permit river temperature limits, were further derated to achieve compliance with the NPDES Permit. There is no corresponding requirement for written follow-up in accordance with 10 CFR 50.73. The licensee has notified the NRC Resident Inspector. |
| ENS 44367 | 27 July 2008 08:30:00 | Security notified Limestone County Sheriffs dispatch at 0340 CDT that Browns Ferry had experienced a fire and arc-over event at the area of cooling tower switchgear 'C'. At about 0225 CDT a fault occurred in the cable tray between the 161 kV supply transformer and the 4 kV cooling tower 'C' switchgear resulting in clearing of the 161 kV line and a subsequent fire. The fire was extinguished at 0236 CDT by on-site responders. At 0340 CDT, Limestone County Sheriff department was contacted by site security and informed of the event and asked to increase patrols in the area of the county road adjacent to Browns Ferry. It was known that the event was likely due to operational issues and not sabotage but the final determination had not been made at that time. Limestone County Sheriff was contacted again at 0520 CDT and informed that the problem was operational and that their assistance is not required. There are no safety related plant equipment reportability issues resulting from this event. The reportability is solely due to notification of an outside agency. This condition is reportable within 4 hours according to 10 CFR 50.72(b)(2)(xi). Any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactive contaminated materials. Licensee has notified NRC Resident Inspector. |
| ENS 44232 | 21 May 2008 21:41:00 | At 1515 hrs, the control room staff discovered that normal control room phone communications were not working. Additionally, the NRC emergency notification system (red phone) was not functional. The following communication systems were maintained: Satellite phone system, Chattanooga CECC ring down line, 2 lines in the Technical Support Center, Nextel mobile phone network, and site radio frequencies. Even though some communications capabilities were maintained, this condition was determined to be reportable as a major loss of communication capability via 10 CFR 50.72(b)(3)(xiii) given that normal control room phone lines and the NRC red phone were not available. All normal means of communication were returned to service at 1724 hrs by placing a spare battery charger in service to the communications system. The Browns Ferry NRC Resident Inspector was notified of this occurrence and restoration. |
| ENS 43940 | 29 January 2008 00:51:00 | On 1/28/08 at 1843 CST, Browns Ferry Unit 1 was performing 1-SR-3.3.5.1.3(D) HPCI System Condensate Header Low Level Switch Calibration and Functional Test when 1-LS-73-56A failed to actuate. Per TS 3.3.5.1, 1-LS-73-56A is inoperable. 1-SR-3.3.5.1.3(D) defeats the logic relay normally actuated by 73-56A & B. This causes HPCI to be inoperable per TS 3.3.5.1.D if the relay is defeated for greater than 1 hour. Failure of the 73-56A switch prevented restoration of the relay within the 1 hour time frame. This event is reportable under 10CFR 50.72(b)(3)(v)(B) 'any event or condition that at the time of discovery could have prevented the fulfillment of the Safety Function of structures or systems that are needed to: Remove Residual Heat' and 10CFR 50.72(b)(3)(v)(D) 'any event or condition that at the time of discovery could have prevented the fulfillment of the Safety Function of structures or systems that are needed to: mitigate the consequences of an accident.' This event also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(v)(B) and 10CFR 50.73(a)(2)(v)(D). The defeated relay was restored to normal and the HPCI system returned to operable status at 2330 CST on 1/28/08. The licensee notified the NRC Resident Inspector.
On January 28, 2008, Browns Ferry Unit 1 entered an LCO to perform a planned maintenance activity, High Pressure Coolant Injection System Condensate Header Low Level Switch Calibration and Functional Test, 1-SR-3.5.5.1.3(D). During the calibration of 1-LS-073-0056A and 1-LS-073- 0056B, 1-LS-073-0056A was found inoperable. The removal of both level switches from service (and as a result the HPCI transfer on low condensate header level function) was a planned maintenance activity, performed in accordance with an approved procedure and in accordance with the plants TSs. During this time no condition was discovered that could have prevented HPCI from performing its intended function because 1-LS-073-056B was considered OPERABLE. Therefore, this event is not reportable under 10 CFR 50.72(b)(3)(v)(B) or 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. R2DO(Lesser) notified. |