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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 53678 | 21 October 2018 05:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Shut Down Due to Chemistry Limit Exceeded | At 0200 Central Daylight Time on 10/21/2018, Browns Ferry Nuclear Plant Unit 3 commenced a reactor shutdown as required by the Technical Requirements Manual Limiting Condition for Operation 3.4.1 Coolant Chemistry Condition D due to conductivity greater than 10 micro mho/cm at 25 degrees Celsius. The required action for this condition is to immediately initiate an orderly shutdown and be in Mode 4 as rapidly as cooldown rate permits. This event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified.
ENS Event Number 53678, made on 10/21/18, is being retracted. NRC notification 53678 was made to ensure that the four-hour non-emergency reporting requirements of 10 CFR 50.72 were met when the licensee discovered a condition requiring shut down of a reactor. 10 CFR 50.72 requires a report in accordance with 50.72(b)(2)(i) for any Technical Specifications (TS) required reactor shutdown. NUREG-1022 only specifies TS applicability and makes no mention of a Technical Requirements Manual (TRM) required shutdown. Because the shutdown comes from the TRM and not the TS as discussed in 10 CFR 50.72 and NUREG-1022, an EN was not required. TVA's evaluation of this event notification is documented in the corrective action program. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ehrhardt). | ||
| ENS 51231 | 14 July 2015 22:15:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown | At 1810 (Central Daylight Time) on July 14, 2015, Browns Ferry Units 1, 2, and 3 initiated actions to commence a reactor shutdown to comply with TS (Technical Specifications) LCO 3.0.3. TS LCO 3.0.3 was entered at 1715 (Central Daylight Time) due to concurrent losses of the A and B Control Bay Chillers. This resulted in a loss of cooling to the U1 and U2 4kV Shutdown Board Rooms. Required actions for the loss of cooling to the U1 and U2 4kV Shutdown Board Rooms are to declare the electrical equipment in the 4kV Shutdown Board Rooms inoperable. The declaration of inoperability of the equipment supported by the U1 and U2 4kV Shutdown Boards resulted in TS LCO 3.0.3 for Units 1, 2, and 3. TS LCO 3.0.3 requires actions to be initiated within one hour to place the affected units in MODE 2 within 10 hours; MODE 3 within 13 hours; and MODE 4 within 37 hours. This event requires a 4-hour report in accordance with 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. Condition Report #1056829 has been initiated in the Corrective Action Program. The 4kV shutdown electrical boards are required in all modes of operation. | 05000259/LER-2015-003 | |
| ENS 50265 | 10 July 2014 09:45:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown | At 0445 (CDT) on July 10, 2014, Browns Ferry Unit 2 initiated actions to commence a reactor shutdown to comply with TS LCO 3.0.3. TS LCO 3.0.3 was entered at 0355 (CDT) and was required due to the 'C' Emergency Diesel Generator becoming inoperable after isolating a leak on the Emergency Equipment Cooling Water System. Currently, a 7 day TS LCO Action 3.5.1.A is in effect due to ongoing scheduled Core Spray Loop I maintenance outage. The declaration of inoperability of the equipment supported by the 'C' Emergency Diesel Generator, Core Spray Loop II, along with the redundant Core Spray system inoperable for maintenance resulted in TS LCO 3.0.3 for Unit 2. TS LCO 3.0.3 requires actions to be initiated within one hour; to place the unit in MODE 2 within 10 hours; MODE 3 within 13 hours; and MODE 4 within 37 hours. This event requires a 4 hour report lAW 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' Actions were taken to restore the Core Spray System to Operable status and LCO 3.0.3 was exited at 0735 (CDT) on July 10, 2014. The NRC Resident Inspector has been notified. This event was entered into the Corrective Action Program. Browns Ferry Unit 2 had reduced power to 98% when LCO 3.0.3 was exited, the power reduction was suspended, and preparations are being made to return power to 100%. There is no impact on Units 1 or 3. | Emergency Diesel Generator Core Spray Emergency Equipment Cooling Water | |
| ENS 50207 | 17 June 2014 22:29:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Shutdown Due to Loss of a and B Control Bay Chiller Trains | At 1729 (CDT) on June 17, 2014 Browns Ferry Units 1, 2, and 3 initiated actions to commence a reactor shutdown to comply with TS LCO (Technical Specifications Limiting Condition of Operation) 3.0.3. TS LCO 3.0.3 was entered at 1632 and was required due to the concurrent loss of 'A' and 'B' Control Bay Chiller units which resulted in loss of cooling to the U1 and U2 4kV Shutdown Board Rooms. Required actions for the loss of cooling to the U1 and U2 4kV Shutdown Board Rooms is to declare the electrical equipment in the 4kV Shutdown Board Rooms inoperable. The declaration of inoperability of the equipment supported by the U1 and U2 4kV Shutdown Boards resulted in TS LCO 3.0.3 for Units 1, 2 and 3. TS LCO 3.0.3 requires actions to be initiated within one hour to place the units in MODE 2 within 10 hours; MODE 3 within 13 hours; and MODE 4 within 37 hours. This event requires a 4 hour report IAW 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. Service Request 899734 was initiated in the Corrective Action Program. The 'B' Control Bay Chiller train was operating when the control room entered the Loss of Chilling procedure. The 'A' Control Bay Chiller train was started to supply Control Bay cooling, and the 'A' train unit tripped during start. The failure affected unit-3 because each unit has a 50% capacity common standby gas treatment system. The loss of the the U1 and U2 4kV shutdown boards failed two of the three systems.
"'A' Control Bay Chiller declared operable at 2035 CDT. TS LCO 3.0.3 exited and plant shutdowns have been secured." The licensee will be notifying the NRC Resident Inspector. Notified R2DO (Haag) | Standby Gas Treatment System | |
| ENS 48822 | 14 March 2013 13:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Unit 2 Shutdown Required by Technical Specifications Due to Inoperable Rcic | At 0300 (CDT) on 3/02/2013, radiography results for the U2 (Unit 2) RCIC (Reactor Core Isolation Cooling) system stop-check valve 2-HCV-71-14 showed this valve to be in the fully open position thus, not meeting the acceptance criteria of 2-SI-3.2.3 (Testing ASME Section XI Check Valves). This valve is the RCIC turbine exhaust stop-check valve. It is one of two PCIVs (Primary Containment Isolation Valves) connecting RCIC directly to primary containment. TS (Technical Specification) 3.6.1.3 (PCIVs) Condition 'A' was entered. In order to comply with TS 3.6.1.3 Condition 'A', the RCIC system was required to be isolated to prevent flow through the RCIC turbine exhaust check valve 2-CKV-71-0580. This valve is the upstream RCIC turbine exhaust line PCIV. 2-FCV-71-003, RCIC Steam Line Outboard Isolation valve, was closed and deactivated to meet TS 3.6.1.3 (PCIV) Condition 'A' at 0525 (CDT) on 03/02/2013. This resulted in RCIC being unable to auto initiate and required U2 to enter TS 3.5.3 (RCIC System) Condition 'A', which requires RCIC to be returned to service within 14 days. Because valve 2-HCV-71-14 is unisolable to primary containment, a unit S/D (Shutdown) is required to perform repairs. The RCIC TS LCO (Limiting Condition for Operation) 3.5.3 requires that U2 be in MODE 3 by 1725 (CDT) on 03/16/2013 and <150psig by 1725 (CDT) on 03/17/2013. Action to initiate Reactor S/D was taken on 03/14/2013 at 0800 (CDT). The licensee informed the NRC Resident Inspector. | Primary containment | |
| ENS 48782 | 25 February 2013 19:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Automatic Scram Due to a Turbine Trip from a Loss of Condenser Vacuum | At 1313 (CST) on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 (CST) with positive indication of isolation, and isolation valves deactivated at time 1514 (CST) to satisfy TS LCO 3.6.1.3 required actions. This event is reportable within 4 hours per 10CFR50.72(b)(2)( iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). At 1415 (CST), Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 (CST), water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2. | Feedwater High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Residual Heat Removal Emergency Core Cooling System Safety Relief Valve Control Rod | |
| ENS 47130 | 6 August 2011 11:17:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Initiation of Technical Specification Required Shutdown | On August 6, 2011, Reactor Protection System (RPS) power supply 1B failed resulting in a partial loss of power to Primary Containment Isolation System (PCIS) groups and an invalid actuation of those PCIS groups. PCIS groups 1 and 2 received partial isolation signals with no subsequent system isolations, as designed. PCIS group 3, 6, and 8 received partial isolation signals with resulting system isolations, also as designed. The combination of loss of RPS 1B and PCIS group 6 isolation resulted in the isolation of the Drywell Floor Drain Sump and the Drywell Continuous Atmospheric Monitor for both particulate and gaseous activity. Thus, both means of automatic monitoring of Reactor Coolant System leakage became inoperable. Unit 1 entered Technical Specification Limiting Condition for Operation (LCO) 3.4.5.D (all required leakage detection systems inoperable) and immediately entered LCO 3.0.3 as required. At the time of occurrence, RPS 1A was being supplied from its alternate source for scheduled maintenance. Thus, the alternate source was not available to RPS 1B. Unit 1 entered LCO 3.0.3 at 0524 (CDT), 'Initiate actions within one hour to place the unit in MODE 2 within 10 hours; MODE 3 within 13 hours; and MODE 4 within 37 hours.' At 0617, Unit 1 began reducing reactor power to comply with LCO 3.0.3. This event requires a 4 hour report IAW 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The PCIS isolations which occurred at 0524 CDT are also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) 'Any event or condition that results in a valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B)(2), 'General Containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs)), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The event time for the PCIS isolations is 0524 CDT. The NRC resident inspector has been notified. Service Request 412927 was initiated in the Corrective Action Program.
Browns Ferry restored power to the 1B Reactor Protection System power supply at 1208 CDT, reset all isolations and exited LCO 3.0.3. The licensee plans to return the unit to full power. The licensee notified the NRC Resident Inspector. Notified R2DO (Binoy Desai). | Reactor Coolant System Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Drywell Continuous Atmospheric Monitor | |
| ENS 46713 | 30 March 2011 20:42:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Initiated | On March 30, 2011 at 1443 CDT, during a refueling outage, Browns Ferry Unit 2 received an invalid Common Accident Signal (CAS) as a result of maintenance activities. All four Unit 1/2 diesel (generators) auto started and all four Unit 3 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 RHR (Residual Heat Removal) 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 was tagged out for maintenance. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) received auto initiation signals, however the steam isolation valves were tagged closed and the systems did not start. MSIVs (Main Steam Isolation Valves) isolated as a result of the CAS signal. Unit 1 was at 100% power when the CAS on Unit 2 occurred. This caused initiation of the Preferred Pump Logic which designates Division I CS (Core Spray) and LPCI (Low Pressure Coolant Injection) pumps on Unit 1 (1A and 1C) and the Division II pumps on Unit 2 (2B and 2D) and prevents Unit 1 Division II pumps from auto starting. This resulted in the Division II RHR and CS pumps (being) inoperable for Unit 1. Unit 1 subsequently entered Technical Specification 3.5.1, Condition H (two or more low pressure ECCS (Emergency Core Cooling System) injection/spray subsystems inoperable) which requires entering LCO 3.0.3 immediately. Unit 1 entered LCO 3.0.3 at 1443, which requires that actions shall be initiated within one hour to place the unit in Mode 2 within 10 hours; Mode 3 within 13 hours; and Mode 4 within 37 hours. At 1542 Unit 1 began lowering power in order to comply with LCO 3.0.3. CAS logic was reset at 1812 and Unit 1 exited LCOs 3.5.1.H and 3.0.3. This condition requires a four hour report in accordance with 50.72(b)(2)(i) - 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector was notified. Service Request #346544 was initiated in the Corrective Action Program. | Shutdown Cooling Core Spray | |
| ENS 45728 | 26 February 2010 01:35:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Rhr Inoperability | At 1840 on 2/25/10, Unit 2 RHR Division II became inoperable due to exceeding 260 degrees F on the injection line. The temperature is elevated due to leakage past the injection line check valve (2-CKV-74-68) and the inboard injection valve (2-FCV-74-67). The 260 degree F value is based on engineering calculations to ensure the injection line does not become steam voided. RHR Division I was already inoperable for scheduled maintenance when this condition occurred. Entered TS LCO 3.0.3 based on TS LCO 3.5.1 Condition H - Two or more Low pressure ECCS injection/spray subsystems inoperable for reasons other than Condition A. Action was taken to lower power at 1935 when the conditions requiring entry into LCO 3.0.3 could not be corrected. Tech Spec LCO 3.0.3 was exited at 2000 when RHR Loop II system pressure was raised using an alternate keep fill flow path to ensure that the injection line would remain filled. Power will be returned to 100%. This event requires a 4 hour report in accordance with 10 CFR 50.72(b)(2)(i). The licensee informed the NRC Resident Inspector. | ||
| ENS 45124 | 11 June 2009 20:55:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Unit 2 Initiated a Tech Spec Required Shutdown Due to a Rise in Unidentified Drywell Leakage | At noon on 06/11/2009 Browns Ferry Unit 2 experienced a rise in drywell leakage during reactor startup. The 4 hour unidentified leak rate from 0800 to 1200 on 6/10/2009 was 0 GPM, while the 4 hour unidentified leak rate from 0800 to 1200 on 6/11/2009 was 3.88 GPM. This increase in leak rate exceeded the allowable limit of a 2 GPM increase in unidentified leakage in a 24 hour period per Technical Specification 3.4.4. At 1555 (CDT) on 06/11/2009, Unit 2 inserted a manual reactor SCRAM to comply with Technical Specification 3.4.4 Condition C to be in Mode 3 in 12 hours and Mode 4 within 36 hours. All systems responded as required except that when reset of the SCRAM was attempted after all control rods had inserted, a portion of the 'B' RPS channel (B2/B3) failed to reset. The 'A' RPS channel was then actuated by a spike on the Intermediate Range Monitors while they were being driven into the core resulting in a full RPS SCRAM actuation. The licensee notified the NRC Resident Inspector." Unit 2 is currently removing decay heat using normal feedwater with bypass steam to the main condenser. The electrical system is in a normal shutdown configuration. The licensee has not identified the source of the unidentified leakage and anticipates taking Unit 2 to Cold Shutdown (Mode 4).
Plant personnel entered the drywell after the reactor was shutdown. The resulting investigation revealed two sources of increased drywell leakage. One source of leakage was from 2-CKV-10-511 and 2-CKV-10-526, SRV Tailpipe Vacuum Breakers, which were apparently damaged by leakage past the pilot valve for 2-PCV-001-0023, Main Steam Safety Relief Valve. The combination of damaged vacuum breakers and a leaking pilot valve resulted in steam leakage into the drywell atmosphere. The other source of leakage was noted through the packing for 2-DRV-10-505, RPV Drain to RWCU. The vacuum breakers, pilot valve, and drain valve have been repaired. The 'B' RPS contactors were repaired and the scram reset. The NRC Resident Inspector has been notified." R2DO (McCoy) notified. | Feedwater Intermediate Range Monitor Safety Relief Valve Main Condenser Control Rod Main Steam | |
| ENS 41861 | 22 July 2005 08:16:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | T/S Shutdown Due to Dc Subsystem Declared Inoperable | During performance of 2-SR-3.8.6.2(I-A) QUARTERLY CHECK FOR SHUTDOWN BOARD A BATTERY, the DC subsystem was declared inoperable due to cell voltage not meeting Technical Specification (TS) acceptance criteria. Concurrently a second DC subsystem was inoperable in support of scheduled maintenance activities. As a result of these subsystems being inoperable, Unit 2 entered TECHNICAL SPECIFICATION LCO 3.0.3. at 0220 and commenced a shutdown at 0316. In accordance with 10CFR50.72(b)(2)(i) this notification is being made. Plant conditions are stable at this time and no other plant systems were affected. At 0450 one DC subsystem was returned to an operable status. As a result of this, Unit 2 exited LCO 3.0.3, shutdown activities were terminated and unit 2 has been returned to 100 % power. The licensee notified the NRC Resident Inspector. |