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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57006 | 5 March 2024 06:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 0132 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main feedwater isolation signal which resulted in steam generator lo-level reactor trip. The reactor trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the main feedwater isolation is being investigated. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 56970 | 16 February 2024 03:24:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of Emergency Diesel Generator System | The following information was provided by the licensee via email: At 2224 EST on February 15, 2024, with both units 1 and 2 in mode 1 at 100 percent power, an actuation of the emergency diesel generator (EDG) system on 1A-A, 1B-B, and 2B-B EDGs occurred while removing clearances. The 2A-A EDG did not start because it was still under a clearance. The reason for the emergency diesel generator system auto-start was clearance removal sequencing errors. The emergency diesel generator system automatically started as designed when the common emergency start signal was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the emergency diesel generator system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: In accordance with NUREG-1022, Section 2.8 and Section 4.2.3, Watts Barr is retracting the previous report EN 56970 pursuant to 10 CFR 50.72(b)(3)(iv)(A). The start signal for the 1A-A, 1B-B, and 2B-B emergency diesel generators (EDG)s was from activation of the common emergency start of the 2A-A EDG. The actuation was not from a loss of offsite power (LOOP) to any shutdown board or from any parameters that would initiate a safety injection (SI) signal, for which the EDG is designed to provide a design basis safety function. Also, the starts were not from intentional manual actuation. Starting the EDGs did not make them inoperable and each EDG was able to perform its design safety function. The common emergency start relay for each diesel is not safety related. It is an anticipatory and redundant circuit to start other EDGs in the event of a LOOP or SI related to the specific EDG. With the 2A-A EDG out of service, the associated common emergency circuit would not be required to perform any function. The starts were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the system. Since the starts were not initiated via an automatic signal from a LOOP, SI, or traditional operator action, the signal is not a valid actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). Therefore, EN 56970 is being retracted. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Miller) | Emergency Diesel Generator | |
| ENS 56935 | 28 January 2024 02:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 2141 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. The licensee notified the NRC Resident Inspector. | Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 56660 | 4 August 2023 21:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via phone and email: At 1746 EDT on 08/04/2023, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to number 2 steam generator low low level. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the number 2 steam generator low low level is being investigated. | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 56593 | 27 June 2023 20:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via phone and email: At 1626 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The (reactor) trip was not complex with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. | Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 55143 | 17 March 2021 15:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | At 1004 EDT on March 17, 2021, with Unit 2 in Mode 1 at 90 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the Auxiliary Feedwater and Steam Dump Systems. Unit 1 is not affected. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All controls rods fully inserted and the electrical system is in normal shutdown alignment. The cause of the turbine trip is being investigated. | Reactor Protection System Auxiliary Feedwater | |
| ENS 54997 | 15 November 2020 06:44:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of the Emergency Diesel Generators | At 0144 EST on November 15, 2020, with Unit 1 in Mode 1 at 100 percent power and Unit 2 in Mode 5 at 0 percent power, an actuation of the Emergency Diesel Generator (EDG) system occurred while transferring the 2A-A 6.9 kV Shutdown Board (SDBD) from the maintenance feed to its normal power supply. The reason for the 2A-A 6.9 kV SDBD failing to transfer to the normal power supply is under investigation. The EDGs automatically started as designed when the valid actuation signal was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the EDGs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 54720 | 20 May 2020 09:21:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Emergency Diesel Generators | At 0521 EDT on May 20, 2020, with Unit 2 in Mode 3 at 0 percent power and Unit 1 defueled, an actuation of the Emergency Diesel Generator (EDG) System occurred while transferring the 1B-B 6.9kV Shutdown Board (SDBD) from the maintenance feed to its normal power supply. The reason for the 1B-B 6.9kV SDBD failing to transfer to the normal power supply is under investigation. The EDGs automatically started as designed when a valid actuation signal was received. The event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the EDGs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 54532 | 19 February 2020 14:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Loss of Control of Steam | At 0936 EST on February 19, 2020, the Watts Bar Unit 1 reactor was manually tripped while operating at 100 percent power in response to loss of control of water level for steam generator number 3. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. There is no impact to Unit 2. The manual actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50. 72(b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. | Steam Generator Reactor Protection System Auxiliary Feedwater Decay Heat Removal | |
| ENS 54252 | 1 September 2019 00:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Trip Due to Loss of Steam Generator Level Control | While operating at 100 percent power, the Watts Bar Unit 1 reactor was manually tripped at 2055 EDT on August 31, 2019 due to loss of steam generator #2 level control. The trip was not complex. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems, including Auxiliary Feedwater, actuated as designed. Operations responded and stabilized the plant. Decay heat is being removed by the Auxiliary Feedwater and the Steam Dump System. Unit 2 is not affected. The cause of the loss of steam generator water level control is being investigated. The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The automatic actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 54077 | 22 May 2019 06:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Failure of Main Feedwater Regulating Valve | On May 22, 2019, at 0233 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 reactor was manually tripped due to a failure of the #2 Main Feedwater Regulating Valve during power ascension following a refueling outage. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Dumps. Unit 2 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). There was no impact to WBN Unit 1. The NRC Senior Resident has been notified. | Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 53697 | 27 October 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following Startup | On October 27, 2018, at 1533 EDT, Watts Bar Nuclear (WBN) Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. (Main Steam Isolation Valves) MSIVs were required to be isolated due to cooldown. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Generator Atmospheric Dump Valves. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident has been notified. | Steam Generator Auxiliary Feedwater Decay Heat Removal | |
| ENS 53557 | 22 August 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 0943 EDT on August 22, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The reactor automatically tripped due to a main turbine trip signal. An investigation is in progress. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. | Reactor Protection System Auxiliary Feedwater Decay Heat Removal | |
| ENS 53467 | 22 June 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 0841 EDT on June 22, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 95% power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The reactor automatically tripped due to a main turbine trip. The turbine trip was caused by main generator electrical trip. An investigation is in progress. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip. | Reactor Protection System Auxiliary Feedwater Decay Heat Removal | |
| ENS 53327 | 12 April 2018 13:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 0920 EDT on April 12, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The cause of the automatic reactor trip is being investigated. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip. | Reactor Protection System Auxiliary Feedwater Decay Heat Removal | |
| ENS 53132 | 20 December 2017 15:40:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | System Actuations Due to Opening of Feeder Breaker to Shutdown Board | On December 20, 2017, at 1040 Eastern Standard Time (EST), the Watts Bar Nuclear Plant (WBN) 1B-B 6.9kV Shutdown Board (SDBD) normal feeder breaker opened. The loss of voltage to the 1B-B SDBD resulted in the start of the 1B-B Motor Driven Auxiliary Feedwater (MDAFW) pump, the Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) pump, and the start of all four Emergency Diesel Generators (EDGs). Power was restored to the 1B-B 6.9 kV SDBD when it loaded on to its associated EDG. Following initial investigation, the 1B-B 6.9 kV SDBD was transferred to its alternate offsite power source, Common Station Service Transformer (CSST) C at 1217 EST. At 1230 EST, the 1B-B 6.9 kV SDBD alternate feeder breaker opened. The loss of voltage to the 1B-B SDBD did not result in the restart of the 1B MDAFW pump, the Unit 1 TDAFW pump, or EDGs; this equipment remained running from the earlier event. Power was restored to the 1B-B 6.9 kV SDBD when it loaded on to its associated EDG. Restoration of normal offsite power to the 1B-B SDBD was completed at 1654. Other than several common Unit Technical Specifications having not been met, Unit 2 was not operationally impacted by the transfer of the 1B-B Shutdown Board to onsite power and remains in Mode 1 at 100% power. This report is made per 10 CFR 50.72(b)(3)(iv)(A). NRC Resident Inspector has been notified. The licensee investigation continues for the cause of the event. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 53112 | 11 December 2017 13:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip in Response to Indication of Multiple Dropped Control Rods | While operating at 97% power, the Watts Bar Unit 2 reactor was manually tripped at 0857 EST on December 11, 2017 due to multiple dropped control rods. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and the Steam Dump System. The cause of the dropped rods is being investigated. The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. No safety or relief valves lifted during this event. | Reactor Protection System Auxiliary Feedwater Decay Heat Removal Control Rod | 05000391/LER-2017-006 |
| ENS 52912 | 17 August 2017 16:05:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Actuation of Turbine Driven Auxiliary Feedwater Pump Due to Loss of 6.9Kv Shutdown Board | On August 17, 2017, at 1205 Eastern Daylight Time (EDT), the Watts Bar Nuclear Plant (WBN) lost power to the 1B-B 6.9kV Shutdown Board. The loss of power to this safety related bus resulted in an automatic start of the Unit 1 Turbine Driven Auxiliary Feedwater Pump (TDAFWP). The motor driven AFW pumps did not start and were not expected to start for this event. No other system actuations occurred as a result of loss of power to the 1B-B 6.9kV Shutdown Board. Power to the 1B-B Shutdown Board was restored at 1505 EDT on August 17, 2017. During the loss of power to the 1B-B Shutdown Board, a reduction in containment and control rod drive mechanism cooling occurred. At 1233 EDT, lower containment average temperature exceeded the procedural limitations specified, and Technical Specification 3.6.5 Condition A was entered for containment average air temperature not within limits. Lower containment average temperature was restored to within limits at 1525 EDT on August 17, 2017. This is reportable as a potential loss of safety function. Other than several common Unit Technical Specifications having not been met, Unit 2 was not operationally impacted by the loss of power to the 1B-B Shutdown Board and remains in Mode 1 at 100% power. The NRC Resident Inspector has been notified. The loss of power occurred during restoration from Emergency Diesel Generator maintenance. | Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 52872 | 25 July 2017 08:28:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Rod Position Indication System Being Inoperable | On July 25, 2017, at 0428 Eastern Daylight Time (EDT) Watts Bar Nuclear Plant (WBN) Unit 2 was in Mode 3, beginning a Reactor Startup. While in the initial phase of withdrawing the first of four Control Rod banks, the two associated group demand position indicators deviated greater than 2 steps from each other. In accordance with Technical Requirement 3.1.7, Position Indication System, Shutdown, with one or more group demand position indicators inoperable, the reactor trip breakers are to be opened immediately. Operations personnel opened the reactor trip breakers immediately by initiating a manual trip of the Reactor Protection System (RPS). The Auxiliary Feedwater system was in service and controlling Steam Generator water levels at the time of the event and did not receive any valid actuation signals. No other system actuations occurred as a result of this reactor trip and all systems operated as designed. The cause of the position indication system inoperability is currently under investigation. NRC Resident Inspector has been notified. | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 52732 | 4 May 2017 21:09:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Failed Reactor Coolant Pump Power Transfer | On May 4th, 2017, at 1709 EDT, Watts Bar Nuclear Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Generator Atmospheric Dump Valves. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72(b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident (Inspector) has been notified. | Steam Generator Auxiliary Feedwater Decay Heat Removal | |
| ENS 52725 | 2 May 2017 23:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Failed Reactor Coolant Pump Power Transfer | On May 2nd, 2017, at 1945 EDT, Watts Bar Nuclear (WBN) Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All control and shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via auxiliary feedwater and main steam dump systems. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident (Inspector) has been notified. | Auxiliary Feedwater Decay Heat Removal Main Steam | |
| ENS 52630 | 23 March 2017 04:14:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of Auxiliary Feed Water | On March 23, 2017, at 0014 EDT, Watts Bar Nuclear Plant Unit 2 (WBN2) experienced an unplanned trip of both Turbine Driven Main Feed Pumps (TDMFP) following a loss of Main Condenser Vacuum. The trip of both TDMFPs caused an automatic start of both Motor Driven Auxiliary Feed Water Pumps and the Turbine Driven Auxiliary Feed Water Pump. (The) cause of the loss of Main Condenser Vacuum is currently under investigation. The plant was performing a normal startup, and had just synced the main generator to the grid. Subsequent to the event, the plant was transitioned to Mode 3. All rods are fully inserted. Decay heat is being removed via the atmospheric relief valves. Unit 1 remains in Mode 5 for a refueling outage. The licensee has notified the NRC Resident Inspector. | Auxiliary Feedwater Turbine Driven Main Feed Pump Main Condenser | |
| ENS 52625 | 20 March 2017 12:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip as a Result of Secondary Plant Transient | On March 20, 2017 at 0813 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 operations personnel manually tripped the plant from approximately 91 percent power based on lowering steam generator levels. Prior to the plant trip, the 2A Hotwell pump tripped at 0758 EDT and the 2C Condensate Booster Pump subsequently tripped at 0802 EDT. Operations personnel commenced to lower plant power after the 2A Hotwell pump trip in an attempt to maintain steam generator levels, but were unable to recover level and manually tripped the unit. All control rods fully inserted and all automatically actuated safety related equipment operated as designed. At 0905 EDT, operations personnel exited the emergency operating instructions after the plant was stabilized. The cause of the event is under investigation. This event is reportable to the NRC within four hours under 10 CFR 50.72(b)(2)(iv)(B) as a result of the actuation of the Reactor Protection System and in eight hours under 10 CFR 50.72(b)(3)(iv)(A) as a result of actuation of the Auxiliary Feedwater system. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 52210 | 31 August 2016 01:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Notice of Unusual Event - Fire in a Main Bank Transformer | A fault occurred on the unit 2 "B" main bank transformer resulting in an oil fire. The main turbine tripped resulting in a reactor trip. All control rods fully inserted and no safety or relief valves lifted. Decay heat is being removed via steam dumps to the main condenser and feeding steam generators with auxiliary feedwater. Electrical power is through the normal shutdown electrical lineup. Offsite assistance was requested from the county and off duty fire brigade members. At 2228, the fire was reported as out. Spray is continuing and a reflash watch is being set. Unit 1 continued to operate at 100% power throughout the event. Notified the DHS SWO, FEMA Ops Center, DHS NICC, FEMA National Watch Center (E-mail) and Nuclear SSA (E-mail).
On August 30, 2016, at 2110 EDT, Watts Bar Nuclear Plant Unit 2 reactor tripped due to an electrical fault affecting the 2B Main Bank Transformer, resulting in a fire in the transformer. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and main steam dump systems. Unit 2 is in a normal shutdown electrical alignment. The fire was out at 2230 EDT. The cause of the fire is currently under investigation. The fire was reported at 2149 EDT. Local Fire Departments responded to the site as requested. The reactor trip and system actuation is being reported under 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 1. The NOUE was exited at 2342. The NRC Senior Resident Inspector has been notified. The Licensee notified the State of Tennessee. Notified the R2DO (Bartley), IRD MOC (Stapleton), NRR EO (Miller), DHS SWO, FEMA Ops Center, DHS NICC, FEMA National Watch Center (E-mail) and Nuclear SSA (E-mail). | Steam Generator Auxiliary Feedwater Main Turbine Decay Heat Removal Main Condenser Control Rod Main Steam | |
| ENS 52194 | 23 August 2016 17:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Trip Due to a Loss of Main Feedwater | On August 23, 2016, at 1356 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 reactor was manually tripped due to a loss of main feedwater. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All control and shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and main steam dump systems. Unit 2 is in a normal shutdown electrical alignment. The cause is currently under investigation. This is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72(b)(2)(iv)(B). There was no effect on WBN Unit 1. The NRC Senior Resident Inspector has been notified. | Feedwater Auxiliary Feedwater Decay Heat Removal Main Steam | |
| ENS 52177 | 13 August 2016 07:36:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of Turbine Driven Auxiliary Feed Water Pump | On August 13, 2016 at 0330 EDT, Watts Bar Nuclear Plant Unit 2 (WBN2) was being stabilized following a pre-planned reactor trip. Both motor driven auxiliary feed water pumps and the turbine driven auxiliary feed water pump (TDAFW pump) were in operation maintaining steam generator water level 6 - 50 percent in accordance with the Reactor Trip Response Procedure. At 0333 EDT the TDAFW pump was secured by procedure and steam generator water level lowered to the Low Low Alarm setpoint (17 percent). The trip time delay at 0 percent power is three (3) minutes. At 0336 EDT, the TDAFW pump automatically started with steam generator water levels less than the Low Low alarm setpoint (lowest level reached was 15 percent). Steam generator water level was restored to the normal shutdown control band (38 percent). Unit 1 remains in Mode 1 at 85 percent. Unit 2 is stable in Mode 3 with normal shutdown power alignment. The reactor coolant system is being cooled down to 400 degrees F for a planned maintenance period, with decay heat removal via the Main Condenser Steam dumps and the AFW system. This event is being reported pursuant to 10 CFR 50.72(b)(3)(iv)(A). TVA has notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Decay Heat Removal Main Condenser Auxiliary Feed Water | |
| ENS 52026 | 20 June 2016 19:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | On June 20, 2016 at 1540 EDT, Watts Bar Nuclear Plant Unit 2 reactor tripped due to (reaching the) automatic Lo-Lo steam generator trip (setpoint) on (the) #4 steam generator. Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. All control and shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and main steam dump systems. The station is in a normal shutdown electrical alignment. The cause is currently under investigation. This is being reported under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72 (b)(2)(iv)(B). The NRC Senior Resident has been notified. There was no effect on Unit 1. | Steam Generator Auxiliary Feedwater Decay Heat Removal Main Steam | |
| ENS 51982 | 5 June 2016 16:27:00 | 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)(iv)(A), System Actuation - ECCS Discharge | Reactor Trip and Eccs Actuation Caused by Turbine Governor Valve Failure | On June 5, 2016 at 1227 Eastern Daylight Time (EDT), Watts Bar Nuclear Plant (WBN) Unit 2 was in MODE 1 at approximately 12.5% power when a safety injection actuation occurred, followed by an automatic reactor protection system (RPS) trip. Preliminary data suggests that the #1 high pressure turbine governor valve failed open causing a steam header pressure rate of decrease safety injection (SI) actuation signal. As designed, the safety injection actuation caused both trains of the shared Emergency Gas Treatment System (EGTS) to align to Unit 2, requiring WBN Unit 1 to enter Technical Specification (TS) LCO 3.0.3 at 1227 (EDT). Also as designed, Unit 1 annulus pressure momentarily increased, causing operation personnel to enter LCO 3.6.15 Condition B at 1240 (EDT). At 1242 (EDT), after annulus pressure normalized, operations personnel exited LCO 3.6.15. At 1245 (EDT), Operations personnel secured safety injection and Unit 2 was stabilized in MODE 3 at normal operating pressure and temperature. By 1349 (EDT), Unit 1 Operations personnel had restored both trains of EGTS to standby readiness, and exited TS 3.0.3. No primary safety barriers (RCS, containment and fuel clad) were challenged and no primary or secondary safety or relief valves actuated during the event. The Unit 2 plant trip was uncomplicated and safety equipment operated as expected. Unit 2 is stable in Mode 3 at normal operating temperature and pressure, in normal shutdown power alignment. Unit 1 is stable in Mode 1 at 100% power. The Senior Resident Inspector has been notified of this event. The MSIVs are shut with the steam generators (SG) discharging steam using the atmospheric dump valves. There is no primary to secondary leakage. Motor driven AFW pumps are running to maintain SG levels. RCS pressure remained above ECCS (Emergency Core Cooling System) discharge pressure. | Steam Generator Reactor Protection System Emergency Gas Treatment System Emergency Core Cooling System | 05000391/LER-2016-004 |
| ENS 51940 | 17 May 2016 20:30:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Loss of 1 B-B Electrical Buss | On May 17, 2016, at 1630 hours while restoring from a plant modification related to new 'loss of phase' circuitry, the 1 B-B 6.9kV buss de-energized resulting in a loss of voltage on the buss. The loss of voltage was caused by the loss of voltage relays that separated offsite power from the 1 B-B 6.9kV buss. At the time, the 1 B-B emergency diesel generator was removed from service for planned maintenance. In response to the loss of power on the 1 B-B 6.9kV buss, the operators entered abnormal operating instruction, AOI 43.02, Loss of Unit 1 Train B Shutdown Boards, and started emergency diesel generators 1 A-A, 2 A-A, and 2 B-B. All equipment operated properly. The emergency diesel generators were not required to be paralleled to the boards. Offsite power was restored to the 1 B-B 6.9kV buss at 1802 hours on May 17, 2016. This condition did not result in any adverse impact on the health and safety of the public. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' The NRC Resident Inspector has been notified. | Emergency Diesel Generator | 05000390/LER-2016-008 |
| ENS 51815 | 22 March 2016 15:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Watts Bar Automatic Reactor Trip | On March 22, 2016, at 1131 EDT, the Watts Bar Nuclear Plant Unit 1 (WBN1) reactor tripped due to the actuation of the Over Temperature Delta Temperature bistables. Concurrent with the reactor trip, the Auxiliary Feedwater (AFW) system actuated. All control rods inserted upon the reactor trip. Engineered Safety Function systems functioned as expected. WBN1 is currently stable in Mode 3. The Reactor Coolant system is being maintained at normal operating pressure and no-load temperature with decay heat removal via the Main Condenser steam dumps and the AFW system. The station is in a normal shutdown electrical alignment. Watts Bar Unit 2 is in mode 4 preparing for power ascension testing and was unaffected by the WBN1 event. The cause of the event is under investigation. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. | Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | 05000390/LER-2016-004 |
| ENS 50839 | 21 February 2015 15:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Inserted Due to Rapid Loss of Main Condenser Vacuum | On February 21, 2015 at 10:32 EST, Watts Bar Nuclear Plant Unit 1 reactor was manually tripped due to rapidly dropping main condenser vacuum. Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and S/G PORVs. Main Steam Isolation Valves are closed. The Station is in a normal shutdown electrical alignment. The cause is currently under investigation. This is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). The NRC Senior Resident has been notified. This event had no affect on Unit 2 (Under Construction) | Main Steam Isolation Valve Auxiliary Feedwater Decay Heat Removal Main Condenser | 05000390/LER-2015-001 |
| ENS 50278 | 13 July 2014 23:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Automatic Feedwater Heater Isolation | On July 13, 2014 at 1937 (EDT), Watts Bar Nuclear Plant Unit 1 reactor was manually tripped due to automatic isolation of all low pressure feedwater heaters. Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater, steam dumps and the main condenser. The station is in a normal shutdown electrical alignment. The cause is currently under investigation. This is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). The NRC Senior Resident (Inspector) has been notified. Initial indications are that a level switch failed on the #7 heater drain tank which caused the heater drain tank pumps to trip. When the high level tank setpoint was reached, the feedwater system isolated. | Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 49154 | 28 June 2013 17:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Offsite Electrical Fault | On June 28, 2013 at 1330 (EDT), Watts Bar Nuclear Plant Unit 1 reactor automatically tripped due to an electrical fault causing a main generator lockout and subsequent turbine trip. The electrical fault generated an 'A' Main Bank Transformer Differential Relay actuation. Suspected cause is due to an offsite fault. This is being reported under 10CFR50.72(b)(2)(iv)(B). Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. This is being reported under 10CFR50.72(b)(3)(iv)(A). All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater, steam dumps and the main condenser. The station is in a normal shutdown electrical alignment. The NRC Senior Resident has been notified. There were no primary or secondary relief valve actuations during the plant transient. The cause of the electrical fault is being investigated. | Auxiliary Feedwater Decay Heat Removal Main Condenser | |
| ENS 48414 | 17 October 2012 03:30:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | All Four Emergency Diesel Generators Auto-Start on Valid Undervoltage Signal | On October 16, 2012, Watts Bar Nuclear Plant Unit 1 experienced a momentary loss of power to the 1B 6.9 KV Shutdown board. This resulted in an emergency start of all 4 Emergency Diesel Generators due to actuation of the designed undervoltage EDG start signal. A manual transfer of the shutdown board from the Normal power supply to the Alternate was in progress to support maintenance. The momentary loss of power lasted approximately 2 seconds, which caused load shed and Diesel Generator start to occur. Core cooling was maintained throughout the transient. This condition is being reported under 10 CFR 50.72 (b)(3) (iv) (A). All safety systems responded as designed. The unit remains stable in Mode 5. The cause for this failure is not yet known. Investigation is in progress. The NRC Senior Resident has been notified. | Emergency Diesel Generator | |
| ENS 48238 | 28 August 2012 07:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Following Failure of a Feedwater Regulating Valve | On August 28, 2012, Watts Bar Nuclear Plant Unit 1 reactor automatically tripped due to low level in steam generator (SG) #2. The low level resulted when the Main Feedwater Control Valve for SG#2 (1-FCV-3-48) failed closed. This is being reported under 10CFR 50.72 (b)(2) (iv) (B). Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. This is being reported under 10CFR 50.72(b)(3) (iv)(A). All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3. The NRC Senior Resident has been notified. Decay heat is being removed to the main condenser via condenser steam dumps. The plant is in its normal shutdown electrical lineup. No steam safety or relief valves lifted during the event. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser | |
| ENS 46902 | 29 May 2011 05:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | At 0155 EDT, the Watts Bar Unit 1 reactor tripped from 100% power due to a turbine trip above P-9 (reactor trip on turbine trip permissive). The cause of the turbine trip is under investigation at this time. All systems functioned as designed with the exception of Pressurizer Backup Heaters which failed to energize on lowering Pressurizer pressure. The unit is stable in Mode 3 with Auxiliary Feedwater supplying the Steam Generators. The electrical system is in normal shutdown alignment with all Emergency Diesel Generators available in standby. There are no abnormal radiological conditions at this time. The reason the Pressurizer Backup Heaters failed to energize is unknown at this time. All control rods fully inserted. No relief valves or safety valves lifted. The licensee notified the NRC Resident Inspector. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Control Rod | |
| ENS 46418 | 14 November 2010 11:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Loss of Cooling to the 'A' Phase Main Bank Transformer | At 0617 (hrs. EST), Watts Bar experienced a failure of the cooling system to the 'A' phase Main Bank Transformer. Due to rising oil temperatures on the 'A' phase Main Bank Transformer, the reactor was manually tripped at 0652 hrs. All systems responded as designed with no issues. All rods inserted during the trip. There were no primary or secondary relief valves that lifted during the transient. The grid is stable and the plant is in a normal shutdown electrical line-up. The reactor is at normal pressure and temperature with decay heat being removed via the steam dumps to condenser with auxiliary feedwater providing steam generator make-up. The cause of the loss of cooling to the 'A' phase Main Bank Transformer was the failure of a control power transformer that supplies the Main Bank Transformer cooling system components. The licensee has notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater | |
| ENS 46177 | 15 August 2010 19:24:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Aux Feedwater System Actuation During Power Reduction for Turbine Maintenance | At 1524 EDT, on 8/15/2010, Watts Bar Nuclear Plant Unit 1 had a valid actuation of the Auxiliary Feedwater System in response to a trip of all Main Feedwater Pumps. The Auxiliary Feedwater Pumps were started manually in anticipation of the trip of all Main Feedwater Pumps. Main Feedwater Pump B had been tripped manually as part of the power reduction in preparation for taking the Main Turbine offline to perform repairs on the Electrohydraulic Control System. Main Feedwater Pump A was tripped manually before it experienced a loss of suction in response to an unplanned isolation of the Intermediate Pressure Heater string. Following manual actuation of the Auxiliary Feedwater Pumps, reactor power was further reduced, and at 1526 (EDT), Watts Bar Unit 1entered MODE 2. No plant safety systems beyond Auxiliary Feedwater were required or actuated in response to this event. The plant was stabilized using Auxiliary Feedwater and the Main Steam dump valves. Operators followed their Abnormal Operating Instruction, and reactor trip was not required. This event is being reported as a valid actuation of the auxiliary feedwater system in accordance with 10 CFR 50.72(b)(3)(iv). Offsite power is normal and EDG's are operable. The licensee notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Main Turbine Main Steam | |
| ENS 45944 | 21 May 2010 23:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on Turbine Trip | At 1937 Eastern Daylight Saving Time (EDT), Watts Bar Nuclear Power Plant Unit 1 experienced a reactor trip due to a turbine trip. This caused an automatic AFW Pump start from P-4 coincident with Lo Tave signal. The cause of the turbine trip has not yet been identified, and is under investigation. The plant is stable and is being maintained in Mode 3, at normal operating pressure and temperature, with steam generator and pressurizer levels normal. Plant systems responded to return the plant to a stable condition without complication, and all systems performed as expected with one exception: The 'B' Motor Driven Auxiliary Feedwater Backpressure Control Valve failed closed, but the Steam Driven Auxiliary Feedwater Pump provided sufficient feedwater so that all Steam Generators were provided sufficient feedwater to maintain cooling and normal steam generator level. Plans for plant restart are pending awaiting the cause investigation. All control rods inserted into the core. Plant decay heat removal is through the steam dumps to the main condenser. Offsite power is available and lined up to plant system loads. Watts Bar (NRC) Resident Inspector has been notified of this event. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 45210 | 17 July 2009 21:42:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of Emergency Diesel Generators Due to Loss of Power to Safety Related Bus | At 1742 EDT on July 17, 2009, while attempting to restore normal alignment providing offsite power following repair of the 'D' Common Station Service Transformer (CSST), the transfer from the alternate to the normal power supply for the 6.9kV Shutdown Board 2B-B failed because of an apparent failure of the interlock that should have prevented closing the normal breaker until an undervoltage condition existed on the alternate feed. The emergency diesel generators received a blackout signal and all four diesel generators started. Loads transferred and shed properly, and all systems functioned as expected. WBN (Watts Bar Nuclear) had been in LCO 3.8.1 condition A since 0032 EDT on July 16, 2009 due to the outage of the D CSST. Because loads did not transfer to the offsite power source, WBN remains in LCO 3.8.1 A which requires restoration of the offsite power supply by 0032 EDT 7/19/2009. All systems are operating properly and the plant is stable. Troubleshooting and maintenance has begun and WBN anticipates return of the normal offsite power supply to the 2B-B Shutdown Board within the time required by technical specifications. This event is reportable under 10 CFR 50.72(b)(3)(iv) as an event or condition that resulted in valid actuation of the emergency diesel generators. The licensee has notified the NRC Resident Inspector The 6.9kV Shutdown boards at Watts Bar are safety related busses. The electrical configuration prior to the event had the 1A-A and 2A-A Shutdown Boards aligned to their normal offsite power source. The 1B-B and 2B-B Shutdown Boards were aligned to an alternate offsite power source because their normal source, the 'D' CSST had been out of service for repairs. Following completion of the repairs to the 'D' CSST, the licensee attempted to restore normal offsite power to the 2B-B Shutdown Board from the 'D' CSST using a fast transfer from the alternate power supply. For unknown reasons, the normal supply breaker attempted to close onto the 2B-B bus before the alternate supply breaker had opened. This, in effect, would have paralled both the primary and alternate sources of power to the 2B-B bus. An interlock prevents paralleling these two sources and resulting in both supply breakers tripping and the bus being de-energized. This generated a blackout signal that started all four emergency diesels generators (EDGs). All the EDGs started as required, only the 2B-B loaded onto its associated shutdown bus, as expected, because of the bus had been de-energized. The other busses (1A-A, 1B-B, and 2A-A) remained energized and the associated EDGs did not load. The 1A-A, 1B-B, and 2A-A EDGs were shutdown and returned to a standby condition. The 2B-B EDG continued to power the 2B-B Shutdown Board while the licensee investigated the fast transfer problem. All systems functioned as required during this event except for the 2B-B fast bus transfer from the alternate to normal offsite power supply. | Emergency Diesel Generator | |
| ENS 44506 | 20 September 2008 13:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Turbine Trip | At 0906 EDT Watts Bar Unit 1 experienced a Reactor trip in response to a Turbine trip. This caused an automatic AFW Pump start from P-4 coincident with Lo Tave signal. First indications are that the Exciter Field Breaker tripped open. The cause is under investigation. All ESF systems responded as designed with no other issues. The plant is currently stable and is being maintained in Mode 3. Plans for plant restart are pending awaiting the cause investigation. All control rods fully inserted into the core. Plant decay heat removal is through the steam dumps to the main condenser. The offsite power is available and lined up to plant system loads. The licensee notified the NRC Resident Inspector. | Decay Heat Removal Main Condenser Control Rod | 05000390/LER-2008-004 |
| ENS 44388 | 7 August 2008 06:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Trip from 50% Power Following Low Pressure Heater String Isolation | On August 7, 2008 at 0228 EDT, TVA was reducing reactor power in preparation for a planned reactor shut down. While at approximately 50% power, the feedwater system isolated due to high levels in the low pressure heater strings. Based on the condition stated, the reactor was manually scrammed. All systems functioned as designed in response to the scram. The plant is currently being maintained in Mode 3 Hot Standby condition. All control rods fully inserted on the manual trip. Decay heat is being removed via auxiliary feedwater to the steam generators steaming to the main condenser. Offsite power is supplying safety buses and emergency diesel generators are available if required. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Emergency Diesel Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 42744 | 31 July 2006 16:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Main Generator Trip | At approximately 1213 hours on July 31, 2006, with Watts Bar Nuclear Plant Unit 1 operating normally at 100% power, the main generator tripped resulting in a reactor trip per design. All control rods inserted (fully) and the auxiliary feedwater system (AFW) automatically actuated per design and the reactor was stabilized in mode 3. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip (4-hour report) and under 10 CFR 50.72(b)(3)(iv)(A ) for the RPS (reactor protection system) and AFW actuations (8-hour report). The cause of the generator trip is currently under investigation. Steam is being released via steam dump to the condenser and all systems functioned as required. The licensee notified the NRC Resident Inspector. | Auxiliary Feedwater Control Rod | |
| ENS 42610 | 30 May 2006 21:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip on High Turbine Vibration | At approximately 1700 hours on May 30, 2006, with Watts Bar Nuclear Plant Unit 1 operating normally at 100% power, main turbine vibration increased to a value above the procedure limit and reactor operators manually tripped the reactor in accordance with site procedure requirements. All control rods inserted and the auxiliary feedwater system automatically actuated per design. No other significant equipment issues were identified and the reactor was stabilized in mode 3. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) for the manual reactor trip (4-hour report) and under 10 CFR 50.72(b)(3)(iv)(A) for the RPS and AFW actuations (8-hour report). Watts Bar had been monitoring indications of slightly elevated turbine vibration on the main turbine, but the reason for the increase above the procedure limit of 14 mils is not known at this time. TVA will be investigating the cause of the increased vibration to make necessary repairs before turbine startup. Decay heat is being removed by dumping steam to the main condenser. No safety or relief valves lifted. The licensee notified the NRC Resident Inspector.
As a result of the initial assessment of the turbine vibration discussed above, TVA has identified damage to the turbine end of the 'C' low pressure turbine. This is consistent with the initial indications of high vibration on the number 7, 8 and 9 bearings and not thought to be associated with previous condition monitoring of the number 11 bearing. Assessment and repair of secondary plant components damaged in the transient are in progress. The licensee will notify the NRC Resident Inspector. Notified R2DO (C. Ogle) | Auxiliary Feedwater Main Turbine Main Condenser Control Rod | |
| ENS 41054 | 19 September 2004 08:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip in Response to Indication of Dropped Control Rods | The following was received from the licensee via facsimile: While operating at 100% power, the Watts Bar Unit 1 reactor was manually tripped at approximately 0456 EDT on September 19, 2004, in response to an apparent four dropped rods in control bank 'B'. The cause of the dropped rods is being investigated. Safety systems functioned as expected in response to the trip. All control rods inserted properly in response to the reactor trip. The Auxiliary Feedwater (AFW) System actuated as designed in response to the trip. One reactor coolant pump (RCP) did not transfer to its alternate power supply. The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv). The actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv). Notification to NRC was made by Rick O'Rear at 0744 EDT. The NRC Senior Resident Inspector was notified of this event. The licensee also reported that the reactor is being maintained at the normal temperature and pressure for this condition. All station service electrical systems and the emergency diesel generators are available and in normal configuration. All emergency core cooling systems are available. Decay heat is being removed using steam dump to the main condenser and feedwater to the steam generators is being provided by the electric main feedwater pump. | Steam Generator Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System Main Condenser Control Rod | |
| ENS 40454 | 16 January 2004 16:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | At 1137 hours EST on January 16, 2004, with Watts Bar Unit 1 in Mode 1 at 100 % power, a turbine trip and reactor trip occurred. This is reportable as a 4-hour notification under 10 CFR 50.72 (b)(2)(iv)(B) and an 8-hour notification under 10 CFR 50.72 (b)(3)(iv)(A). In addition, per design, there was auto start of the Auxiliary Feedwater system which is reportable as an 8-hour notification under 10 CFR 50.72 (b)(3)(iv)(A). Plant safety systems performed as designed and the reactor is currently stable in mode 3. At the time of the trip, plant personnel were performing scheduled Solid State Protection System (SSPS) surveillance testing on the B Train Reactor Trip Breaker. Plant personnel are currently investigating the cause of the event. The licensee also reported that all control rods inserted on the reactor trip, no primary or secondary system relief valves operated, and that reactor temperature is being maintained using steam dump to the condenser. Steam generator water levels are being maintained using auxiliary feedwater. The station electrical system is available and in a normal configuration. All ECCS equipment is available. The reactor is currently stable at 2230 psig, 559 degrees Fahrenheit. The licensee has notified the NRC Resident Inspector. | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 40100 | 25 August 2003 13:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Turbine Trip/Reactor Trip Due to a Sudden Pressure Relay Signal from Main Transformer Bank "1C | On August 25, 2003, Watts Bar (WBN) Unit 1 was operating at 100 percent power when there was an operation of a "Sudden Pressure Relay" for Main Transformer Bank 1C. The actuation of the relay resulted in a turbine trip and a subsequent reactor trip at approximately 0945 EDT. The cause of the relay actuation is under investigation at this time. All control rods inserted as required and the safety systems actuated as designed including the motor and turbine driven pumps for the Auxiliary Feedwater (AFW) System. AFW pump 1B-B was inoperable at the time of the trip due to work on an area cooler. However, the pump was available for service and started as required. Unit 1 is currently stable in Mode 3 and will remain in this mode until the completion of the investigation into the cause of the trip. At the time of the Sudden Pressure Relay for Main Transformer Bank 1C, an oil sample was being taken of the Transformer. Fire Brigade was sent but there was no fire and no explosion to the Transformer. The electrical grid is stable, and Emergency Core Cooling systems and the Emergency Diesel Generators are fully operable if needed. At this time only the 1B-B Motor Driven Auxiliary Feedwater pump is still operating. The NRC Resident Inspector was notified of this event by the licensee. | Emergency Diesel Generator Auxiliary Feedwater Main Transformer Emergency Core Cooling System Control Rod |