Semantic search
Jump to navigation
Jump to search
| Discovered date | Reporting criterion | Title | Event description | |
|---|---|---|---|---|
| ENS 57175 | 16 June 2024 17:33:00 | 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: On June 16, 2024, at 1233 CDT, Waterford Steam Electric Station Unit 3 was operating at 93 percent power when an automatic reactor trip occurred. Immediately following the reactor trip, emergency feedwater (EFW) actuated automatically. The unit is currently in Mode 3. All control rods fully inserted. Decay heat removal is via the main condenser. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip, except steam generator (SG) feedwater pump 'A' tripped and SG '1' main feed regulatory controller went to manual. Steam generator water levels are being controlled with the SG feedwater pump 'B'. The cause of the trip is currently being investigated. This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system when the reactor is critical and as an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as a valid actuation of the EFW system. The NRC Resident Inspector has been notified. |
| ENS 57042 | 22 March 2024 04:37:00 | 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 | Notification of Unusual Event Due to Fire in the Protected Area | The following information was provided by the licensee: A Notification of Unusual Event, HU4.4 (see note below) was declared based a fire in the protected area requiring off site assistance to extinguish. The fire was in the main transformer yard. The fire was detected at 2328 CDT on March 21, 2024, and the fire was declared out at 0009 CDT on at March 22, 2024. An automatic reactor trip was initiated due to a loss of offsite power to the "B" train and a failure to automatically transfer from unit auxiliary transformer "B" to startup transformer "B. The licensee notified State and local authorities and the NRC Resident Inspector. The NRC remained in Normal. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). NOTE: Due to a typographical error initiating condition HU4.1 was initially recorded for the event. The correct initiating condition is HU4.4 as now shown.
The licensee terminated the Notification of Unusual Event at 0221 CDT on 3/22/24. The licensee notified the NRC Resident Inspector. Notified R4DO (Gepford), IR-MOC (Grant), NRR-EO (Felts), DHS-SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
RPS ACTUATION The following information was provided by the licensee via email: On March 21, 2024, at 2328 CDT, Waterford 3 Steam Electric Station, Unit 3 was operating at 98 percent power when an automatic reactor trip was initiated due to a loss of offsite power to the B train and a failure to automatically transfer from unit auxiliary transformer B to startup transformer B. Emergency feedwater actuation signal 2 (EFAS), safety injection actuation signal (ECCS), containment isolation actuation signal and emergency diesel generators automatically actuated. The unit is currently stable in Mode 3. All control rods fully inserted and all other plant equipment functioned as expected. Forced circulation remains with one reactor coolant pump per loop running. Decay heat removal is via the main condenser. A train safety bus is being supplied by off-site power, and B train safety bus is being supplied by emergency diesel generator B. Following the loss of offsite power to the B train, it was reported that main transformer B and startup transformer B were both on fire. The Emergency Director declared an Unusual Event at time 2337 CDT. The fire was reported extinguished at 0009 CDT on March 22, 2024, and the Unusual Event was terminated at 0221 CDT on March 22, 2024. Offsite assistance was requested. The local fire department responded to the site but the fire was extinguished by the on-shift fire brigade. NRC Region IV management was contacted regarding the emergency plan entry at 0030 CDT on March 22, 2024. This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system, ECCS, Containment Isolation and Emergency Diesel Generators. The NRC Resident Inspector has been notified. Notified R4DO (Gepford)
The initial notification in event notice #57042 by Waterford Steam Electric Station, Unit 3, reported a Notice Of Unusual Event (NOUE) emergency declaration due to a fire in the protected area requiring off site support to extinguish. The basis for retraction of the initial emergency notification is that this event did not meet the definition of a fire in the protected area that requires off site support to extinguish. Guidance provided in Nuclear Energy Institute (NEI) 99-01, Rev. 6 and implemented in Waterfords Emergency Plan procedure, initiating Condition HU4.4 states, The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. (NOTE: The Initial Notification Form sent from the Control Room at 2341 CDT on March 21, 2024, requested by and provided to the NRC Headquarter Operations Center via e-mail at 0302 CDT on March 22, 2024, stated that the Emergency Classification had been made on Initiating Condition HU4.4 rather than HU4.1)" When the event occurred on March 21, 2024, the Emergency Director declared an Unusual Event at 2337 CDT and requested offsite support based on the information available at that time including the initial assessment by the fire brigade leader and expected need for offsite support to extinguish the fire. As reported in the 0420 EDT update on March 22, 2024, the fire was reported extinguished at 0009 CDT on March 22, 2024, by the Waterford Fire Brigade without the need of offsite support." Notified R4DO (Kellar).
The initial notification in EN 57042 by Waterford Steam Electric Station, Unit 3, reported an emergency declaration of an Unusual Event due to a fire in the protected area requiring off site support to extinguish. The basis for the update to the initial notification is that this event did not meet the definition of a Fire in the Protected Area that requires offsite support to extinguish. As provided in NEI 99-01, Rev. 6 and implemented in Waterfords emergency plan procedure, initiating condition HU4.4 states, The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Additionally, EAL 4.1 for a fire not extinguished within 15 minutes of detection in any Table H-1 fire area was not applicable because the fire did not occur in a Table H-1 fire area. When the event occurred on March 21, 2024, the Emergency Director declared an Unusual Event at 2337 CDT and requested offsite support based on the information available at that time including the initial assessment by the fire brigade leader and expected need for offsite support to extinguish the fire. As reported in the 0420 EDT update on March 22, 2024, the fire was reported extinguished at 0009 CDT on March 22, 2024, by the Waterford Fire Brigade without the need of offsite support. (NOTE: The Initial Notification Form sent from the Control Room at 2341 CDT on March 21, 2024, requested by and provided to the Headquarters Operation Center via e-mail at 0302 CDT on March 22, 2024, stated that the emergency classification had been made on initiating condition HU4.4 rather than HU4.1) In accordance with NRC Approved guidance in FAQ 21-02 (ML21117A104), Waterford 3 is retracting the initial event notification made at 0117 EDT on March 22, 2024. The remaining events that were reported in EN 57042 as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW (emergency feedwater) system, ECCS (emergency core cooling system), containment isolation and emergency diesel generators are still applicable and require no additional update at this time. The licensee also provided a site map. Notified R4DO (Kellar) |
| ENS 57004 | 4 March 2024 00:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Main Turbine Trip on LOW Condenser Vacuum | The following information was provided by the licensee via email: On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram. Additionally, following the scram a low RPV (reactor pressure vessel) level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR (residual heat removal) shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system. This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration. |
| ENS 56645 | 30 July 2023 19:26:00 | 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 LOW Main Turbine ELECTRO-HYDRAULIC Control (EHC) Oil Level | The following information was provided by the licensee via email: On July 30, 2023 at 1526 EDT, with unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to low main turbine electro-hydraulic control oil level. The trip was uncomplicated with all systems responding normally post-trip. Operations stabilized the plant in mode 3. Decay heat removal is being accomplished using the steam dumps in steam pressure mode to the main condenser. Emergency Feedwater actuated due to low-low steam generator level as expected. 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. |
| ENS 56328 | 1 February 2023 05:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Scram Due to Turbine Trip | The following information was provided by the licensee via fax and telephone: Generator trip due to power load unbalance which caused a turbine trip and subsequent reactor scram. Experienced a trip on circulating water pump A. NRC Resident Inspector notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Off-site power available and unaffected. Decay heat removal via main steam line and drains to condenser. Plant is stable in mode 3. |
| ENS 56174 | 22 October 2022 17:58:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | ECCS Injection While Shutdown | The following information was provided by the licensee via fax: During Mode 5 Refueling operations, while attempting to establish flow through the Fuel Pool Cooling system filter demineralizers, an air operated valve to a radioactive waste tank failed to close automatically. This caused the Fuel Pool Cooling system to pump water from the Skimmer Surge Tanks (SST) to the radioactive liquid waste system. In response to the loss of inventory from the SSTs, the Control Room operating crew started Core Spray Pump A to restore normal operating level In the SST. This prevented the loss of the Fuel Pool Cooling/Alternate Decay Heat Removal system which was the only in service system meeting the safety function of decay heat removal. Core Spray Pump A was used for Injection for less than 3 minutes. This is reportable as a discharge of ECCS into the RCS in response to an event, but not part of a pre-planned sequence under 10 CFR 50.72(b)(2)(iv)(A) and actuation of a specified system under 10 CFR 50.72(b)(3)(Iv)(A). The resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Licensee reported approximately 6000-7000 gallons of water was injected into the RCS. The stuck open air operated valve was closed. Proceeding with refueling outage operations. |
| ENS 55612 | 27 November 2021 10:19:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of the Emergency AC Electrical Power System | At 0519 EST on November 27, 2021, with Unit 2 in Mode 5 at zero percent power, an actuation of the Emergency AC Electrical Power System occurred. The reason for the Emergency AC Electrical Power System auto-start was a lockout of the CT-2 transformer; causing a temporary loss of AC power to the main feeder bus. The Keowee Hydroelectric Units 1 and 2 automatically started as designed when a main feeder bus undervoltage 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 AC Electrical Power System. Additionally, the temporary loss of AC power resulted in a loss of Decay Heat Removal (DHR) that was restored upon power restoration to the main feeder bus. Therefore, this condition is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v) for an event or condition that could have prevented fulfillment of a safety function. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The loss of the CT-2 transformer is under investigation. Main feeder bus power was restored within a minute so no plant heat up occurred as a result of the loss of the decay heat removal system. |
| ENS 55527 | 16 October 2021 10:28:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to a Fire in the Protected Area | (An) Unusual Event (was declared) due to a fire in the protected area not extinguished in less than 60 minutes. Main power transformer 3 faulted, the unit auto scrammed, all rods are in. The fire went out at 0622 CDT. (The licensee) is monitoring for re-flash. The unit automatically scrammed and all rods fully inserted. Decay heat removal is through the condenser. Unit 2 is unaffected and remains at 100 percent power. The licensee notified the NRC Resident Inspector and R3 Branch Chief (Riemer). Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), and DHS Nuclear SSA (email).
The Unusual Event was terminated at 0709 CDT. Notified R3DO (Pelke), NRR EO (Felts), IR MOC (Kennedy), DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), and DHS Nuclear SSA (email). |
| ENS 55504 | 4 October 2021 04:31:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid System Actuation | The 'A' Steam Generator Narrow Range Water Level went less than 17 percent causing an Auxiliary Feed Water System valid actuation signal. The Auxiliary Feed Water System was in service at the time of the event providing decay heat removal. There was no adverse effect on plant systems. The Steam Generator Narrow Range Water Level was restored to normal operating band. This is being reported per 10 CFR 50.72(b)(3)(iv)(A), which states, '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.' (Reactor Coolant System) RCS Pressure 340 pounds and RCS Temperature 340 Degrees F. The NRC Resident Inspector was notified. |
| ENS 55346 | 9 July 2021 01:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Main Turbine Trip | At 2154 EDT on 7/8/2021, with the Unit in Mode 1 at 100% power, the reactor automatically tripped due to trip of the main turbine, caused by failure of a non-safety related breaker during functional testing. Following the reactor trip the Steam Feed Rupture Control System automatically initiated on low Steam Generator 1 level, actuating both turbine-driven Auxiliary Feedwater Pumps. The operators subsequently started the high pressure injection pumps manually per procedure in response to overcooling indications. Operations responded and stabilized the plant. Decay heat was initially being removed via the Main Condenser. During post-trip response actions, while attempting to shut down the Auxiliary Feedwater Pumps, a low pressure condition was experienced in Steam Generator 2, resulting in isolation of the Main Condenser and steam being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage. 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). This event is also being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a four-hour, non-emergency notification of emergency core cooling system (ECCS) discharge into the reactor coolant system, and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an eight-hour, non-emergency notification of an event that results in a valid actuation of the Auxiliary Feedwater System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55104 | 15 February 2021 11:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Low Steam Generator Level | At 0526 (CST) on 02/15/2021, Unit 1 automatically tripped due to low steam generator levels. The low steam generator levels were due to loss of Feedwater pumps 11 and 13 (cause unknown). Auxiliary Feedwater and Feedwater Isolation actuated as designed. All Control and Shutdown Rods fully inserted. No primary or secondary relief valves opened. There were no electrical problems. Normal operating temperature and pressure (NOT/NOP) is 567 degrees F and 2235 psig. There were no significant TS LCOs entered. This event was not significant to the health and safety of the public based on all safety systems performed as designed. Unit 2 was not affected. Decay heat removal is being controlled via Steam Dumps. (Auxiliary Feedwater is supplying water to the Steam Generators.) Offsite power is in the normal electrical lineup. The NRC Resident inspector has been notified. Unit 2 was not affected and remains at 100% power. |
| ENS 55078 | 20 January 2021 23:22:00 | 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 Trip of Motor Control Center | On 1/20/2021 at 1822 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a loss of Motor Control Center 2B2. The trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in Mode 3. Auxiliary feed-water automatically actuated on the 2A Steam Generator post trip. Current decay heat removal is the 2B main feedwater pump to both steam generators and the Steam Bypass Control System to the main condenser. Unit 1 is not affected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. |
| ENS 54847 | 21 August 2020 03:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Steam Generator Water Level Control | On 08/20/20 at 2354 (EDT), with Unit 3 in Mode 1 at approximately 34% RTP (Rated Thermal Power), the reactor was manually tripped. This was due to Steam Generator Water Level control issues that resulted in the only Steam Generator (S/G) Feed Pump tripping on low suction pressure. Unit 3 reactor was tripped manually upon the loss of the last running feed pump. All other systems operated normally. Auxiliary Feedwater initiated as designed to provide S/G water level control. EOPs have been exited and General Operating Procedures (GOPs) were entered. Unit 3 is stable in Mode 3 at normal operating temperature and pressure. 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 bas been notified. Decay heat removal is by the steam dumps to atmosphere. Unit 4 is not affected. The cause of the low suction feed pressure to the steam generator feed pump is under investigation. |
| ENS 54740 | 6 June 2020 13:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip When Group One, Bank 'B' Control Rods Inserted Into the Core | At 0920 (EDT), with the unit in Mode 1 and 100 percent power, the reactor was manually tripped due to group 1 of control rod bank 'B' fully inserting into the core. All systems responded normally post trip. Operations has stabilized the plant in mode 3 at NOP/NOT (normal operating pressure and temperature). Decay heat removal is being accomplished via the steam dumps in the steam pressure mode to the main condenser. Emergency feedwater actuated due to low low steam generator level as expected. This event is being reported pursuant to 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A) The senior NRC Resident Inspector has been notified. The plant response to the trip was uncomplicated. All safe shutdown equipment is available. There were no reliefs or safeties actuated during the transient. The licensee manually tripped eight days ago for the same condition. See EN #54731. |
| ENS 54731 | 29 May 2020 18:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Trip Due to Rod Bank Unexpectedly Inserting | At 1403 EDT, with the unit in Mode 1 and 100 percent power, the reactor was manually tripped due to Group 1 of Control Rod Bank 'B' unexpectedly inserting. All systems responded normally post-trip. Operations stabilized the plant in Mode 3 at 557 degrees Fahrenheit. Decay heat removal is being accomplished via the steam dumps in the steam pressure mode to the main condenser. Emergency feedwater actuated due to low low steam generator level as expected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified. |
| ENS 54725 | 25 May 2020 09:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Main Turbine Trip | An (automatic) reactor SCRAM occurred at 0433 CDT, on 05/25/2020, from 66 percent core thermal power. The cause of the SCRAM was due to a Main Turbine Trip. The cause of the Turbine Trip is under investigation. All systems responded as designed. No loss of offsite power or (Emergency Safety Feature) (ESF) power occurred. No (Emergency Core Cooling System) (ECCS) or Emergency Diesel Generator initiations occurred. Main Steam Isolation valves remained open and no radioactive release occurred due to this event. The plant is stable in mode 3. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. The NRC Resident Inspector has been notified. Decay heat removal is through the Feedwater and Condensate System. |
| ENS 54708 | 13 May 2020 06:08:00 | 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Trip | At 0208 EDT on 05/13/2020, Sequoyah Unit 1 was at 100% power when an automatic reactor trip signal was received concurrent with a low steam line pressure safety injection signal. The low steam line pressure safety injection signal was actuated from the steam pressure rate of decrease feature. Main steam isolation valves (MSIVs) automatically closed as designed and steam generator pressures stabilized following the isolation. All other safety-related equipment operated as designed, with the exception of 1-FCV-61-122 Glycol inboard containment isolation valve which failed to automatically isolate on a Phase A containment isolation signal. The corresponding outboard containment isolation valve, 1-FCV-61-110, automatically isolated as designed which isolated penetration X-114. Safety injection was terminated at 0221 EDT 5/13/20, and Unit 1 is currently being maintained in Mode 3 at normal operating temperature and pressure with auxiliary feedwater supplying the steam generators and decay heat removal via steam generator atmospheric relief valves. There is no indication of any primary to secondary leakage. The electrical alignment is normal with shutdown power supplied from off-site power. There is no current operational impact to Unit 2. There is no impact on public health or safety. Post safety injection actuation investigation is in progress. The NRC Resident Inspector has been notified. |
| ENS 54663 | 13 April 2020 19:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Trip Due to Low Condenser Vacuum | At 1550 EDT on 4/13/2020, with Millstone Power Station Unit 3 operating at approximately 82 percent reactor power, an automatic reactor trip occurred following a turbine trip due to low condenser vacuum caused by the trip of multiple circulating water pumps. Due to the loss of the circulating water pumps, decay heat removal was established by the steam generator atmospheric dump valves. All other systems responded as expected to the trip. Auxiliary feedwater actuated automatically as expected following the trip due to low-low levels in the steam generators. There was no risk to the public. There was no impact to Millstone Unit 2. The Senior Resident Inspector has been informed. This event is being reported as a four hour report under 10 CFR 50.72(b)(2)(iv)(B) as a condition that resulted in actuation of the reactor protection system while the reactor was critical, and as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(3)(iv)(B) for actuation of the auxiliary feedwater system. The licensee also notified the state of Connecticut, the Connecticut Department of Energy and Environmental Protection, and the city of Waterford. |
| ENS 54631 | 1 April 2020 20:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Circuit Fault | On April 1, 2020, at 1625 EDT, Milllstone Unit 3 was in Mode 1, at 100 percent power, when an automatic reactor trip occurred following a main generator trip. The cause was due to a circuit fault between the main generator breaker and the offsite switchyard. The reactor trip was not complicated and the reactor remains stable in Mode 3. One of the two offsite electrical sources remain inoperable with an investigation of the circuit fault underway. Decay heat removal is maintained by the main condenser. There was no effect on Unit 2. The NRC Resident Inspector was notified. The licensee notified State and local government agencies. |
| ENS 54597 | 22 March 2020 16:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram | At 1255 Eastern Daylight Time (EDT) on March 22, 2020, with Unit 1 in Mode 2, stabilized at 2 percent power, coming out of a refueling outage, all 4 main turbine Bypass Valves (BPVs) opened unexpectedly. As a result, the main control room inserted a manual reactor scram. All control rods inserted as expected during the scram. In accordance with plant procedures, the main control room closed all Main Steam Line Isolation Valves (MSIVs) to arrest the cooldown resulting from BPVs remaining open. The condensate system remained aligned for injection and pressure control was initially via main steam line drains. RHR (residual heat removal) shutdown cooling was placed in operation for decay heat removal and pressure control once the MSIVs were closed. All systems responded as designed, with the exception of the BPVs. 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. There was no impact to Unit 2. |
| 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. |
| ENS 54449 | 18 December 2019 00:29:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown | At 1929 EST, on 12/17/19, Millstone Unit 3 began preparations for shutting down the reactor as the 'A' Emergency Diesel Generator (EDG) could not be restored to operable status within the 14-day outage time, requiring a Technical Specification (Tech Spec) shutdown. Per Tech Spec 3.8.1.1., the reactor must be in Hot Standby in six (6) hours, and Cold Shutdown within the following 30 hours. Hot Standby is estimated by midnight, and Cold Shutdown by 1800 EST on 12/18/19. All other safety and shutdown systems are operable. Decay heat removal will be through the Shut Down Cooling and Residual Heat Removal systems. There was no impact to Unit 2. There was no impact to the health and safety of the public or plant personnel. The licensee notified the state of Connecticut, Waterford County, and the NRC Resident Inspector. |
| ENS 54351 | 25 October 2019 01:51:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Undervoltage Actuation During Edg Sequencer Maintenance | On October 24, 2019, at 2051 Central Time, while performing Train C Sequencer maintenance, a valid undervoltage actuation signal was sent to Unit 2 Emergency Diesel Generator (EDG) 23. The ESF Train C bus loads were shed but EDG 23 did not automatically start because it had been placed in Pull-To-Stop to support the sequencer maintenance activities. EDG 23 was taken out of Pull-To-Stop by Control Room staff to allow it to auto start and load the bus. As a result of the bus strip signal, the in service Spent Fuel Pool Cooling Pump secured. Spent Fuel Pool Cooling was restored with no measurable increase in pool temperature. The reactor was not critical and reactor decay heat removal was not challenged throughout the event. This actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A) due to the automatic actuation of a system listed in 10 CFR 50.72(b)(3)(iv)(B). The NRC Resident Inspector has been notified. |
| ENS 54262 | 7 September 2019 12:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | On September 09, 2019 at 0824 EDT, with St. Lucie Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped on Low Reactor Coolant System Flow due to a trip of the 1A1 reactor coolant pump. The trip was uncomplicated with all systems responding normally post-trip. Operators responded and stabilized the plant in Mode 3. The cause of the loss of the 1A1 reactor coolant pump is currently under investigation. St. Lucie Unit 2 was unaffected and remains in Mode 1 at 100 percent power. This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip. Decay heat removal is being accomplished by main feed water and the main condenser using the turbine steam bypass valves. The licensee notified the NRC Resident Inspector. |
| 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. |
| ENS 54027 | 25 April 2019 13:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Turbine Generator Trip | At 0918 (EDT) on 4/25/19, with (Saint Lucie) Unit 1 in Mode 1 at 100% power, the reactor automatically tripped due to a Turbine Trip. The reactor trip was uncomplicated with all systems responding normally. Operations is maintaining the plant stable in Mode 3. Decay heat removal is being accomplished by main feed water and the main condenser using the turbine steam bypass valves. Unit 2 is not affected and remains at 100% power. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. |
| ENS 53954 | 24 March 2019 18:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Automatic Trip on Turbine Trip | On March 24, 2019, at 1445 EDT, Indian Point Unit 2 automatically tripped on a turbine trip due to a loss of excitation. All control rods fully inserted and plant equipment responded normally to the unit trip. This RPS (reactor protection system) actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater system actuated following the automatic trip as expected. This specified system actuation is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. Unit 2 is in Mode 3 at normal operating temperature and pressure. Decay heat removal is via the steam generators to the atmospheric steam dumps. No radiation was released. Indian Point Unit 3 was unaffected by this event and remains defueled in a scheduled refueling outage. A post trip investigation is in progress. The licensee has notified the NRC Resident Inspector The New York State Public Service Commission, Consolidated Edison System Operator, and New York State Independent System Operator were also notified. |
| ENS 53703 | 29 October 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Inadequate Feedwater Flow | On October 29, 2018 at 1317 EDT, with St. Lucie Unit 1 in Mode 1 at 100% power, the reactor was manually tripped due to inadequate feedwater flow to both 1A and 1B Steam Generators (S/Gs). The trip was uncomplicated with all systems responding normally post-trip. (All control rods fully inserted and there were no specified system actuations.) Operators responded and stabilized the plant in Mode 3. The cause of the inadequate feed flow to the 1A and 1B Steam Generators is currently under investigation. Decay Heat removal is being accomplished through forced circulation with stable conditions from Main Feedwater and the Steam Bypass Control System to the Main Condenser. Currently maintaining Pressurizer pressure at 2250 psia and Reactor Coolant System temperature at 532 degrees F. St. Lucie Unit 2 was unaffected and remains in Mode 1 at 100% power. This report is submitted in accordance with 10CFR50.72(b)(2)(iv)(B) for the reactor trip. The NRC Senior Resident Inspector has been notified. |
| 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. |
| ENS 53665 | 12 October 2018 04:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | On October 12, 2018 at 1353 EDT, St. Lucie Unit 2 experienced an automatic RPS actuation and Reactor Trip due to a fault on the 2A1 6.9kv bus during a transfer of the bus power supply from the 2A Auxiliary Transformer to the 2A Startup Transformer. The bus fault caused a fire in the 2A1 6.9kv switchgear that has been extinguished. Offsite support was not required to extinguish the fire. The specific cause of the fault is currently under investigation. Following the reactor trip, both Steam Generators are being supplied by main feedwater. All (Control Element Assemblies) (CEAs) fully inserted into the core. Decay Heat removal is being accomplished through forced circulation. Main Feedwater and Steam Bypass Control Systems are maintaining stable conditions in Mode 3. St. Lucie Unit 1 was unaffected and remains in Mode 1 at 100 percent power. This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the Reactor Trip. The fire was extinguished within 28 minutes. Plant loads are being supplied by the 2B Auxiliary Transformer. The licensee notified the NRC Resident Inspector. |
| 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. |
| 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. |
| ENS 53396 | 10 May 2018 20:17:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event | Unit 3 experienced a loss of AC power while in Mode 6. Power was regained automatically from Keowee via the underground path. Decay heat removal has been restored. Spent fuel cooling has been restored. Emergency procedures (are) in progress. The Licensee notified the senior NRC resident inspector, State of South Carolina and local authorities. The total loss of 4160 volt AC power was for approximately 30 seconds. The unit is refueled and reactor reassembly complete up to bolting on the reactor head. There was no effect on Units 1 and 2. Notified DHS SWO, FEMA Ops Center, FEMA NWC, DHS NICC, and NuclearSSA
At 0530 EDT, Oconee terminated the notification of unusual event on Unit 3. The basis for termination was that offsite power was restored and the plant is now in its normal shutdown electrical lineup. The licensee has notified Oconee and Pickens counties and will be notifying the NRC Resident Inspector. Notified R2DO (Ehrhardt), NRR EO (Miller), IRD MOC (Gott), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email) and NuclearSSA (email). |
| ENS 53386 | 7 May 2018 07:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip | This 4 and 8 hour notification is being made to report that Salem Unit 2 initiated a manual reactor trip and subsequent automatic Auxiliary Feedwater system actuation. The trip was initiated due to a 21 Reactor Coolant Pump reaching its procedural limit for motor winding temperature of 302F. Salem Unit 2 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 PSIG and Reactor Coolant System temperature is 547 F with decay heat removal via the Main Steam Dump and Auxiliary Feedwater Systems. Unit 2 has no active shutdown technical specification action statements in effect. All control rods inserted on the reactor trip. All ECCS (emergency core cooling systems) and ESF (emergency safety function) systems functioned as expected. No safety related equipment or major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event. The NRC Resident Inspector was notified. The Lower Alloways Creek Township will be notified. |
| ENS 53348 | 20 April 2018 01:07:00 | 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 Turbine Issue | While performing a turbine startup, a turbine control anomaly caused a steam generator level transient. The rise in steam generator level above the setpoint caused the turbine to automatically trip. The high steam generator level of 73 percent caused a feedwater isolation signal at 2107 EDT, which also tripped both Main Boiler Feed Pumps. The tripping of the Main Boiler Feed Pumps auto started the motor driven Aux Boiler Feed Pumps 21 and 23. The reactor was manually tripped at 2108 EDT in accordance with AOP-FW-1 Loss of Main Feedwater. All control rods inserted. Electrical power is being provided from offsite via the Station Aux Transformer. Decay heat removal is being provided via the Atmospheric Dump Valves. An investigation into the cause of the turbine control anomaly is underway. The NRC Resident Inspector has been notified. The event did not have an affect on Unit 3 and there is no primary to secondary leakage. |
| 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. |
| ENS 53216 | 16 February 2018 07:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Trip Due to Loss of Excitation on Main Generator | On February 16, 2018 at 02:01 EST Indian Point Unit 3 automatically tripped on a turbine trip due to a loss of main generator excitation. All control rods fully inserted and all plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater system actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. The plant is stable, in Mode 3, at no load operating temperature and pressure. Decay heat removal is via the steam generators to the main condenser via the condenser steam dumps. No radiation was released. Indian Point Unit 2 was unaffected by this event and remains at 100 percent power. A post trip investigation is in progress. The licensee has notified the NRC Resident Inspector. |
| 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. |
| ENS 53060 | 8 November 2017 00:57:00 | 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 | On 11/7/2017 at 1957 (EST), VC Summer Nuclear Station automatically tripped due to a turbine trip. The cause of the turbine trip is under investigation at this time. All systems responded as expected. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The unit is currently stable in Mode 3 with decay heat removal via the Main Steam to the Main Condenser. The NRC Resident Inspector has been notified. The licensee will notify the South Carolina State Emergency Management Division, the Fairfield, Richland, Lexington and Newberry Counties.
All systems responded as expected, with the exception of 'B' Steam Generator Feedwater Isolation Valve XVG1611 B-FW. This valve did not appear to automatically close and was slow to indicate closed from the Main Control Board. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. Notified the R2DO (Musser). |
| ENS 53052 | 4 November 2017 00:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Indian Point Unit 3 Reactor Trip on Low Steam Generator Level | On November 3rd, 2017 at 2022 EDT, the Indian Point Unit 3 Reactor Protection system automatically actuated at 100 percent power. Annunciator first out indication was from 33 SG (Steam Generator) Low Level. This automatic reactor trip is reportable to the NRC under 10 CFR 50.72(b)(2)(iv)(B). All control rods fully inserted on the reactor trip. All safety systems responded as expected. The Auxiliary Feedwater System actuated as expected. Offsite power and plant electrical lineups are normal. All plant equipment responded normally to the unit trip. No primary or secondary code safeties lifted during the trip. The Auxiliary Feedwater System actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). The Emergency Diesel Generators did not start as offsite power remained available and stable. The Unit remains on offsite power and all electrical loads are stable. Unit 3 is in Hot Standby at normal operating temperature and pressure with decay heat removal using auxiliary feedwater to the steam generators and normal heat removal through the condenser via the high pressure steam dumps. Unit 2 was unaffected and remains at 100 percent power. A post trip investigation is in progress. The licensee indicated that Radiation Monitor number 14 spiked twice during the transient, however, is currently not indicating any signs of radiation. The licensee will notify the NRC Resident Inspector and the NY Public Service Commission.
The initial notification stated that Indian Point Unit 3 Reactor Tripped on 33 SG (Steam Generator) Low Level, this is incorrect. Indian Point Unit 3 Reactor Tripped on a Turbine Trip. The Turbine Trip was caused by a Generator Back-up Lockout Relay. The Turbine Trip was the 'first' annunciator first-out but was acknowledged instead of silenced during initial operator actions. The Turbine Trip first-out being acknowledged allowed a Low Steam Generator first-out to later annunciate. A Low Steam Generator Level is an expected condition post trip. This update does not change any actions taken by the operating team or required notifications under 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). A post trip investigation remains in progress. The licensee will notify the NRC Resident Inspector and the NY Public Commission. Notified the R1DO(Cook) |
| ENS 53036 | 26 October 2017 06:12:00 | 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 a Loss of Load | On October 26, 2017 at 0212 EDT St. Lucie Unit 2 experienced a reactor trip due to a loss of load event resulting in an RPS (Reactor Protection System) actuation. The cause of the loss of load is currently under investigation. Following the reactor trip, an Auxiliary Feedwater Actuation Signal occurred due to low level in the 2A Steam Generator. One of the two Main Feed Isolation Valves to the 2A Steam Generator did not close on the Auxiliary Feedwater Actuation Signal. 2A Steam Generator level was restored by Auxiliary Feedwater. The 2B Steam Generator level is being maintained by Main Feedwater. All CEAs (Control Element Assemblies) fully inserted into the core. Decay heat removal is being accomplished through forced circulation with stable conditions from Auxiliary Feedwater/Main Feedwater and Steam Bypass Control System. Currently maintaining pressurizer pressure at 2250 psia and Reactor Coolant System temperature at 532 degrees F. St. Lucie Unit 1 was unaffected and remains in Mode 1 at 100 percent power. This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72(b)(3)(iv)(A) for the Specified System Actuation. The licensee notified the NRC Resident Inspector. |
| ENS 52992 | 26 September 2017 09:43:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of the 2B Emergency Diesel Due to a Valid Undervoltage Signal | At approximately 0543 (EDT), while (recovering from the performance of) 2B Emergency Diesel Generator and ESFAS testing, a (subsequent) valid undervoltage actuation signal was sent to the 2B Emergency Diesel Generator (EDG). The 2B AC emergency bus (2BA03) was load shed, the 2B EDG automatically started, and tied to 2BA03. The 2BA03 bus was loaded by the automatic load sequencer. The actuation was identified by the Control Room operators and the 2B EDG was locally monitored while in service. This actuation is reportable due to the automatic actuation of a system listed in 10 CFR 50.72(b)(3)(iv)(B). The reactor was not critical at the time of the event and not challenged throughout the event. Decay heat removal and spent fuel pool cooling were not challenged throughout the event. The NRC Resident Inspector has been notified. The cause of the undervoltage condition is under investigation. |
| ENS 52950 | 6 September 2017 15:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram on Low Reactor Vessel Water Level | On September 6, 2017 at 1157 (EDT), Nine Mile Point Unit 1 experienced an automatic reactor scram with a Main Steam Isolation Valve (MSIV) and Containment isolation. The scram was due to reactor vessel low water level. The cause of the reactor vessel low water level is under investigation. All control rods fully inserted. Following the scram, pressure was momentarily controlled through the use of the Emergency Condenser (EC) system. At 1205, pressure control was established through the main steam lines to the condenser through the turbine bypass valves. All plant systems responded per design following the scram. The reactor scram is a 4-hour report per 10 CFR 50.72(b)(2)(iv)(B). The following systems automatically actuated after the scram as expected. These system actuations are an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A): 1. The High Pressure Coolant Injection (HPCI) system. HPCI initiated at 1157 and was reset at 1158 when RPV level was restored above the HPCI system low level actuation set point. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. 2. The Core Spray system actuated, but did not discharge to the Reactor Coolant system. The Core Spray system was secured at 1216. 3. Containment and MSIV isolation on reactor vessel low-low water level signal. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor vessel water level and pressure maintained within normal bands. Decay heat is being removed via steam to the main condenser using the turbine bypass valves. The offsite grid is stable with no grid restrictions or warnings in effect. The licensee has notified the NRC Resident Inspector. No safety relief valves lifted during the transient. The main steam isolation valves were opened after the isolation signal cleared to facilitate decay heat removal. Offsite power is supplying all plant loads. There was no effect on Unit 2. The licensee notified New York State Department of Environmental Protection and will be issuing a press release. |
| ENS 52932 | 28 August 2017 12:37:00 | 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 | On 8/28/2017 at 0837 (EDT), VC Summer Nuclear Station automatically tripped due to a turbine trip. The turbine trip was caused by the Main Generator Differential Lockout due to a fault on the center phase lightning arrester on the Main Transformer (XTF-001). There were no complications with the trip. All control rods fully inserted. Balance of Plant (BOP) buses automatically transferred to their alternate power source XTF 31/32. All Emergency Feedwater pumps started as required. All systems responded as required. The plant is stable in Mode 3. Station personnel are investigating the cause of the fault on the main transformer lightning arrester. This event is reportable per 10 CFR 50. 72(b)(2)(iv)(B) and 10 CFR 50. 72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. The unit is currently stable in Mode 3 with decay heat removal via the Main Steam to the Main Condenser. The licensee will inform both State and local authorities. |
| ENS 52767 | 24 March 2017 18:25:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid System Actuation During Testing | On March 24, 2017, at 1425 EDT, while performing Engineered Safeguards Actuation System (ESAS) quarterly High Pressure Injection/Low Pressure Injection Logic and Component testing, an unintended test signal was generated when a test switch was moved to the OFF position but went slightly past this position and engaged contacts for the Test no. 1 position. When examined, the test switch was found to be degraded which allowed the switch to move past the center position and engage the test no. 1 contacts. This resulted In a partial actuation of 'B'- train ESAS components. It also resulted in an injection to the reactor coolant system (RCS). The test signal was immediately removed by operators and the inadvertently started equipment secured. The plant was operating at 100% power when the event occurred. There were no valid signals or plant conditions present to warrant the safety system actuation. The 'B' Emergency Diesel Generator rolled on air start but did not get up to full speed. Decay Heat Removal Pump 'B' started and the Decay Heat Removal Injection valve 4B opened, Make-Up Pump 'C' started, Make-Up Pump suction valve 14B opened, Make-Up pump discharge valves 16C and 16D opened, Spent Fuel Pump 1B tripped off, Air Handling Fan 18 tripped off and Air Handling Fan 1C trip tripped off. These components properly functioned from the inadvertent test signal and were secured prior to any adverse impact to plant operation. There was a small injection of borated water into the RCS. The plant remained stable at 100% power operation. Pursuant to 10 CFR 50.73(a)(1) the following information is provided as a sixty (60) day telephone notification to the NRC. This notification, reported under 50.73(a)(2)(iv)(A), is being provided in lieu of the submittal of a written LER to report a condition that resulted in an invalid partial actuation of the 'B' train of the Engineered Safeguards Actuation System (ESAS) as it was not part of a pre-planned sequence. The Licensee notified the NRC Resident Inspector. |
| 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. |
| 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. |
| ENS 52718 | 29 April 2017 22:44:00 | 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 Steam Generator Hi-Hi Level Signal and Feedwater Isolation | At 1844 (EDT) on 04/29/2017, while the unit was in a low power condition exiting from a refueling outage, the reactor was manually tripped following a P-14 signal (Steam Generator Hi-Hi Level) and a resulting feedwater isolation signal. All control rods were verified to be fully inserted. The cause of the ('B') steam generator high level is currently being investigated. Emergency feedwater actuated at 1845 due to a low-low water level in steam generator 'D'. Plant equipment response is being evaluated and the plant is stabilized in Mode 3 with decay heat removal through the steam dump system to the condensers. There was no release and the emergency feedwater system has been restored to standby. The event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. The licensee notified the NRC Resident Inspector. |
| ENS 52469 | 30 December 2016 18:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Control Rod Not Withdrawing as Expected | On 12/30/16 at 1302 EST, Unit 1 began withdrawing control bank rods for an approach to criticality following a refueling outage. At 1305 EST, operators observed that control rod H-6 did not withdraw. Operators entered the applicable Abnormal Operating Procedure. Operators tripped the reactor as required by plant procedures and entered applicable Emergency Operating Procedures. The act of manually tripping the reactor generated a valid trip signal in the plant Reactor Protection System. Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater was already supplying the Steam Generators; a Feedwater Isolation occurred due to plant conditions. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators, and decay heat removal via the steam dumps. Method of decay heat removal is Steam Generators via the steam dumps. Current reactor coolant system conditions: Temperature at 548 degrees F and stable, pressure 2235 psig and stable. All control and shutdown banks are inserted. Electrical alignment is normal, supplied by offsite power. No impact to Unit 2, Unit 2 is in Mode 1 at 100 percent power. The licensee notified the NRC Resident Inspector. |
| ENS 52327 | 28 October 2016 14:24:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Shutdown Cooling System Declared Inoperable | At 0851 (EDT) on October 28, 2016, Division 1 RHR was started in shutdown cooling (SDC) mode of operation. Prior to starting the RHR system, the Alternate Decay Heat Removal (ADHR) system was maintaining RPV and Spent Fuel Pool temperature. At 0924 on October 28, 2016, RHR (pump A) tripped due to RHR-MOV-17 (SDC suction valve) closing. This is considered to be an event or condition that could have prevented fulfillment of a safety function, and is reportable under 10 CFR 50.72(b)(3)(v)(B). RHR SDC subsystem A was declared inoperable. CNS (Cooper Nuclear Station) entered LCO 3.9.7, Condition A - Required Action A.1: Verify an alternate method of decay heat removal is available within 1 hour and once per 24 hours thereafter; Condition C - Required Action C.1: Verify reactor coolant circulation by an alternate method within 1 hour from discovery of no reactor coolant circulation and once per 12 hours thereafter, and Required Action C.2: Monitor reactor coolant temperature hourly. All LCO conditions specified have been met. ADHR remained in service throughout the event and the plant remained aligned for natural circulation. Spent fuel pool weir temperature monitoring was commenced to verify natural circulation. No increase in RPV (reactor pressure vessel) temperature has been observed. There was no impact to plant operations. Initial investigation indicates that installation of PCIS relay K27 during a maintenance activity physically agitated the adjacent relay, K30, which actuated and caused RHR-MOV-17 to close. The NRC Resident Inspectors have been informed. |