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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56632 | 21 July 2023 15:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 1148 EDT on 07/21/2023, with Unit 3 in Mode 1 at 32 percent power, the reactor automatically tripped on low reactor coolant pump (RCP) speed due to decaying RCP motor voltage during power ascension testing. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to the atmosphere, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 are 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Reactor Protection System | |
| ENS 56614 | 9 July 2023 17:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 1328 EDT on 07/09/2023, with Unit 3 in Mode 1 at 45 percent power, the reactor automatically tripped during power ascension testing due to low reactor coolant flow from decaying voltage to the reactor coolant pumps. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to the atmosphere, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 are 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Reactor Protection System | |
| ENS 56497 | 2 May 2023 08:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip | The following information was provided by the licensee via email: At 0423 EDT on 05/02/2023, with Unit 3 in Mode 1 at 14 percent power, the reactor was manually tripped due to securing all main feed pumps, due to sudden high differential pressure on their suction strainers. The trip was not complex, with all safety-related systems responding normally post-trip. No equipment was inoperable prior to the event that contributed to the event or adversely impacted plant response to the reactor trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the steam dumps, and startup feedwater is supplying the steam generators. Units 1, 2, and 4 were 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Reactor Protection System Main Condenser | |
| ENS 56460 | 10 April 2023 04:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 0048 EDT on 4/10/2023, with Unit 3 in Mode 1 at 18 percent power, the reactor automatically tripped due to low reactor coolant flow due to voltage decaying to the reactor coolant pumps during main generator testing activities. The trip was not complex, with all safety-related systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam via steam generator power operated relief valves to atmosphere. Units 1, 2, and 4 are 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Reactor Protection System | |
| ENS 56414 | 16 March 2023 01:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 2157 EDT on 03/15/2023, with Unit 3 in Mode 1 at 18 percent power, the reactor automatically tripped due to the loss of two reactor coolant pumps when their electrical buses failed to transfer after a main generator excitation protective relay tripped. Operations responded and stabilized the plant. Decay heat is being removed by steam generator power operated relief valves. Units 1, 2, and 4 are not affected. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, nonemergency 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. | Steam Generator Reactor Protection System | |
| ENS 56311 | 14 January 2023 12:21:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Actuation of Reactor Protection System (RPS) | The following information was provided by the licensee via email: At 0721 EST on 01/14/2023, with Unit 3 in Mode 3 at 0 percent power and reactor trip breakers open, a manual actuation of the RPS occurred while conducting pre-criticality testing. The RPS manual actuation was procedurally driven in response to low gland steam pressure, resulting in the necessity to break condenser vacuum following a trip of the auxiliary boiler. The reactor trip breakers were in an open state at the time of the event. 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 RPS. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Protection System | |
| ENS 56294 | 4 January 2023 03:59:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Offsite Power Source | The following information was provided by the licensee via email: At 2259 EST on 1/3/2023, with Unit 1 and Unit 2 in Mode 1 at 100 percent power, an actuation of the Unit 1 B and Unit 2 A emergency diesel generator (EDG) systems, as well as an actuation of the associated auxiliary feedwater (AFW) systems on each unit occurred. The reason for the EDG auto-starts was due to a loss of an offsite power source (loss of one of the two reserve auxiliary transformers (RAT) on each unit) to the Unit 1 B and Unit 2 A safety related buses. The EDG and AFW systems automatically started as designed when the valid undervoltage signal on the affected safety related bus was received. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Emergency Diesel Generator and the Auxiliary Feedwater Systems for both Unit 1 and Unit 2. 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 Auxiliary Feedwater | |
| ENS 56147 | 6 October 2022 06:44:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Reactor Protection System (RPS) | The following information was provided by the licensee via email: At 0244 EDT on 10/06/2022, with Unit 3 Defueled at 0 percent power, an actuation of the RPS occurred during restoration of Division B Class 1E DC and uninterruptible power supply system. The reason for the RPS actuation was due to the opening of the Division B passive residual heat removal (PRHR) heat exchanger outlet flow control valve. The reactor trip breakers were in an open state at the time of the event when the RPS 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 RPS. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Protection System Residual Heat Removal | |
| ENS 55875 | 3 May 2022 19:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Automatic Actuation of Auxiliary Feedwater System | The following information was provided by the licensee via email: At 1541 EDT on May 3, 2022, with Unit 1 in Mode 1 at 100 power, the reactor was manually tripped due to the loss of one of the main feed pumps. The trip was not complex, with all systems responding normally post-trip. No equipment was inoperable prior to the event that contributed to the event or adversely impacted plant response to the scram. Operations responded and stabilized the plant. Decay heat is being removed by Auxiliary Feedwater through the steam dumps to the condenser. Unit 2 is not affected. An automatic actuation of the Auxiliary Feedwater System (AFW) also occurred. The AFW auto-start is an expected response from the reactor trip. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an eight-hour, nonemergency notification per 10 CFR 50.72(b)(3)(iv)(A) as 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. | Reactor Protection System Auxiliary Feedwater | |
| ENS 54995 | 12 November 2020 22:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Turbine Trip | At 1732 EST on November 12, 2020, with Unit 2 in Mode 1 and 100 percent power, the reactor automatically tripped due to a 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 steam generators through the steam dumps into the condenser. Unit 1 is not affected. An automatic actuation of the Auxiliary Feedwater System (AFW) also occurred. The AFW auto-start is an expected response from the reactor trip. Due to the Reactor Protection System actuation while critical, this event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B), as well as in accordance with 10 CFR 50.72(b)(3)(iv)(A) as 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. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 54596 | 21 March 2020 20:44:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of the 1B Emergency Diesel Generator | At 1644 EDT with Unit 1 in Mode 6 at 0 percent power, an actuation of the Unit 1 Bravo Train Emergency Diesel Generator system (EDG) occurred during Engineered Safety Feature Actuation System (ESFAS) testing. The reason for the EDG auto-start signal was a loss of voltage on the Bravo train safety related electrical bus due to the EDG output breaker opening. The EDG was already running at the time of the loss of voltage on the bus. 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 EDG system. 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. | Emergency Diesel Generator | |
| ENS 54318 | 9 October 2019 14:23:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of Emergency Diesel Generator and Auxilliary Feedwater System | At 1023 EDT, on October 9, 2019, with Unit 2 in Mode 1 and 100 percent power, an actuation of the Emergency Diesel Generator and Auxiliary Feedwater Systems occurred. The reason for the Emergency Diesel Generator auto-start was the loss of power to the 4160V 1E electrical bus 2AA02 due to a fault at an offsite electrical switchyard. The Emergency Diesel Generator started and energized the 4160V safety bus, and Auxiliary Feedwater Systems automatically started as designed when the undervoltage condition on the safety bus was detected. 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 and Auxiliary Feedwater Systems. There was no impact on the health and safety of the public or plant personnel. The NRC resident has been notified. | Feedwater Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 54175 | 19 July 2019 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 Reactor Trip Due to Main Steam Isolation Valve Failing Shut | At 0945 (EDT) on July 19, 2019, with Unit 2 in Mode 1 and 100 percent power, the reactor automatically tripped due to Loop 2 'B' Main Steam Isolation Valve failing shut. The Auxiliary Feedwater system (AFW) started automatically as a result of the automatic reactor trip. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser. 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). Due to the valid AFW actuation from the reactor trip, this event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 1 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods fully inserted. | Reactor Protection System Main Steam Isolation Valve Auxiliary Feedwater Main Steam Line Control Rod | |
| ENS 53967 | 31 March 2019 01:30:00 | 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 Msiv Failing Closed | At 2130 (EDT) on March 30, 2019, with Unit 2 in Mode 1 at 30 percent reactor power, the reactor was manually tripped due to a main steam isolation valve failing closed. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser. 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). Unit 1 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. | Main Steam Isolation Valve Auxiliary Feedwater Main Steam Line | |
| ENS 53643 | 4 October 2018 04:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | En Revision Imported Date 10/22/2018 | EN Revision Text: MANUAL REACTOR TRIP DURING LOW POWER PHYSICS TESTING At 0544 EDT on October 4, 2018, with Unit 1 in Mode 2 with reactor power in the intermediate range performing low power physics testing, the reactor was manually tripped due to a rod control urgent failure alarm. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam system. Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods inserted as expected. The cause of the rod control urgent failure is being investigated.
This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A). During Dynamic Rod Worth Measurement testing, Control Bank Charlie was inserted approximately 153 steps when the urgent failure occurred (CBC positioned at 75 steps out). Following the scram, additional analysis concluded that the reactor was subcritical when the Reactor Protection System was actuated." The licensee notified the NRC Resident Inspector. Notified the R2DO (McCoy). | Reactor Protection System Control Rod Main Steam | |
| ENS 53484 | 3 July 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 Due to High Steam Generator Water Level | At 0954 (EDT) on July 3, 2018, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to high steam generator water level. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser. 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). Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods inserted and Unit 1 is in an electrical shutdown lineup. The cause of the high steam generator water level transient is being investigated. | Steam Generator Auxiliary Feedwater Main Steam Line Control Rod | |
| 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. | Emergency Diesel Generator Decay Heat Removal | |
| ENS 52619 | 17 March 2017 19:17:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Emergency Diesel Generator | At approximately 1517 (EDT), while restoring protective relay power to the 1B Reserve Auxiliary Transformer, a valid undervoltage actuation signal was sent to the 1B Emergency Diesel Generator (EDG). The 1B EDG automatically started and tied to the safety bus (1BA03). The 1BA03 bus was loaded by the automatic load sequencer. This actuation was identified by the Control Room Operators and the 1B 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. Containment Cooler Number 8 did not automatically start in 'Fast Speed' as expected. Containment Cooler Number 8 was successfully started in 'Fast Speed' manually by the Control Room Operators. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 52534 | 3 February 2017 20:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Manual Reactor Trip Due to Loop 1 Msiv Starting to Fail Closed | At 1545 EST on 2/3/17, Vogtle Unit 1 was manually tripped from 100% power when loop 1 Main Steam Isolation Valve (MSIV) started to fail closed. Non-Safety Related 4160V bus 1NA01 failed to transfer to alternate incoming power supply automatically and was successfully manually energized. All control rods fully inserted and AFW (Auxiliary Feedwater) and FWI (Feedwater Isolation) actuated as expected. Unit 1 is in Mode 3 and stable with decay heat being removed by AFW. The licensee informed the NRC Resident Inspector. | Main Steam Isolation Valve Control Rod | |
| ENS 51956 | 25 May 2016 06:06:00 | 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 Lowering Steam Generator Water Level | At 0206 EDT 5/25/16, Vogtle Unit 2 tripped from 100% when SG (Steam Generator) #1 Level began to lower for an unknown reason. Cause for level issue is under investigation. All control rods fully inserted and AFW (Auxiliary Feedwater) and FWI (Feedwater Isolation) actuated as expected. Unit 2 is in Mode 3 and stable with decay heat being removed by Aux Feedwater. Prior to the trip, I & C (Instrumentation & Calibration) was performing a loop #1 narrow range instrument calibration. Unit 2 is in a normal post trip electrical lineup with all source of offsite power available. The licensee informed the NRC Resident Inspector. | Steam Generator Feedwater Control Rod | |
| ENS 50894 | 14 March 2015 16:07:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Actuation During Surveillance Test | On March 14, 2015 at 1207 EDT, operators were performing steps to disable an automatic actuation signal to the B-train Auxiliary Feedwater system (AFW), when a valid actuation signal was received on B-train only. Both motor driven Auxiliary Feedwater pumps were already running and feeding forward. The B-train discharge valves went from throttled to fully open. The AFW discharge valves were restored to their previous positions without any adverse impacts on the plant. Decay heat removal was still being removed through the Atmospheric Relief Valves. The licensee notified the NRC Resident Inspector. | Auxiliary Feedwater Decay Heat Removal | |
| ENS 50893 | 14 March 2015 08:29:00 | 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 | Automatic Reactor Trip on System Injection Actuation Signal | Vogtle Unit 2 was operating in Mode 1 at 100% rated thermal power. At 0429 EDT a Unit 2 automatic Reactor Trip and Safety Injection / Steamline Isolation occurred. All systems operated as expected and all control rods fully inserted. The Safety Injection was terminated at 0447 EDT and the emergency operating procedures were exited at 0522 EDT. Unit 2 is stable in Mode 3 with decay heat removal via the Auxiliary Feedwater system and the atmospheric relief valves. A response team is investigating the cause of the event. Unit 1 was unaffected by the event. NRC Senior Resident Inspector was notified and is at plant site for investigation. | Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 50526 | 12 October 2014 13:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip During Reactor Startup | VEGP (Vogtle Electric Generating Plant) Unit 2 was performing startup and had taken reactor critical at 0929 EDT. When attempting to stabilize power to collect critical data, control rods were inserted with Control Bank D the expected group to insert. Control Bank A inserted instead of Control Bank D. Power had reached 6 E-2 percent as indicated by IR (intermediate range) indication when control room crew performed a manual reactor trip. AFW (auxiliary feed water) was in service to support plant conditions prior to the trip and did not receive any actuation signal. All equipment operated as expected. Unit 2 is currently stable in Mode 3 at normal operating temperature and pressure. The licensee has notified the NRC Resident Inspector. | Control Rod | |
| ENS 50314 | 27 July 2014 18:09:00 | 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 Loss of Main Feed Pump During Testing | VEGP (Vogtle Electric Generation Plant) Unit One was at 100 percent power, with a Main Feed Pump (MFP) Turbine A trip mechanism test in-progress, when MFP A Trip alarm was received in the Main Control Room. Control Room crew identified MFP A speed and steam generator levels lowering and initiated a manual reactor trip. All control rods fully inserted and nothing unusual was noted. Auxiliary feed water and feedwater isolation actuated as expected. The unit is currently stable in MODE 3 at normal operating temperature and pressure. A forced outage response team has been formed to determine the cause of the MFP A trip and determine restart criteria and time of restart. The unit is in a normal shutdown electric plant lineup. No effect on Unit 2. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 50031 | 13 April 2014 00:08:00 | 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 Main Steam Isolation Valve Failing Shut | At 2008 EDT, Vogtle Unit One was manually tripped in response to loop 1 outboard Main Steam Isolation Valve failing shut. All systems operated correctly in response to the reactor trip. All control rods fully inserted. System response allowed for an uncomplicated reactor trip response. Unit 1 is stable in Mode 3 and cause investigation is in progress. The NRC Senior Resident Inspector was notified. There was a normal post trip feedwater isolation due to low Tave. Offsite power remains available. Decay heat is being removed by the main condenser. The plant is stable in Mode 3. There was no impact on Unit 2. | Feedwater Main Steam Isolation Valve Main Condenser Control Rod | |
| ENS 50006 | 8 April 2014 08:28:00 | 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 Low Steam Generator Level | VEGP Unit 2 was at 100% power, normal activities, when digital feedwater trouble alarms were received on all 4 steam generators (SG) with level stable in all generators. Operating crew entered abnormal operating procedure for feedwater malfunction when SG #3 level began rapidly lowering. Operators attempted to take manual control of SG #3 main feedwater regulating valve and were unable to raise SG #3 level. SG #3 level lowered to the Lo-Lo Level setpoint causing an automatic reactor trip. All control rods fully inserted and SG #3 level remained off scale low on narrow range indications. Auxiliary feedwater and feedwater isolation actuated as expected. (Unit 2) is currently stable in Mode 3 at normal operating temperature and pressure. A forced outage response team has been formed to determine the cause of the low SG water level and determine restart criteria and time of restart. All control rods fully inserted on the trip. Decay heat is being removed via auxiliary feedwater to steam generators steaming to the condenser steam dumps. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 49461 | 22 October 2013 15:44:00 | 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 Decreasing Condenser Vacuum | At 1144 EDT, Vogtle Unit Two was manually tripped in response to lowering Main Condenser vacuum. The Unit 2 Bravo Main Feed Pump was tagged out for scheduled maintenance and the casing was being removed when condenser vacuum started lower due to isolation valve not holding pressure. Main Condenser vacuum lowered to less than procedural limits for continued plant operation. All systems operated correctly in response to the reactor trip. All control rods fully inserted. AFW was placed in service to control Steam Generator levels. System response allowed for an uncomplicated reactor trip response. Plant is stable in Mode 3 while performing a cause investigation. The NRC Senior Resident Inspector was notified and at plant site. The plant is in a normal post-trip electrical line-up. Decay heat is being removed via the steam dumps to the Main Condenser. | Steam Generator Main Condenser Control Rod | |
| ENS 49453 | 19 October 2013 10:43:00 | 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 | VEGP (Vogtle Electric Generating Plant) unit two was at 100% power, normal activities, when the unit two turbine tripped causing an automatic reactor trip. All control rods fully inserted and nothing unusual was noted. Auxiliary feed water and feedwater isolation actuated as expected. The unit is currently stable in Mode 3 at normal operating temperature and pressure. A forced outage response team has been formed to determine the cause of the turbine trip and determine restart criteria and time of restart. During the transient, no relief valves lifted. The electrical grid is stable and supplying plant safety loads. Decay heat is being removed via steam dumps to the condenser. The licensee has notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Control Rod | |
| ENS 48788 | 27 February 2013 04:02:00 | 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 Excessive Reactor Coolant Pump Seal Leakoff Flow | At 2302 EST, Vogtle Unit Two was manually tripped in response to excessive Reactor Coolant Pump #4, seal #1 leakoff flow. Seal leakoff flow exceeded the procedural limits for continued operation of the pump. Following the reactor trip, RCP #4 was shutdown per procedure guidance. All systems operated correctly in response to the reactor trip. All control rods fully inserted. The Auxiliary Feed Water (AFW) system automatically actuated as expected. System responses allowed for an uncomplicated reactor trip response. The plant is stable in Mode 3 during cause investigation. The NRC Senior Resident was notified and is enroute to the plant for investigation. AFW is supplying the steam generators and decay heat removal is to the condenser via steam dumps. No safety valves or relief valves lifted during the transient. The unit is in a normal post-trip electrical line-up. There was no impact on Unit One. | Steam Generator Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 48385 | 5 October 2012 09:14:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Auxiliary Feedwater Actuation | On October 5, 2012 at 05:14 EDT, restoration of steam generator narrow range level instrumentation from a bypassed condition was in progress when a valid Auxiliary Feedwater Actuation signal was received due to steam generator levels being below the Lo-Lo Level setpoint. As a result, both motor driven auxiliary feedpumps automatically started and two turbine driven auxiliary feedwater discharge valves automatically opened. The turbine driven auxiliary feedwater pump and the motor driven auxiliary feedwater discharge valves had been previously removed from service under administrative controls. The system was aligned per procedure so that water injection into the steam generators did not occur and the motor driven pumps were operated on mini-flow. Unit 1 is off line for a planned refueling and maintenance outage. There were no adverse impacts on the plant. At the time of the event, Unit 1 was in Mode 5 with RHR in service and preparations for Mode 4 entry in progress. The NRC Resident Inspector has been informed. | Steam Generator Auxiliary Feedwater | |
| ENS 47836 | 14 April 2012 17:46:00 | 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 Feedwater Flow | At 1346 EDT, Vogtle Unit 1 reactor was manually tripped from 100% power due to Main Feedwater Pump 'B' discharge flow lowering unexpectedly. All control rods fully inserted. AFW system automatically actuated as expected. System responses allowed for an uncomplicated reactor trip response. Plant is stable in Mode 3 during cause investigation. The electrical lineup remained normal. No safety valves lifted due to the trip. Decay heat is being removed via the steam dumps to the main condenser. The licensee has notified the NRC Resident Inspectors. | Feedwater Main Condenser Control Rod | 05000424/LER-2012-002 |
| ENS 47224 | 31 August 2011 13:06:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip on Hi-Hi Steam Generator Level Results in Automatic Reactor Trip | While preparing Loop 2 Main Feed Reg Valve (MFRV) for maintenance, it was placed on its 'air gag' that maintained the MFRV in position, while any small changes in feed flow would be modulated by the associated Bypass Feed Reg Valve (BFRV). Approximately 5 minutes after the MFRV was placed on the air gag, with Steam Generator (S/G) level stable, control room operators observed S/G level start to increase. The operators observed for a short time to see if the associated BFRV would control the level change. When it became apparent that level was not being controlled automatically, the operators took manual control of the BFRV, eventually closing it all the way, and observed that S/G level was then increasing very slowly. While level was still slowly rising, two hi-hi level bistables actuated, generating a P-14 (hi-hi S/G level trip) signal which tripped the main turbine, which then caused an automatic reactor trip. As a result of the reactor trip, all systems functioned as required and there was nothing unusual or not understood. During the transient, no safeties, primary relief valves or secondary relief valves lifted. All control rods inserted into the core. Auxiliary Feed Water automatically initiated and is supplying the steam generators. Decay heat is being removed via the steam dumps to condenser. The grid is stable with all safety buses powered from offsite power via a normal shutdown electrical lineup. Unit 2 was unaffected by the trip. The licensee will be issuing a press release. The NRC Resident Inspector has been notified. | Steam Generator Auxiliary Feedwater Main Turbine Control Rod | |
| ENS 46772 | 20 April 2011 21:34:00 | 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 1734 EDT, (Vogtle) unit one automatically tripped from 100% power. No significant activities were in progress that should have challenged the Reactor Protection System. All control rods fully inserted. AFW system actuated as expected on S/G Lo-Lo-Level and AMSAC (ATWS Mitigation System Actuation). System responses allowed for an uncomplicated reactor trip response. Plant is stable and will remain in Mode 3 during cause investigation. Cause of the reactor trip is under investigation. The plant is in its normal shutdown electrical lineup. Decay heat is being sent to the main condenser through the turbine bypass valves. The steam generators are being fed from auxiliary feedwater. There was no effect on unit two. The licensee informed the NRC Resident Inspector. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser Control Rod | 05000424/LER-2011-001 |
| ENS 45588 | 23 December 2009 20:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Rps Actuation Due to Loss of Instrument Air | At 1525 EST, Vogtle Unit 2 was manually tripped from 100% power due to a loss of instrument air to the turbine building. System operators were releasing a tagout and restoring one of two instrument air dryers that had been isolated for maintenance. Instrument air low pressure alarms were received in the control room and secondary side valves were responding to the loss of instrument air. Control room operators responded according to procedures. Main feed pump 'B' tripped on a loss of suction pressure and operators manually tripped the reactor. The reactor was manually tripped in anticipation of a loss of feed water to the steam generators. All systems responded as required. AFW (Auxiliary Feed Water) actuated as required for loss of feed water. All control rods fully inserted on the reactor trip. Instrument air has been restored to the turbine building and steam dumps are controlling RCS temperatures. Cause of the loss of instrument air is being investigated. (The NRC) Senior Resident (Inspector) was notified. | Steam Generator Control Rod | |
| ENS 45557 | 10 December 2009 04:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram Following High Turbine Vibration | At 2310 EST, Vogtle Unit 1 was manually tripped from 24% reactor power while the main turbine was rolling at 1800 rpm, preparing for synchronization to the grid. As Vogtle 1 was preparing to bring the Unit 1 generator on line following a forced outage, high vibration levels were experienced on the HP turbine bearings while the Turbine was at rated speed and synchronization preparations were in progress. The Turbine was manually tripped in accordance with plant procedures. Vibrations continued to increase as the Turbine began to coast down, warranting that vacuum be broken in accordance with procedures. The reactor was manually tripped in anticipation of trip of the main feedwater pump (due to loss of condenser vacuum) and condenser vacuum was broken to slow the turbine. When condenser vacuum was broken, the in-service Main Feedwater Pump auto tripped as expected, causing an automatic actuation of the Motor Driven Auxiliary Feedwater system. The cause of high vibrations on the Turbine is being investigated. All systems responded as expected on the trip. All control rods fully inserted into the core following the reactor trip. Atmospheric relief valves are being used to remove decay heat. There is no known primary to secondary leakage. The plant is in a normal post-trip electrical line-up. The licensee notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Main Turbine Control Rod | |
| ENS 45547 | 7 December 2009 23:01:00 | 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 a Turbine Trip Caused by Low Condenser Vacuum | Vogtle Unit 1 tripped from 100% power due to a (turbine trip/ reactor trip) RPS (reactor protection system) actuation. The turbine tripped due to low condenser vacuum. Initial investigation indicates that a loss of a non-1E electrical switchgear initiated the event. All systems responded as expected. The AFW (auxiliary feedwater) systems responded as required. Reactor temperature (and decay heat removal) is being maintained on SG (steam generator) ARVs (atmospheric relief valves). Both NRC Resident Inspectors were notified of the trip. There were no complications. All rods inserted during the trip and there was no primary to secondary leakage. There was no impact on Unit 2. | ||
| ENS 43313 | 23 April 2007 14:24:00 | 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 Generator Ground Relay Actuation | Vogtle Unit 2 tripped from 53% power due to a turbine trip /P9 reactor trip RPS actuation during power ascension following completion of refueling outage 2R12. Initial investigation indicates that a generator neutral ground relay actuated causing an automatic main generator turbine trip. No visual damage is apparent on any plant equipment. All systems responded as expected. Both motor driven auxiliary feedwater pumps and the turbine driven started on lo-lo steam generator level and AMSAC. The NRC Resident Inspector was notified. All control rods fully inserted upon RPS actuation, the atmospheric relief valves lifted momentarily and reseated as expected, and no safety valves lifted. After the trip, steam generator level was being maintained with auxiliary feed pumps and steam was being dumped to the condenser. The plant was placed in the normal shutdown electrical lineup. | Steam Generator Auxiliary Feedwater Control Rod | |
| ENS 42942 | 28 October 2006 16:23:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation | During plant cooldown/depressurization in Mode 3, an Over Temperature (OT) Delta Temperature reactor trip signal was generated. The reactor Trip Breakers opened as designed (all rods were previously inserted). Additionally, a low Tavg/P-4 Feedwater Isolation Signal was generated and all Bypass Feedwater Regulation Valves closed (open for long-cycle recirculation operation). No Feedwater Isolation Valves were open at the time. All systems performed as designed. Actual loop Delta T's (Th - Tc) never exceeded 3.3 degrees F, well below the calculated OT Delta T setpoint. But plant Delta T calculations are based on narrow range Th and Tc instruments with scaling to low limits. When actual plant temperature was lowered to the lower limits of each instrument, calculated Delta T increased to the OT Delta T setpoint causing the reactor trip signal. Existing procedure guidance did not adequately ensure that the reactor trip breakers are open prior to initiating a partial plant cooldown. The plant was stabilized. The long-cycle recirculation was re-established and plant cooldown/depressurization was recommenced, as originally planned to 340 degrees F and 925 psig. The licensee notified the NRC Resident Inspector.
Alvin W. Vogtle, Unit 1, Operating License no. NPF-68, Event Notification #42942 is retracted. The event reported in the notification which occurred during the Unit 1 cooldown in Mode 3 was not representative of the conditions under which the Over Temperature Delta Temperature (OTDT) setpoint is intended to operate. The setpoint is intended to provide protection while the reactor is in Modes 1 and 2. As the cooldown progressed from Normal Operating Pressure and Temperature (NOPT), eventually the Reactor Coolant System (RCS) hot and cold leg temperature indications reached the low end of their ranges or scales. Further reductions of temperature resulted in artificial indications of core power (delta-T) that cannot be relied upon as a valid input to the OTDT protective function. The actuation of the Reactor Protection System (RPS) is not considered to be a valid actuation. As such, this is not reportable under 10 CFR 50.72(b)(3)(iv)(A). The Unit 1 Feedwater Isolation caused from the OTDT reactor trip is not a function is listed in 10 CFR 50.72(b) (3)(iv)(B) as a system whose actuation is required to be reported under 10 CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector. | Reactor Coolant System Feedwater Reactor Protection System | |
| ENS 42806 | 27 August 2006 10:31:00 | 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 Reactor Coolant Pump Trip | On Sunday, August 27, 2006 at 0631 EDST, Unit Two was at 99.7% RTP when RCP #4 tripped generating a Low Flow Reactor Trip. All systems functioned as required. AFWAS (Aux Feedwater Actuation Signal) was actuated as expected due to lo-lo Steam Generators levels. RCS Letdown isolated on a momentary low level signal on one channel of Pressurizer level, and has since been restored. The reactor is currently stable in Mode 3 while the cause of the trip of the RCP is investigated. Decay heat is being rejected to the condenser via the steam dumps. ESF systems remain operable and the electrical grid is stable. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 42506 | 17 April 2006 04:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Manually Tripped Due to Failure of a Main Feed Regulating Valve to Control Steam Generator Level | Unit 1 reactor was manually tripped from 33% RTP due to the Loop 3 Main Feed Regulating Valve (MFRV) 1LV-0530 not controlling steam generator level in manual or AUTO. A unit shutdown was in progress at the time of the trip due to concerns with the Loop 3 MFRV. The unit shutdown had commenced from 100% power on 04/16/2006 at 1603 hrs. All systems functioned as required on the reactor trip including a feedwater isolation and the Loop 3 MFRV closing as expected. The unit is presently in Mode 3 at normal operating temperature and pressure. An investigation team is being assembled concerning the Loop 3 MFRV. All control rods fully inserted. No primary or secondary reliefs/safeties lifted during the transient. Unit 1 is currently in Mode 3 Hot Standby controlling Steam Generator Water Level using Auxiliary Feedwater supplied by both Motor Driven and the Steam Driven pumps. Decay Heat is being removed via the Bypass Valves to the Main Condenser. Offsite power is stable and all EDGs are available, if necessary. The licensee informed the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 42058 | 17 October 2005 22:16:00 | 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 Steam Generator Level | At 1816 EDT on 10/17/2005 Vogtle Unit 1 was manually tripped from 100% power due to lowering Steam Generator level on Loop 2. The Main Feedwater Regulating Valve for Loop 2 failed closed and operator attempts to re-open it were unsuccessful. The operators initiated a manual reactor trip when it was apparent that Steam Generator level would not be restored. Following the manual reactor trip, an automatic actuation of the Motor Driven and Turbine Driven Auxiliary Feedwater Pumps occurred due to low level in the Steam Generators. The Main Feedwater Regulating Valve for Loop 1 did not close as expected for the feedwater isolation signal (P-4 / Tavg 564 degrees F) that resulted from the manual trip. The Loop 1 Main Feedwater Regulating Valve was manually closed by the operators. All control rods are fully inserted. This incident did not affect Unit 2. Unit 1 is stable in Mode 3 and removing heat by dumping steam to the condensers. All safety related systems or equipment are available and functioning as required. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 41653 | 30 April 2005 01:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip - Lowering Steam Generator Level on Loop #1 | At 2155 EDT on 4/29/2005, Vogtle Unit 1 was manually tripped from 100% power due to lowering steam generator level on loop #1. The main feedwater regulating valve was in manual control and a repair plan was in progress to replace a controlling card which failed earlier in the day. The manual reactor trip caused an automatic aux. feedwater actuation of the motor driven and turbine driven feedwater pumps. All other equipment responded as expected on the trip. At approximately 1600, the loop #1 main feed regulating valve had failed shut while in the automatic mode of operation. The operator shifted control to manual and opened the valve, preventing a reactor trip. At 2155 the recovery plan was being implemented using a plant procedure. While performing the procedure, the loop #1 feed regulating valve shut. The reactor was manually tripped on lowering steam generator level. All rods fully inserted after the manual reactor trip. Decay heat removal is to the main condenser with steam generator level being maintained using the motor driven aux. feedwater pumps. The plant is in its normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Decay Heat Removal Main Condenser | |
| ENS 41323 | 11 January 2005 12:19:00 | 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 Turbine Trip | At 0719 EST, Vogtle Unit 1 tripped from 100% power due to a Turbine Trip/P9 Reactor Trip, RPS actuation. Initial investigation indicates that Generator relaying was involved. The investigation is ongoing. As expected for the trip, both Motor Driven Auxiliary Feedwater pumps and the Turbine Driven pump started on Lo-Lo Steam Generator level and AMSAC. All equipment actuated as expected except the Group A Pressurizer Heaters tripped. Technical Specification 3.4.9 Condition B was entered. All control rods fully inserted. Decay heat is being removed via the steam dumps to the main condenser. The electrical grid is stable. The licensee notified the NRC Resident Inspectors. | Steam Generator Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 41213 | 20 November 2004 16:40:00 | 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 | Unexpected Automatic Reactor Trip During Surveillance Testing with Si Injection | On 11/20/2004 at 1140 EST during the performance of 14421-2 'Solid State Protection System And Reactor Trip Breaker Train B Operability Test' an automatic Reactor Trip occurred and a Safety Injection occurred a short time thereafter. The Reactor Trip was caused by an error made in the performance of the 14421-2 procedure. The operator working with a peer checker mistakenly placed the 'A' train multiplexer test switch in the 'A + B' position instead of the 'B' train one. The 'B' train multiplexer test switch was still in the inhibit position. When the 'A' Train multiplexer switch went through the inhibit position, it caused the second general warning which tripped the reactor. The Safety Injection is believed to have been caused by a failure of the loop 2 Reactor Coolant Average Temperature which impacted the Steam Dump control system. The Steam Dumps did not close as expected during the Reactor Trip response and Pressurizer pressure lowered below the Safety Injection System setpoint and a Safety Injection was actuated. All systems operated as expected during the Safety Injection and Reactor Trip. All the Control Rods fully inserted on the Reactor Trip. The Main Steam Lines were isolated by the operator due to the lower than expected decrease of the Reactor Coolant System Average Temperature. Both Motor Driven AFW pumps started as expected as did the Turbine Driven AFW pump. All ECCS pumps started as expected on the Safety Injection. The Containment Coolers all started in low speed on the Safety Injection as expected. Both Diesel Generators started on the Safety Injection actuation. The Containment Isolation systems isolated containment as expected. During the recovery from the Safety Injection pressurizer level did reach 100%, but the Pressurizer PORVs were not required to open to control Pressurizer pressure. Unit 2 is currently stable in mode 3 removing decay heat via the Atmospheric Steam Dumps. The licensee informed the NRC Resident Inspector. | Reactor Coolant System Main Steam Line Control Rod | |
| ENS 40682 | 18 April 2004 16:38:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Auxiliary Feedwater System Actuation Due to Main Feed Pump Trips | An Auxiliary Feedwater (AFW) actuation occurred with the plant in Mode 3, Main Feedwater out of service and a cooldown in progress. An AFW actuation was received when both main feed pumps tripped due to low condenser vacuum. AFW was already in service with both motor-driven pumps running and discharge valves throttled to maintain steam generator level. A surveillance to test AFW actuation on the trip of main feedwater pumps was in progress. The main feed pump trip signal was generated when condenser vacuum was broken prior to completing the test and blocking the actuation signal from the main feed pumps to AFW. Steam generator blowdown and sample valves isolated and MDAFW (Motor Driven AFW) discharge valves opened as expected. Operators took action to throttle AFW feed to the steam generators. No adverse affects on the plant occurred. The AFW actuation signal from the trip of main feed pumps has been blocked as allowed by plant conditions. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 40615 | 28 March 2004 03:05:00 | 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 Feedwater Pump Speed Control | On 3/27/2004, the Unit 1 Reactor was manually tripped when the 1B Main Feedwater Pump speed could not be controlled in automatic or manual. An unexpected increase in Main Feedwater Pump speed was observed which increased the Main Feedwater header pressure to higher than expected values. An attempt was made to manually control the speed of the 1B Main Feedwater Pump from the various controllers in the Main Control Room. Preparations to start the 1A Main Feedwater Pump were initiated. The speed of the 1B Main Feedwater Pump continued to increase to a point where level control of the Steam Generators and overspeed of the 1B Main Feedwater Pump became a concern. At this point the Reactor was manually tripped and the 1B Main Feedwater Pump was tripped. An Auxiliary Feedwater System automatic actuation occurred on the trip a the 1B Main Feedwater Pump as expected because the 1A Main Feedwater Pump was also tripped at the time due to being at a low power level. All systems responded as expected on the Reactor Trip. The plant is currently stable in Mode 3. An Event Review Team will be performing a review of the event and making recommendations related to restarting the Reactor. All control rods fully inserted. Decay heat is being removed using the steam dumps and auxiliary feedwater. Plant pressure is 2235 psig and temperature is 557 degrees F. The licensee notified the NRC Resident Inspector. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 40268 | 22 October 2003 05:03:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | High Pressure Hose Connection Failed Resulting in a Train "a" Manual Steam Line Isolation | Manually initiated a steam line isolation signal upon report of a failed connection on a high pressure hose being used to vent a main steam isolation valve bonnet cavity (to open pressure locked valves). The manual isolation resulted in closure of the four open main steam isolation valves (Train "A"). The isolation was conducted as a preventative measure for personnel protection (three people were in the South Main Steam Valve Room when the fitting failed). No personnel were injured. Plant conditions remained stable. The NRC Resident Inspector was notified of this event by the licensee. | Main Steam Isolation Valve Main Steam |