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The query [[Category:ENS Notification]] [[Site.Company::Southern Nuclear]] [[Scram::+]] was answered by the SMWSQLStore3 in 0.8372 seconds.


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 Entered dateSiteScramRegionReactor typeEvent description
ENS 5417519 July 2019 13:05:00VogtleAutomatic ScramNRC Region 2At 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.
ENS 540401 May 2019 21:03:00FarleyManual ScramNRC Region 2

EN Revision Text: MANUAL REACTOR TRIP DUE TO MISALIGNED CONTROL ROD At 1643 (CDT), with Unit 2 in Mode 2 during low power physics testing, the reactor was manually tripped per procedure due to a misaligned control rod. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the atmosphere using the atmospheric relief valves. Unit 1 is not affected. Due to the Reactor Protection System actuation, 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.

  • * * UPDATE ON 05/08/2019 AT 1212 EDT FROM MIKE CONNER TO JEFFREY WHITED * * *

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). The reactor was tripped during low power physics testing. The misaligned rod was encountered during rod group insertion and the affected bank had been inserted to the extent that the reactor was subcritical when the operators tripped the reactor. The licensee notified the NRC Resident Inspector. Notified R2DO (Lopez)

ENS 5396731 March 2019 00:17:00VogtleManual ScramNRC Region 2At 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."
ENS 5395324 March 2019 05:23:00HatchManual ScramNRC Region 2At 0159 (EDT), with Unit 2 in Mode 1 at 25 percent power, the reactor was manually tripped due to degrading condenser vacuum. After the turbine was tripped, the station service electrical buses did not transfer to alternate supply resulting in loss of the condensate feedwater system and level being controlled by the RCIC system. Operators responded and stabilized the plant. Reactor water level is being maintained via the RCIC system. Pressure is being controlled and decay heat is being removed by the HPCI system in pressure control mode. Unit 1 is not affected. Additionally, an actuation of the primary containment isolation system occurred during the reactor scram. The reason for the actuation was a group II isolation signal was received on reactor water level and a group I isolation was received on decreasing vacuum. 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). Additionally, 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 primary containment isolation system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
ENS 536434 October 2018 07:57:00VogtleManual ScramNRC Region 2

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.

  • * * UPDATE FROM KEVIN LOWE TO DONALD NORWOOD AT 1408 EDT ON 10/19/2018 * * *

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).

ENS 534843 July 2018 12:00:00VogtleManual ScramNRC Region 2At 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.
ENS 5269620 April 2017 05:57:00HatchAutomatic ScramNRC Region 2GE-4On 04/20/2017 at 0302 EST during a reactor startup, a reactor scram resulted from upscale spike on two Intermediate Range Monitors (IRMs), 1C51K601A and 1C51K601B. IRM A, 1C51K601A is in Reactor Protection System Channel A and IRM B, 1C51K601B is in Reactor Protection System Channel B. All control rods fully inserted. No PCIS (Primary Containment Isolation System) actuations occurred and none were expected to occur based upon plant condition following the reactor scram. Investigation is in progress. Condition was not due to a true flux event. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) as an event or condition that resulted in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. CR 10356172 The NRC Resident has been notified. The reactor was at 0.5% (percent) power at the time of the event and will remain in Hot Shutdown pending the results of the root cause investigation.
ENS 525343 February 2017 18:56:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 5239527 November 2016 03:08:00FarleyManual ScramNRC Region 2Westinghouse PWR 3-LoopAt 0026 (CST) on November 27, 2016, Farley Unit 1 was manually tripped from 100% reactor power due to voltage swings suspected to be caused by the Auto Voltage Regulator. All control rods fully inserted and Auxiliary Feedwater (AFW) auto-started as expected. All systems responded as expected. The plant is currently stable in Mode 3 (Hot Standby). The cause of the main generator voltage oscillations is under investigation. The NRC Resident Inspector has been notified. The trip was uncomplicated. Decay heat is being removed via the steam dumps to condenser. The plant is at normal operating pressure and temperature with auxiliary feedwater supplying the steam generators. The electrical grid is stable and supplying plant loads. All safety equipment is available, if needed. Unit 2 was unaffected by the event and remains at 100% power.
ENS 523568 November 2016 17:36:00FarleyManual ScramNRC Region 2Westinghouse PWR 3-LoopAt 1331 (CST) on November 8, 2016, Farley Nuclear Plant Unit 1 manually tripped from 32% reactor power. The plant was ramping down to remove the main generator from service due to an unrelated issue. 1A SGFP did not respond to control Steam Generator (SG) level as expected when the miniflow was opened per procedure. SG levels lowered due to lower feed flow and the reactor was manually tripped in accordance with plant procedures. All control rods fully inserted and Auxiliary feedwater (AFW) auto started as expected. The Main Steam Isolation Valves were closed to minimize the cool-down. Decay heat is being removed through the Atmospheric Relief Valves. All other systems responded as expected. The plant is currently stable in Mode 3 (Hot Standby). The failure of the 1A SGFP control is under investigation. Unit 2 was not affected. The NRC Resident Inspector has been notified. There is no primary to secondary leakage.
ENS 522741 October 2016 09:42:00FarleyAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 0512 (CDT) on October 1, 2016, Farley Nuclear Plant Unit 1 automatically tripped from 99 percent reactor power due to the inadvertent closure of a main steam isolation valve (MSIV). The closure of the MSIV caused a turbine trip resulting in an automatic reactor trip. Concurrent with the reactor trip, a safety injection (SI) occurred. The plant is stable in Mode 3 (Hot Standby) and auxiliary feedwater (AFW) autostarted as expected. The cause of MSIV closure and SI actuation is under investigation. Cooldown will continue to Mode 5 (Cold Shutdown) as planned for entry into a scheduled refueling outage. Restart is not planned until the completion of the refueling outage. Unit 2 was not affected. The NRC Resident Inspector has been notified. The MSIVs are open with the steam generators discharging steam to the main condenser using the turbine bypass valves. SI was from high head injection which has been secured.
ENS 5195625 May 2016 04:54:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 5191811 May 2016 10:56:00FarleyManual ScramNRC Region 2Westinghouse PWR 3-LoopAt 0653 (CDT) on 5/11/16, Farley Unit 2 reactor was manually tripped from 29 (percent) power. The initiating event was hi-hi Steam Generator level. Steam Generator levels began to rise following the start of a second condensate pump. The hi-hi steam generator level setpoint was reached causing the only running main feedwater pump to trip, a main feedwater isolation, and an automatic turbine trip. Auxiliary feedwater automatically started as expected. The reactor was manually tripped per procedure. All other systems responded properly for the event and there were no complications. The plant is currently stable in Mode 3. The NRC Resident Inspector has been notified.
ENS 5089314 March 2015 08:07:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopVogtle 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.
ENS 5053314 October 2014 07:37:00FarleyManual ScramNRC Region 2Westinghouse PWR 3-Loop

This notification is being made as required by 10 CFR 50.72(b)(2)(iv)(B) due to a Farley Nuclear Plant Unit 2 manual reactor trip. The trip was initiated when the in service train of CCW cooling to the Reactor Coolant Pumps was lost due to a loss of the 2B Start Up Transformer (SUT). The control room team manually tripped the reactor then tripped all three reactor coolant pumps as required by station procedure. There was a line of severe thunderstorms with lightning passing through the plant site at the time of loss of the 2B Start Up Transformer. The 2B emergency diesel generator was out of service for maintenance therefore there was a loss of the 'B' train emergency power 4160V electrical bus ('B' train LOSP (Loss of Offsite Power)). 'A' train emergency power remained energized from offsite sources. The plant is stable at normal operating pressure and temperature. At 0433 (CDT), 2B Reactor Coolant Pump was re-started when support conditions were re-established. Heat sink is adequate using the 2A Motor Driven Auxiliary Feedwater Pump. Unit 2 'B' train power was restored by starting the 2C emergency diesel generator at 0523 (CDT). This restored power to the Digital Rod Position Indication system, and control rod K-8 in control bank 'C' indicated full out, and all other control rods fully inserted. An emergency boration is in progress to compensate for the stuck rod. Additionally, the reactor trip resulted in a valid actuation of the Aux Feedwater system which is an eight hour non-emergency report per 10 CFR 50.72(b)(3)(iv)(A). During the transient, one primary PORV momentarily opened, then reseated. Decay heat is being directed to the atmospheric relief valves with no indicated primary to secondary leakage. There was no impact on Unit 1. The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 2125 EDT ON 10/15/2014 FROM BLAKE MITCHELL TO MARK ABRAMOVITZ * * *

Digital rod position indication troubleshooting was conducted on 10/14/2014 and confirmed all control rods, including control rod K-8, fully inserted following the reactor trip. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ayres), NRR EO (Davis) and IRD (Gott).

ENS 5052612 October 2014 12:41:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopVEGP (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.
ENS 5031427 July 2014 15:22:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopVEGP (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.
ENS 5003112 April 2014 23:20:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 500068 April 2014 05:21:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopVEGP 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.
ENS 4946122 October 2013 13:18:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 4945319 October 2013 09:36:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopVEGP (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.
ENS 4910612 June 2013 01:33:00FarleyAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopThis is a report of an automatic RPS actuation and automatic ESF actuation per 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). Additionally, this is to report intentions for a press release per 10CFR50.72(b)(2)(xi). At 2105 CDT on 6/11/13, Farley Unit 1 experienced an automatic reactor trip from 100% power. The initiating event was the loss of the 1B Start up Transformer which resulted in de-energization of the B-Train ESF 4KV buses and the 1B and 1C Reactor Coolant Pump Buses. The 1B Emergency Diesel Generator auto started and tied to the B-Train 4KV Emergency buses. Both MDAFW (Motor Driven Auxiliary Feedwater) Pumps and the TDAFW (Turbine Driven Auxiliary Feedwater) Pump auto-started and are supplying AFW flow to the steam generators. Decay heat removal is via the steam dumps to the main condenser. The cause of the loss of the 1B Start-up Transformer is unknown and is currently under investigation. All other systems functioned as expected in response to the loss of the 1B Start-up Transformer and reactor trip. The NRC Senior Resident Inspector has been notified. A press release is planned. All control rods fully inserted. There is no impact on Unit 2. Currently the licensee does not plan to restart the 1B and 1C Reactor Coolant Pumps. Pressurizer spray has been isolated from the 1B loop per procedure. Main Condenser vacuum is adequate for decay heat removal.
ENS 4878827 February 2013 02:07:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 4873810 February 2013 09:11:00HatchManual ScramNRC Region 2GE-4During normal power operations, the crew observed condensate/feedwater conductivity begin to increase at approximately 0530 EST on 02/10/13. The crew responded to the associated alarm response procedures and entered abnormal operating procedure 34AB-N61-001-1 due to degrading reactor water chemistry parameters. A power reduction (from 100%) was initiated at 0555 EST in accordance with station procedures for responding to a suspected condenser tube leak. At 0700 EST, a manual reactor SCRAM (from approximately 47%) was inserted due to the elevated reactor water conductivity in accordance with station abnormal operating procedures. All rods inserted completely and no complications were encountered following the reactor SCRAM, normal feedwater injection remained available. Following the SCRAM, a Group 2 Primary Containment Isolation Signal (PCIS) was received as a result of reactor water level lowering below +3 inches. The lowest reactor water level observed was (minus) 2 inches and was restored to normal operating levels utilizing normal feedwater injection. Following restoration of reactor water level to the normal operating level, the Group 2 PCIS signal was reset. No ECCS injection systems actuated as a result of the reactor SCRAM. The SCRAM was uncomplicated and the plant is stable. Decay heat removal is to the main condenser via the turbine bypass valves. The plant is in a normal offsite electrical power shutdown alignment. Efforts are in progress to isolate the condenser in-leakage. There was no impact on Unit 2. The licensee has notified the NRC Resident Inspector.
ENS 4783614 April 2012 16:12:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 4736924 October 2011 17:01:00HatchManual ScramNRC Region 2GE-4While performing a startup of HNP-2, after reaching criticality, the crew observed erratic indications on two Intermediate Range Monitors (IRMs), 2C51K601A and 2C51K601C. IRM 2C51K601A had been spiking and was subsequently bypassed. The 2C51K601C was spiking downscale and could not be bypassed due to the 2C51K601A being bypassed already. Both IRMs are in the 'A' RPS trip system. At the time when the second IRM was acting erratic, the crew identified the condition as both IRMs in the same quadrant and did not continue withdrawal of control rods. As a result of not withdrawing control rods, reactor power began to decrease and the crew conservatively inserted a manual reactor scram to shutdown the reactor. All rods did fully insert into the core. No PCIS (Primary Containment Isolation System) actuations occurred and none were expected to occur based on plant conditions following the scram. At this time, investigation is in progress, but the investigation and corrective actions have not yet been completed. The crew is maintaining HNP-2 in Hot Standby (Mode 3) at this time. The licensee informed the NRC Resident Inspector.
ENS 4722431 August 2011 12:02:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopWhile 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.
ENS 4594622 May 2010 19:10:00FarleyManual Scram
Automatic Scram
NRC Region 2Westinghouse PWR 3-LoopUnit 2 was operating at 100% power in the normal operating procedure, FNP-2-UOP-3.1, Power Operation, when multiple alarms were received associated with 2C Steam Generator (S/G) level, and a process cabinet failure. The control room team noticed there was no power or control capability on the 2C S/G Feedwater Regulating Control Valve (FRV), and 2C S/G level was decreasing. The control room team attempted to take manual control of the 2C FRV, which did not respond. The reactor was manually tripped when 2C S/G narrow range level reached 40%. The automatic trip set point for S/G level is 28%. All systems responded properly for the reactor trip and there were no complications. The investigation indicates there was an Nuclear Controller Driver (NCD) card failure in Process Control Cabinet 8. The controller card controls the 2C S/G FRV controller, which prevented any automatic, or manual control of the 2C S/G FRV, or 2C S/G level. There were no safety or relief valves that lifted and decay heat is being removed via steam dump control valves. Auxiliary feedwater pumps are maintaining level in the steam generators. Electrical lineup is normal. The licensee has notified the NRC Resident Inspector.
ENS 4558823 December 2009 17:18:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 4555710 December 2009 01:38:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 455477 December 2009 20:07:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopVogtle 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 4514823 June 2009 05:08:00HatchAutomatic ScramNRC Region 2GE-4(The reactor automatically scrammed) on a Main Turbine Trip >27.6% rated thermal power. The main turbine trip was due to reactor high level. Post scram, reactor level decreased to approximately -26 inches. Reactor water level was restored with the condensate system. Both reactor recirc pumps tripped as required on EOC RPT Logic when the main turbine tripped. Both pumps have been restarted. A Group 2 isolation was received at +3 inches reactor water level with all valves closing as required. Investigation as to the cause of the transient is underway. All rods inserted during the scram. No relief valves actuated during the transient. Decay heat is being removed via turbine bypass valves to the main condenser. The plant is within normal shutdown temperature and pressure limits. The electrical grid is stable and the plant is in a normal shutdown electrical lineup. The Group 2 has been reset. There was no effect on Unit 1. The licensee has notified the NRC Resident Inspector.
ENS 4514520 June 2009 15:47:00HatchAutomatic ScramNRC Region 2GE-4Plant Hatch Unit 2 experienced a full reactor scram from the main generator protection circuitry (generator runback circuit). Preliminary indications are that a main generator high temperature signal was received, initiating the generator protection (runback) circuitry and a high reactor pressure scram signal was received during the turbine/generator runback. Investigations into the cause of the generator high temperature signal are ongoing. Reactor water level was recovered using the reactor feed system, and reactor pressure was controlled using main turbine bypass valves. All control rods inserted, as expected, during the scram. Other than the cause of the main generator high temperature signal, all systems functioned as expected. Unit is currently at 837 psig; 540 degrees F in Mode 3. Electrical system is in a normal lineup. The licensee informed the NRC Resident Inspector.
ENS 4467922 November 2008 12:26:00HatchManual ScramNRC Region 2GE-4Manual Rx Scram initiated due to a loss of condensate/feedwater. Condensate Booster Pump 1A tripped due to low suction pressure and then both Reactor Feed-pumps tripped. Both HPCI and RCIC initiated on low level and restored reactor water level to normal band 5 to 50 inches. Both Reactor Water Recirculation Pumps tripped due to low level at RWL (Reactor Water Level) - 60". Lowest RWL was approximately minus 70 inches and a group two isolation occurred at RWL 3 inches. All group two 2 valves closed as required. The cause of the low condensate booster suction pressure is under investigation. Rods fully inserted on the scram. No safety or relief valves lifted after the scram. Reactor water level is being maintained with normal feed and decay heat is being removed to the main condenser. The plant is in its normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector.
ENS 443374 July 2008 10:04:00HatchAutomatic ScramNRC Region 2GE-4A Turbine Trip greater than 30% power caused a Reactor Trip (Scram), both Recirculation Pumps tripped. A low level (Reactor Vessel Water Level) of approximately 2 inches caused a Group 2 containment valve isolation signal, all valves closed as required. The cause of the Turbine Trip is under investigation All control rods fully inserted with no ECCS actuations. Unit 1 is currently stable in mode 3 (Hot Shutdown) with decay heat being removed via the bypass. Following the scram, one SRV lifted and reseated. At the time of the transient, an EHC pump autostart was in progress, however, there is no indication that this was the cause of the turbine trip. Unit 1 is in a normal shutdown electrical lineup. The licensee informed the NRC Resident Inspector.
ENS 440467 March 2008 17:21:00HatchAutomatic ScramNRC Region 2GE-4Unit 2 RPS actuation / unplanned scram with subsequent ECCS discharge to the RCS at 1446 hrs. on 3/07/08. Unit 2 scrammed on Low RPV water level of 3 inches above instrument zero as a result of a loss of condensate feedwater. Water level decreased to approximately 60 inches below instrument zero as a result of the loss of feedwater. (Top of active fuel is approximately 150 inches below instrument zero.) The cause of the loss of feedwater is presently under investigation. At 35 inches below instrument zero, HPCI and RCIC actuated and restored water level. HPCI oscillations were experienced and the system was taken to manual control, at which time the flow oscillations abated. All other systems functioned as required. A team has been assembled to investigate and determine the cause of the initiating event of the loss of feedwater. During the scram, all rods inserted into the core. There were no safety relief valve actuations as a result of the transient. RPV level was restored and is being maintained using control rod drive flow. The electrical grid is stable with normal offsite power supplying safety loads. Decay heat is being removed using the turbine bypass valves to condenser. The licensee has notified the NRC Resident Inspector.
ENS 4331323 April 2007 11:25:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopVogtle 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.
ENS 4280627 August 2006 09:20:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopOn 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.
ENS 4250617 April 2006 01:20:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopUnit 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.
ENS 4205817 October 2005 19:18:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 4172523 May 2005 20:07:00HatchManual ScramNRC Region 2GE-4

Based on increasing conductivity in the reactor vessel and condenser hotwell, a power reduction was initiated from 100 percent power. A manual scram was inserted at 57 percent RTP and 49 percent Core Flow based on Chemistry recommendations due to sulfates and chlorides in the hotwell. Following the scram a reduction in reactor water level to -28 inches resulted in a Primary Containment Group 2 Isolation (ESF) occurring. All isolations and systems responded as expected. Current plant status is Hot Shutdown with plans to proceed to cold shutdown. All control rods fully inserted and decay heat is being removed with the bypass valves into the condenser. The licensee notified the NRC Resident Inspector.

  • * * RETRACTION FROM SHIFT SUPERVISOR (TONY SPRING) TO ABRAMOVITZ AT 16:33 ON 6/6/2005 * * *

After further review and evaluation it has been determined that the four hour call made May 23, 2005 per the guidance of 50.72(b)(2)(iv)(B) should be retracted. A review of the event with respect to NUREG 1022 Revision 2 determined that: The manual scram was part of a pre-planned sequence to shut the plant down due to an equipment problem. The manual scram was part of a pre-planned sequence. The guidance to scram the reactor was established by the plant's Abnormal Operating Procedure addressing a condenser tube leak and was part of a preplanned sequence to prevent future equipment and component failures. The Manual Scram was not inserted to protect the plant against an event that presented a challenge to an FSAR analyzed event. In other words, this was not an Anticipated Operational Occurrence, an Accident, or a Special Event as defined in section 15.1.3 of the Unit 2 FSAR. Rather it was part of a plan to shutdown the reactor to protect against future potential equipment problems due to out of limits chemistry parameters. Further justification is provided by the fact that the manual scram was not initiated in anticipation of an automatic scram. Per NUREG 1022 Rev. 2: 'The staff also considers intentional manual actions, in which one or more system components are actuated in response to actual plant conditions resulting from equipment failure or human error, to be reportable because such actions would usually mitigate the consequences of a significant event. This position is consistent with the statement that the commission is interested in events where a system was needed to mitigate the consequences of the event.' However, the reporting requirement itself indicates that actuations that result from pre-planned sequences are not reportable. An example is provided in the NUREG of an equipment problem involving the loss of recirc pumps. In this example it is stated that: 'Even though the reactor scram was in response to an existing written procedure, this event does not involve a preplanned sequence because the loss of the recirc pumps and the resultant off-normal procedure entry were event driven, not pre-planned.' This is similar to our event, however, in the NUREG example, the reactor is scrammed to protect against the possibility of a stability event and stability is an FSAR analyzed event. In our case we were shutting down for chemistry reasons, not an FSAR type event. It is concluded that when the RPS is used to shutdown the reactor as part of a plan for the resolution of equipment problems, and the RPS is not needed to mitigate the consequences of an FSAR analyzed event, i.e., one which threatens a fission product boundary (i.e., fuel cladding, RCPB, primary and secondary containments), the RPS actuation is not reportable under 50.72(b)(2)(iv)(B).

The licensee notified the NRC Resident Inspector. Notified R2DO (Haag).

  • * * RETRACTION RESCINDED - S. BURTON TO M. RIPLEY 1524 EDT 06/08/05 * * *

On May 23, 2005 a four hour report was made per the guidance of 50.72(b)(2)(iv)(B), 'Any event or condition that results in an actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This was made per event # 41725. The report was made within the four hour time frame of 10 CFR 50.72(b)(2). The four hour report for event # 41725 was retracted on June 6, 2005. After further consideration, the retraction made on June 6, 2005 is being cancelled and the original report re-instated. The licensee notified the NRC Resident Inspector. Notified R2 DO (K. Landis)

ENS 4165330 April 2005 00:00:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 4132311 January 2005 09:20:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 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.
ENS 4121320 November 2004 14:56:00VogtleAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopOn 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.
ENS 4107025 September 2004 02:44:00HatchManual ScramNRC Region 2GE-4Received Group 2 isolation signal on Low Reactor Water Level during initiation of manual reactor scram during planned shutdown. Reactor Water Level decreased to -0.5 inches. Group 2 isolation setpoint is +3.0 inches. All Group 2 isolation valves closed as required. During the planned shutdown, all control rods fully inserted. Decay heat is being removed to the main condenser via the main turbine bypass valves, ESF and ECCS systems remain available, and the electrical grid is stable. The licensee will notify the NRC Resident Inspector.
ENS 4066611 April 2004 12:47:00FarleyAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1247 EDT on 04/11/04, the licensee reported that at 1105 CDT on 04/11/04, control room operators were performing low power physics testing in accordance with FNP-2-STP-101 during startup of Unit 2 following a refueling outage. With reactor power at 10E-8 amps in the intermediate range in Mode 2, the 'B' reactor trip breaker opened for unknown reasons. All control rods inserted completely. The licensee is investigating the cause. The licensee notified the NRC Resident Inspector.
ENS 4061528 March 2004 00:30:00VogtleManual ScramNRC Region 2Westinghouse PWR 4-LoopOn 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.
ENS 4030910 November 2003 12:41:00FarleyAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopReactor Protection System actuation in response to an indicated (not an actual) 2A RCP (Reactor Coolant Pump) breaker open position signal. All reactor protection and support systems operated as expected. The Aux Feedwater System started as required in response to the tripping of both Steam Generator Feed Pumps. All 3 RCPs are running; none have tripped. Not understood is the indication of the 2A RCP breaker open when the breaker has remained closed. The licensee reported that all control rods fully inserted; decay heat is being rejected to the condenser via the steam dumps; a steam generator atmospheric relief may have momentarily lifted during the transient; and that the electrical grid is stable. The licensee will be notifying the NRC Resident Inspector.