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The query [[Category:ENS Notification]] [[Reactor type::Westinghouse PWR 3-Loop]] [[Scram::+]] was answered by the SMWSQLStore3 in 0.0565 seconds.


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 Entered dateSiteScramRegionReactor typeEvent description
ENS 530607 November 2017 22:32:00SummerAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

On 11/7/2017 at 1957 (EST), VC Summer Nuclear Station automatically tripped due to a turbine trip. The cause of the turbine trip is under investigation at this time. All systems responded as expected. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The unit is currently stable in Mode 3 with decay heat removal via the Main Steam to the Main Condenser. The NRC Resident Inspector has been notified. The licensee will notify the South Carolina State Emergency Management Division, the Fairfield, Richland, Lexington and Newberry Counties.

  • * * UPDATE FROM BETH DALICK TO VINCE KLCO ON 11/8/17 AT 1409 EST * * *

All systems responded as expected, with the exception of 'B' Steam Generator Feedwater Isolation Valve XVG1611 B-FW. This valve did not appear to automatically close and was slow to indicate closed from the Main Control Board. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. Notified the R2DO (Musser).

ENS 530567 November 2017 08:29:00Beaver ValleyAutomatic ScramNRC Region 1Westinghouse PWR 3-LoopOn November 7, at 0504 (EST), BVPS (Beaver Valley Power Station) Unit 1 experienced an automatic reactor trip due to Main Unit Generator over current. The Auxiliary Feedwater system activated and remains in service. Offsite power supply is available. Normal and Emergency busses are being supplied by Offsite power. One Source Range channel failed to energize due to its corresponding Intermediate Range instrument being under compensated. It was manually energized and is not indicating as expected. The second Source Range instrument energized but is reading erratically. Both Source Range instruments have been declared inoperable and the appropriate Technical Specification has been complied with by making the Control Rods not capable of withdrawal and isolating all dilution flow paths. Plant trip response was as expected without complications, and all control rods fully inserted in the core. The plant is currently stable in Mode 3. This event is being reported as an actuation of the Reactor Protection system 10 CFR 50.72(b)(2)(iv)(B) and a Specified System Actuation (Auxiliary Feedwater System) 10 CFR 50.72(b)(3)(iv)(A). BVPS Unit 2 is unaffected by this event and remains at 100% power in Mode 1. The NRC Resident Inspector has been notified.
ENS 5296010 September 2017 22:20:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-LoopOn 09/10/17 at 1855 (EDT), (Turkey Point) Unit 4 reactor was manually tripped from 88% RTP (Rated Thermal Power) due to a failure of 4C Steam Generator main feed regulating valve causing lowering S/G (Steam Generator) level. All other systems operated normally. Auxiliary Feed Water initiated as designed to provide S/G water level control. EOP's (Emergency Operating Procedures) have been exited and General Operating procedures (GOP'S) were entered. Unit 4 is stable in Mode 3 at NOT/NOP (Normal Operating Temperature/Normal Operating Pressure). The licensee is investigating the failure of the feed regulating valve. Offsite power is available. Decay heat is being removed via main feedwater with steam discharged to atmosphere using the ADVs (Atmospheric Dump Valves). There is no known primary-secondary steam generator tube leakage. The licensee informed the NRC Resident Inspector.
ENS 5293228 August 2017 10:55:00SummerAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopOn 8/28/2017 at 0837 (EDT), VC Summer Nuclear Station automatically tripped due to a turbine trip. The turbine trip was caused by the Main Generator Differential Lockout due to a fault on the center phase lightning arrester on the Main Transformer (XTF-001). There were no complications with the trip. All control rods fully inserted. Balance of Plant (BOP) buses automatically transferred to their alternate power source XTF 31/32. All Emergency Feedwater pumps started as required. All systems responded as required. The plant is stable in Mode 3. Station personnel are investigating the cause of the fault on the main transformer lightning arrester. This event is reportable per 10 CFR 50. 72(b)(2)(iv)(B) and 10 CFR 50. 72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. The unit is currently stable in Mode 3 with decay heat removal via the Main Steam to the Main Condenser. The licensee will inform both State and local authorities.
ENS 5283329 June 2017 12:43:00SummerAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

On 6/29/2017 at 0857 (EDT), VC Summer Nuclear Station automatically tripped due to a loss of normal feed water flow to the B Steam Generator.

There were no complications with the trip. All control rods fully inserted. All emergency feedwater pumps automatically started and recovered steam generator levels. The plant is stable in Mode 3. Station personnel are investigating the cause of the loss of normal feedwater to the B Steam Generator. This is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. The licensee notified the State of South Carolina as well as Fairfield, Lexington, Richland and Newberry Counties regarding the event.

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 522929 October 2016 06:09:00SurryAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopSurry Unit 2 reactor automatically tripped at 0254 hours on 10/09/2016, due to a Main Generator Differential Lockout Turbine Trip. The cause of the generator differential lockout is under investigation at this time. Reactor Coolant System temperature is currently being maintained at 547 degrees Fahrenheit on the main steam dump valves. All three Auxiliary Feedwater Pumps automatically started as designed on Low-Low Steam Generator Water Level following the trip. Auxiliary feedwater pumps have since been secured and Main Feedwater is in use. All systems operated as required. The source range nuclear instruments had to be manually reinstated following the reactor trip due to indications of undercompensation on Intermediate Range Nuclear Instrument channel N-36. Off site power remains available. There is no impact on Surry Unit 1. This notification is being made pursuant to 10CFR50.72(b)(2)(iv)(B) for 4-hour notification of Reactor Protection System activation and 10CFR50.72(b)(3)(iv)(A) for 8-hour notification of automatic actuation of the Auxiliary Feedwater System. The NRC Resident Inspector has been notified and is responding to the site. There were no radiation releases, personnel injuries, or contamination events due to this event. All control rods fully inserted. Secondary reliefs lifted and reseated as expected following a reactor trip from 100% power.
ENS 522908 October 2016 13:44:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

UE SU1.1 declared due to momentary loss of power from the qualified off-site source. Both Emergency Diesel Generators started and loaded to supply power to both of the Emergency Buses. 'A' Service Water Pump did not start on Blackout sequencer. Sufficient Service Water flow is available from the other three operating pumps. All other systems operated as designed." At 1304 EDT Robinson Unit 2 experienced a momentary grid voltage drop that lowered the 4kV bus voltage and initiated an automatic reactor trip. All rods inserted and decay heat is being removed by steam generator PORVs. In response to the reduced bus voltage, the Emergency Diesel Generators (EDGs) automatically started and loaded onto the emergency busses. At 1317 EDT, the licensee declared an Unusual Event (EAL SU1.1) due to the loss of offsite power. The licensee is currently investigating the cause of the grid voltage instability. The emergency busses will continue to be powered by the EDGs until the licensee has determined the cause for the voltage drop. All offsite power sources and all equipment is available. The licensee has notified the state government and Darlington County. The NRC Resident Inspector has been notified. Notified DHS SWO, FEMA OPS Center, FEMA National Watch (email only), DHS NICC, Nuclear SSA (email only).

  • * * UPDATE FROM ALEX CURLINGTON TO DANIEL MILLS AT 1658 EDT on 10/08/16 * * *

At 1303 EDT on 10/08/2016, a reactor trip occurred. The cause was under voltage to the plant 4kV buses due to an offsite grid disturbance. The cause of the disturbance is under investigation. Following the reactor trip, the Auxiliary Feedwater System actuated as expected on low steam generator level. At the time of the trip, the plant was in Mode 1. Currently, the Plant is in Mode 3. The current RCS Temperature is 550 degrees F (Average), and the Steam Generator Levels are in the range of 42 to 53% (normal range) with levels controlled by the Auxiliary Feedwater System. Decay heat removal is being controlled by the steam generator PORVs. 'A' Service Water Pump did not start on Blackout sequencer. Sufficient Service Water flow is available from the other three operating service water pumps 'B', 'C', and 'D'. All other systems operated as designed. Due to the Automatic Actuation of the Reactor Protection System, this event is being reported as a 4-hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B). Due to the valid actuation of Auxiliary Feedwater System, this event is also being reported as an 8-hour Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A)(B)(6). At no time during this occurrence was the public or plant staff at risk as a result of this event. The Resident Inspector has been notified.

  • * * UPDATE FROM BOBBY STUCKEY TO DANIEL MILLS AT 2347 EDT on 10/08/16 * * *

At 2323 (EDT) Emergency Bus E-2 powered from off-site power." The NRC Resident Inspector will be notified. Notified R2DO (Bonser), IRD (Grant), NRR EO (Miller).

  • * * UPDATE FROM BOBBY STUCKEY TO JOHN SHOEMAKER AT 0028 EDT ON 10/09/16 * * *

At 0011 (EDT) Robinson Nuclear Plant has terminated the Unusual Event. Basis for the Unusual Event termination was restoration of power to Emergency Bus E-2 from off-site power. The licensee has notified the NRC Resident Inspector. Notified R2DO (Bonser), NRR EO (Miller), IRD (Grant), DHS SWO, FEMA OPS Center, FEMA National Watch (email only), DHS NICC, Nuclear SSA (email only).

  • * * UPDATE FROM GEORGE CURTIS TO JOHN SHOEMAKER AT 0253 EDT ON 10/09/16 * * *

At approximately 2323 EDT on 10/08/2016, an auto-start of the Auxiliary Feedwater (AFW) Motor-Driven pumps occurred during the transfer of Emergency Bus power from the 'B' Emergency Diesel Generator (EDG) to offsite power. AFW system auto-start logic associated with Main Feed Pump (MFP) breakers being open is defeated when the EDG output breaker is closed. As such, when the EDG output breaker was opened during the power transfer while the MFP breakers were open, the auto-start logic was thereby met causing the AFW auto-start.

Due to the valid actuation of the AFW System, this event is being reported as an 8-hour Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A). At no time did this occurrence pose undue risk to the health and safety of the public. The NRC Senior Resident Inspector has been notified of this event. H.B. Robinson Unit 2 was in Mode 3 during this event. Notified R2DO (Bonser).

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 5213630 July 2016 03:52:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-Loop

Power had been reduced for planned maintenance on the 3B feedwater heater. During isolation of the feedwater heaters, a repeated water hammer was experienced. A normal shutdown of Unit 3 was performed. Unit 4 was not affected by the water hammer. The decay heat is being removed via the condenser and all offsite and onsite electrical power is available. The investigation of the cause is underway. The licensee will notify the NRC Resident Inspector.

  • * * UPDATE AT 2141 EDT ON 07/30/16 FROM ERIC JUERGENS TO S. SANDIN * * *

The licensee is retracting this report based on the following: At 0352 EDT on July 30, 2016, EN #52136 provided notification of a Reactor Protection System (RPS) actuation during a normal shutdown of Unit 3 in response to a secondary system equipment issue. Upon further investigation, the unit shutdown and manual RPS actuation were in accordance with general plant operating procedures. The manual RPS actuation was in accordance with the general operating procedure and not required to mitigate the consequences of the secondary system equipment issue. As such, the notification made by EN #52136 for a valid actuation of a specified system is hereby retracted. The NRC Resident lnspector will be notified. Notified R2DO (Rose).

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 5106512 May 2015 05:48:00Turkey PointAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopThis is an non-emergency notification to the NRC (Headquarters) Operations Center in accordance with 10 CFR 50.72(b)(2)(iv)(B) for a valid actuation of the reactor protection system (RPS) (four hour notification) and 10 CFR 50.72(b)(3)(iv)(A) for a valid Engineered Safeguards (ESF) actuation (eight hour notification) due to auxiliary feedwater (AFW) initiation. On 5/12/15 at 0430 EDT, Unit 4 experienced an automatic reactor trip due to Generator Differential Trip. Investigation is underway to determine the cause. Auxiliary feedwater automatically initiated as expected. All systems operated correctly in response to the reactor trip. Unit 4 is currently in Mode 3 and stable. All control rods fully inserted. Normal offsite power is available with decay heat is being removed by the atmospheric steam dumps. There is no known primary to secondary reactor coolant system leakage. The licensee has notified the NRC Resident Inspector.
ENS 5098515 April 2015 07:32:00Beaver ValleyManual ScramNRC Region 1Westinghouse PWR 3-LoopAt 0411 EDT on April 15, 2015, Beaver Valley Power Station (BVPS) Unit 1 manually tripped the reactor from approximately 85% power due to the trip of a condensate pump. The unit was performing an emergent power reduction due to a degraded condensate pump prior to the manual reactor trip. An end of cycle Tave coastdown was in progress at the time of the event. All control rods fully inserted into the core. All three auxiliary feed water pumps started as expected and were subsequently secured in accordance (with) station procedures. The main feedwater system remains available and in service. The unit is currently stable in Mode 3. Unit 2 was unaffected and remains at full power. Decay heat removal is via main feedwater system with steam discharge to the main condenser via the steam bypass valves. Unit 1 is in a normal shutdown electrical lineup. No primary or secondary reliefs or safeties lifted during the transient. The licensee informed the NRC Resident Inspector.
ENS 509462 April 2015 06:55:00North AnnaManual ScramNRC Region 2Westinghouse PWR 3-LoopOn April 2, 2015 at 0426 EDT, the Unit 1 reactor was manually tripped while operating at 100 percent power due to a failure of the main generator voltage regulator. This also resulted in a turbine trip. The operations crew entered the reactor trip procedure and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the reactor trip. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated as designed and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 1 is in a normal shutdown electrical lineup. The NRC Resident Inspectors have been notified. The Louisa County Administrator will be notified. There was no effect on Unit 2 as a result of this trip.
ENS 5085126 February 2015 16:39:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopOn February 26, 2015, at 1511 EST, with Unit 1 operating at 95% power in an end of cycle coastdown, the 'B' Main Feedwater Reg Valve failed closed which resulted in a Unit 1 automatic reactor trip due to 'B' Steam Generator low/low level. The operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the reactor trip. This reactor protection system actuation is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater system is reportable per 10CFR50.72(b)(3)(iv)(A) for the valid actuation of an ESF system. The Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 1 is in a normal shutdown electrical lineup. The NRC Resident Inspectors have been notified and are in the Control Room. The Louisa County Administrator will be notified.
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 5052913 October 2014 11:13:00SurryAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopUnit 2 reactor automatically tripped at 0758 (EDT) hours on 10/13/2014, due to a spurious overpower/delta temperature signal on all three channels. The cause of the spurious signal is unknown at this time. Currently, reactor coolant system temperature is being maintained stable at 546 (F) degrees. All three auxiliary feedwater pumps automatically initiated as designed on low-low steam generator level following the trip. All systems responded as expected with the exception (both) of the intermediate range neutron indication(s), which was determined to be under-compensated. The source range indication did not automatically energize and was energized manually. All other systems operated as required. This notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for 4-hour notification of reactor protection system activation and 10 CFR 50. 72(b )(3)(iv)(A) for 8-hour notification of automatic actuation of auxiliary feedwater. The NRC resident has been notified of this event and is on site. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events. There was no testing in progress when the reactor trip occurred. The reactor trip was considered uncomplicated. All control rods fully inserted. Decay heat is being released via main feedwater and the condenser steam dumps. Normal offsite power is available. There was no effect on Surry Unit 1 which continues to operate at 100% power. The licensee is investigating the cause of the overpower/delta temperature actuation.
ENS 5035411 August 2014 14:20:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-LoopThis is an non-emergency notification to the NRCOC (Nuclear Regulatory Commission Operations Center) in accordance with 10 CFR 50.72(b)(2)(iv)(B) for a valid actuation of the reactor protection system (RPS) (four hour notification) and 10 CFR 50.72(b)(3)(iv)(A) for a valid Engineered Safeguards (ESF) actuation (eight hour notification) due to auxiliary feedwater (AFW) initiation , safety injection (SI) initiation, and emergency diesel generator (EDG) auto start. On 8/11/14 at 1028 EDT, Unit 3 experienced a manual reactor trip due to a loss of instrument air. Auxiliary feed water automatically initiated as expected. Unit 3 is currently in Mode 3 and stable. Instrument air has been restored. At 1305 a safety injection signal (SI) occurred due to main steamline high differential pressure during a plant cooldown to 470 degrees Fahrenheit. High head safety injection (HHSI) pump, residual heat removal (RHR) pumps, and emergency diesel generators (EDGs) auto started as expected due to the safety injection signal. Based on plant conditions, the HHSI and RHR pumps did not inject into the reactor coolant system. The plant was stabilized after the reactor trip, however, the loss of instrument air resulted in a loss of letdown and PZR level continued to slowly increase to approximately 91%. All rods fully inserted and normal offsite power was maintained. Operators initiated a plant cooldown to stabilize and lower PZR level using the S/G atmospheric steam dumps. Unequal opening of the S/G atmospheric steam dumps caused the SI to occur due to main steamline high differential pressure. No RCS injection occurred and all SI equipment operated as expected and has been subsequently secured. Unit 3 is stable in mode 3 at 490 degrees Fahrenheit and RCS pressure is 1940 psig. There was no effect on Unit 4 which continues to operate at 100% power. The loss of instrument air is being investigated. The licensee has notified the NRC Resident Inspector.
ENS 5029322 July 2014 07:53:00SummerAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 0414 (EDT on 7/22/2014), VC Summer Nuclear Station automatically tripped due to decreasing water level in the 'C' Steam Generator. The trip occurred when valve XVB-09210 WI System Condensate Bypass Valve failed to open as required while the station was removing the Condensate Polishing System during startup. The Condensate Polishing System is used to purify and filter the condensate from the non-nuclear, secondary side of the plant. This valve failure caused low level in the Deaerator Storage Tank, which consequently tripped all feedwater pumps. This loss of feedwater led to Lo-Lo Steam Generator level in the 'C' Steam Generator. All Emergency Feedwater pumps automatically started on Lo-Lo Steam Generator level and all control rods inserted fully. The Steam Generator levels recovered quickly. Presently the plant is in Mode 3. Decay heat is being removed by dumping steam to the condenser. A station response team is actively investigating the cause of the event. The NRC Resident Inspector has been notified." The licensee will notify the State and local governments.
ENS 5014025 May 2014 07:37:00Turkey PointAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopThis is an non-emergency event notification to the NRC Operations Center in accordance with 10 CFR 50.72(b)(2)(iv)(B) for a valid actuation of the reactor protection system (RPS) (four hour notification). On 05/25/2014 at 0536 EDT, as part of a planned maintenance outage to repair a leaking check valve inside the containment building, Unit 4 power level had been lowered to 19% reactor power for the planned opening of the reactor trip breakers. Unit 4 experienced a loss of condenser vacuum. An automatic turbine trip signal was generated which initiated a reactor trip signal. The reactor tripped as required on the automatic trip signal. Unit 4 is currently in Mode 3 and stable. The loss of vacuum is under investigation at this time. All other plant systems are working as designed. The reactor trip was uncomplicated with no other safety system actuations. Decay heat is being removed via steam dumps to atmosphere. There is no known primary to secondary leakage. The grid is stable and Unit 4 is in its normal shutdown electrical lineup. Unit 3 was not affected and continues to operate at 100% power. The licensee has notified the NRC Resident Inspector.
ENS 5012420 May 2014 10:52:00Beaver ValleyManual ScramNRC Region 1Westinghouse PWR 3-LoopOn May 20, 2014, at 0835 hours during plant startup, Beaver Valley Power Station Unit 2 Operations personnel manually tripped the reactor due to meeting the pre-briefed trip criteria of 85% narrow range level on the 'A' Steam Generator. This manual trip criterion was reached after the steam generator water level began to oscillate following the start of the 'A' Condensate pump. A manual main steam line isolation was performed in order to limit reactor coolant system cool down. Plant trip response was as expected without complications, and all control rods fully inserted in the core. The plant is currently stable in Mode 3. This event is reportable pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Beaver Valley Power Station Unit 1 was not affected by this event. The NRC Resident Inspector has been notified. No relief or safety valves lifted during this event. The unit is maintaining primary temperature using the atmospheric steam dumps and main feedwater pumps. There is no primary to secondary leakage. The plant is in its normal shutdown electrical lineup.
ENS 497842 February 2014 11:01:00North AnnaManual ScramNRC Region 2Westinghouse PWR 3-LoopOn 2-2-2014 at 0859 (EST), with Unit 2 operating at 100% power, a manual reactor trip was initiated by the control room staff following a trip of the 'A' main feedwater pump and automatic start of the 'C' feedwater pump due to crew concerns that both motors of the 'C' feedwater pump had not actuated. When the 'C' feedwater pump auto started, the running indicator light for one of the 'C' feedwater pump motors failed to illuminate. Both motors of the 'C' feedwater pump had started as designed. Following the reactor trip, all control rods fully inserted into the core and Unit 2 was stabilized in Mode 3 at normal reactor coolant system temperature and pressure. Decay heat is being removed using the normal condenser steam dump system. Unit 2 is in a normal shutdown electrical alignment with power being supplied from the Reserve Station Service Transformers. This event is reportable per 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system. Following the reactor trip, the auxiliary feedwater pumps automatically started as designed and provided makeup flow to the steam generators. The steam generator levels were returned to normal operating level and the auxiliary feedwater pumps were returned to the normal standby automatic alignment. This event is reportable per 10CFR50.72(b)(3)(iv)(A) for actuation of an ESF system. Unit 1 is operating at 100% power and was not affected by the event. The licensee informed the NRC Resident Inspector and will inform the Louisa County Administrator.
ENS 4974218 January 2014 10:47:00HarrisManual ScramNRC Region 2Westinghouse PWR 3-Loop

(At 1016 EST, an) Alert (was declared) based on EAL # HA 2.1 Fire or explosion resulting in either: visible damage to any table H-1 structure or system/component required for safe shutdown of the plant, or control room indication of degraded performance of any safe shutdown structure, system, or component within any table H-1 area. Fire in 480V bus 1D2. Reactor was manually tripped 480 VAC safety related transformer fire in switchgear room. Plant reduced power and tripped the reactor manually. Reactor trip was uncomplicated. Fire was extinguished when the 480 VAC bus was de-energized. The licensee has notified the NRC Resident Inspector, the State of North Carolina, and other local authorities. Notified DHS SWO, DOE Ops Center, FEMA Operations Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and NuclearSSA via e-mail.

  • * * UPDATE FROM JOEL DUHON TO JOHN SHOEMAKER AT 1602 EST ON 1/18/14 * * *

Harris Nuclear Plant secured from the Alert at 1551 EST, on 1/18/14. The plant is stable, the fire is out, the TSC and EOF have been secured and plant recovery has been transferred to the outage control center. There were no personnel injuries or radiological releases. Radiation monitor RM-*1TS-3653C (Technical Support Center Radiation Monitor) is out of service. The licensee has notified the NRC Resident Inspector. Notified the R2DO (King), R2RA (McCree), NRR (Leeds), IRD MOC (Grant), OPA (Brenner), NRR EO (Lee) Notified DHS SWO, DOE Ops Center, FEMA Operations Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and NuclearSSA via e-mail.

ENS 4970810 January 2014 00:27:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 2234 hours EST on 01/09/2014, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. At the time of the event, Steam Generator Water level Protection Channel Testing was in progress. While testing was in progress with the 'C' Steam Generator Channel 1 Water Level Protection channel in trip for testing, a Turbine Trip occurred. The cause of the Turbine Trip is under investigation. The (Turbine Driven and Motor Driven) Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. The pressurizer Power Operated Relief Valves (PORVs) and the Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. Currently, the unit is stable in Hot Standby (Mode 3). This event is being reported as a 4 hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B) due to the valid Reactor Protection System Actuation. This event is being reported as an 8 hour Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A) due to the valid actuation of Auxiliary Feedwater System. At no time during this occurrence was the public or plant staff at risk as a result of this event. The NRC Resident Inspector has been notified. State and local authorities will be notified. Estimated restart date is 1/12/2014
ENS 496976 January 2014 19:09:00Beaver ValleyAutomatic ScramNRC Region 1Westinghouse PWR 3-Loop

At 1659 EST hours on January 6, 2014, Beaver Valley Power Station Unit 1 automatically tripped from 100% power. The cause of the reactor trip was a main transformer differential trip. All rods fully inserted into the core and the plant is stable in Mode 3. All three auxiliary feedwater pumps automatically started as expected. Normal and Emergency Busses are being powered by Offsite Power. The cause of the main transformer differential trip is being investigated. All other equipment functioned as expected. At 1757 EST hours the Emergency Operating Procedures were exited. Resident inspector has been notified. Decay heat is being removed via the turbine bypass valves to the condenser. No primary or secondary safety valves lifted. Unit 2 was unaffected.

Licensee notified the States of Pennsylvania, Ohio, and West Virginia and the Counties of Beaver, PA, Hancock, OH, and Columbiana, OH.

ENS 495065 November 2013 21:31:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1800 hours EST on 11/05/2013, with the unit in Mode 1 at 19% power, an automatic reactor trip occurred. Operators were transferring loads from the Startup Transformer to the Unit Auxiliary Transformer in accordance with normal operating procedures. When breaker 52/7, Unit Aux to 4KV Bus 1 Breaker, was taken to the close position, indication on the Reactor Turbine Generator Board (RTGB) went from 'Open' to 'No' indication. Breaker 52/12, Incoming Line Startup Transformer No. 2, cycled open and then re-closed. This resulted in a momentary loss of power to 4KV Bus 2 and 4KV Bus 1. The reactor trip signal was based on a loss of 4KV bus voltage to 2 of the 3 required 4KV buses. The cause of the loss of 480V bus E-1 was a result of loss of power to 4KV Bus 2. As a result of the loss of 480V bus E-1, the 'A' Emergency Diesel Generator (EDG) auto started. The required loads sequenced onto the 'A' EDG with the exception of the 'A' Service Water (SW) pump. The cause of the failure of the 'A' SW pump is under investigation. The one running Main Feedwater Pump ('A' Pump) tripped on the resulting under voltage of 4 kV Bus 1. By design, this condition resulted in an automatic start of Auxiliary Feedwater due to both Main Feed pump breakers being opened. Both 'A' and 'B' Motor-Driven Auxiliary Feedwater (AFW) Pumps started as designed. Steam generator water levels were maintained in the normal operating band. Currently, the unit is stable in Hot Standby (Mode 3). This event is being reported as a 4 hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B) due to the valid Reactor Protection System Actuation This event is being reported as an 8 hour Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A) due to the valid actuation of AFW and EDG auto-start and subsequent starting of required under voltage loads. At no time during this occurrence was the public or plant staff at risk as a result of this event. The (NRC ) Resident Inspector has been notified. The reactor trip was uncomplicated and the is plant is stable in mode 3 with decay heat being released to the main condenser. Normal offsite power is available with the exception of the 480V bus E-1 being supplied by the "A" Emergency Diesel Generator.
ENS 495055 November 2013 19:27:00Beaver ValleyAutomatic Scram
Manual Scram
NRC Region 1Westinghouse PWR 3-Loop

On 11/5/13 at 1746 (EST), Beaver Valley Unit 1 Generator tripped on 'C' Transformer differential protection. The Unit was at 47%, so an automatic trip of the reactor was not expected. The reactor was tripped manually. Entered E-0 Reactor Trip Response. The Aux Feedwater Turbine Driven Pump automatically started as required and one motor driven pump was manually started. At 1748, the control room received report of fire in the Turbine building. The deluge system had actuated and extinguished the fire in the Turbine plant mezzanine. Off-site assistance was staged and available. At 1828, an Unusual Event was declared when fire brigade reports indications that an explosion and fire had occurred in a cable tray in the Unit 1 Turbine Mezzanine.

The (NRC) Resident Inspector was notified of the events that occurred. At 1836, NRC (Resident Inspector) was notified of the Unusual Event. Cause of the cable tray explosion and fire is being investigated. The reactor trip was uncomplicated and Unit 1 is stable in mode 3 with decay heat being released to the main condenser. Normal offsite power is available. Unit 2 continues to operate at 100% and there is no impact on the health and safety of the public.

  • * * UPDATE FROM J. A. DAUGHERTY TO JOHN SHOEMAKER AT 2005 EST ON 11/05/13 * * *

At 1959 on 11/5/13, the licensee terminated the Unusual Event. The licensee has notified the NRC Resident Inspector. Notified the R1DO (Noggle), NRR EO (Skeen), IRD (Marshall). Notified the DHS SWO, FEMA, DHS NICC, and Nuclear SSA via e-mail.

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 4907528 May 2013 18:09:00North AnnaManual ScramNRC Region 2Westinghouse PWR 3-LoopOn May 28, 2013, at 1507 (EDT), Unit 2 was manually tripped from approximately 98 percent power due to decreasing steam generator levels as a result of a main feedwater system transient. The main feedwater system transient was initiated when the 'C' Main Feedwater Pump Discharge Motor-Operated Valve, 2-FW-MOV-250C, spuriously closed. The cause of the spurious closure of 2-FW-MOV-250C is unknown at this time. The Operations crew entered the reactor trip procedure and stabilized Unit 2 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. The reactor protection system actuation is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater (AFW) pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the AFW system is reportable per 10CFR50.72 (b)(3)(iv)(A) for a valid actuation of an ESF system. The AFW pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 2 is in the normal shutdown electrical line-up. Unit 1 was not affected by this event. The licensee notified the NRC Resident Inspector.
ENS 4902110 May 2013 12:42:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-LoopThis is a non-emergency event notification to the NRCOC (Nuclear Regulatory Commission Operations Center) in accordance with 10 CFR 50.72(b)(2)(iv)(B) for a valid actuation of the Reactor Protection System (RPS) (four hour notification). On 5/10/13 (at 1108 EDT) Unit 3 was stabilizing at 25% (power) during a controlled shutdown from an initial power level of 38% (power). Prior to the planned insertion of the trip at 25% (power), a malfunction of the turbine valves (not yet understood) caused a loss of load and a manual reactor trip was performed at 1108 (EDT). Unit 3 is currently in Mode 3 (Hot Standby) and stable. Source Range Nuclear Instrument N-3-32 malfunctioned (loss of detector voltage). All other plant systems are working as designed. Unit 3 is stable in Mode 3 (Hot Standby) at normal operating temperature and pressure. All control rods fully inserted. All other Nuclear Instrumentation is operable. Decay heat removal is via main feedwater and atmospheric steam dumps however, the condenser steam dumps are available but not preferred for Tave control. There is no known primary to secondary leakage. Unit 3 is in a normal shutdown electrical lineup. The cause of the turbine valves malfunction is not known at this time. There is no impact on Unit 4. The licensee has notified the NRC Resident Inspector.
ENS 4902010 May 2013 08:20:00North AnnaManual ScramNRC Region 2Westinghouse PWR 3-LoopOn May 10, 2013 at 0612 hours (EDT), Unit 2 was manually tripped from 60% power due to increased vibrations and a report of arcing on bearing #9 of the main turbine. Unit 2 was in the process of increasing power following a refueling outage when this event occurred. The Operations crew entered the reactor trip procedure and stabilized Unit 2 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. This reactor protection system actuation is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater system is reportable per 10CFR50.72(b)(3)(iv)(A) for a valid actuation of an ESF system. The Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 2 is in a normal shutdown electrical lineup. The #9 bearing is on the main generator exciter. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector and will be notifying local government agencies.
ENS 4881712 March 2013 17:31:00Turkey PointAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopUnit 3 experienced an automatic reactor trip from approximately 3% power. The turbine was off-line and undergoing testing following a control valve position indication repair. A turbine inlet pressure channel perturbation momentarily actuated the 'Reactor Trip by Turbine Trip' protection, causing the Unit 3 reactor to trip. Unit 3 is currently stable in Mode 3. No other complications were noted following the reactor trip. The cause of the turbine inlet pressure channel perturbation is currently under investigations. All control rods fully inserted. Normal feedwater is supplying the steam generators and decay heat removal is to atmosphere via the atmospheric steam dumps. There is no primary to secondary leakage. Unit 3 is in a normal shutdown offsite electrical power alignment. There is no impact on Unit 4. The NRC Resident Inspector has been notified.
ENS 4876418 February 2013 03:19:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-LoopThis is an non-emergency event notification to the NRCOC in Accordance with 10 CFR 50.72(b)(2)(iv)(B) for a valid actuation of the Reactor Protection System (RPS) (four hour notification) and 10 CFR 50.72(b)(3)(iv)(A) for a valid Engineered Safeguards (ESF) actuation due to Auxiliary Feedwater (AFW) initiation (eight hour notification). On 2/18/2013 at 0055 Unit 3 commenced a unit shutdown due to high # 1 seal leak-off on the 3A Reactor Coolant Pump (RCP). At 0129 3A RCP #1 seal leak-off exceeded operating limits. Unit 3 reactor was manually tripped. Auxiliary feedwater actuated as designed based on steam generator levels as a result of the trip. Unit 3 is currently in Mode 3 and stable. No complications were noted as a result of the reactor trip. Steam generator levels are being returned to normal level following the reactor trip. All other plant systems functioned as designed during and after the reactor trip. 3A RCP high #1 seal leak-off is under investigation. All control rods fully inserted on the trip. Decay heat was removed by AFW to the steam generators out the atmospheric reliefs. There was no primary to secondary leakage. Electrical buses are being supplied via offsite power. The licensee notified the NRC Resident Inspector.
ENS 4874412 February 2013 00:45:00Turkey PointAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopThis is a non-emergency event notification to the NRCOC (Nuclear Regulatory Commission Operations Center) in accordance with 10 CFR 50.72(b)(2)(iv)(B) for a valid actuation of the Reactor Protection System (RPS) (four hour notification) and 10 CFR 50.72(b)(3)(iv)(A) for a valid Engineered Safeguards (ESF) actuation due to Auxiliary Feedwater (AFW) initiation (eight hour notification). On 2/11/2013 at 2236 EST Unit 3 experienced a loss of condenser vacuum. An automatic Turbine Trip signal was generated which initiated a Reactor Trip signal. The reactor tripped as required on the automatic trip signal. Auxiliary feedwater water actuated automatically based on low steam generator levels following the trip. Unit 3 is currently in Mode 3 and stable. Steam generator levels have been returned to normal levels . The loss of vacuum is under investigation at this time. All other plant systems are working as designed. The reactor trip was uncomplicated. All control rods fully inserted. The plant is stable at normal temperature and pressure. AFW has been secured and main feedwater is being used for Steam Generator water level control. The main condenser is unavailable and decay heat removal is to atmosphere via the atmospheric steam dumps. There are no indications of primary to secondary leakage. Unit 3 is in a normal offsite power electrical alignment. There is no impact on Unit 4. The licensee has notified the NRC Resident Inspector.
ENS 4843624 October 2012 02:40:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

On 10/24/12 at 0147, North Anna Unit 2 reactor tripped automatically. The reactor first out is the 'C' steam generator lo-lo level. The turbine first out is reactor tripped, turbine trip. The event was apparently initiated by a loss of load on the secondary side. The cause of the loss of load is still being investigated. All systems responded as expected. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) due to actuation of the Reactor Protection System. The Auxiliary Feedwater pumps received an automatic start signal due to low-low level in all steam generators at the time of the trip, Steam generator levels have been restored to normal operating level. The Auxiliary Feedwater System operated as designed with no abnormalities noted. This event is reportable per 10 CFR 50.72(b)(3)(iv)(A) due to actuation of an ESF system. All control rods inserted into the core at the time of the trip and decay heat is being removed via the main condenser steam dumps. Several secondary (feedwater) relief valves lifted and reseated during the event. North Anna Unit 2 is currently stable at no load temperature and pressure in mode 3. At 0147 EDT, the Unit 2 Pressurizer Power Operated Relief Valve (PORV) , 2-RC-PCV-2455C, opened during an automatic reactor trip of Unit 2. The valve indicated open for less than 1 second. During this time, the identified leakage threshold for EAL SU6.1 (25 gpm) was exceeded. The cause of the loss of secondary load, which is believed to have caused the low steam generator water level and the lifting of the pressurizer PORV, is still under investigation. The licensee is focusing on the high pressure to low pressure turbine intercept valves or reheat valves going shut for reasons unknown at this time. The licensee's data shows that a pressurizer PORV opened momentarily. The instantaneous leak rate exceeded the unusual event threshold leak rate of 25 gpm. The PORV reseated and no ongoing leakage occurred during the transient. The rest of the transient was characterized as uncomplicated. The unit is in a normal post-trip electrical configuration. All systems functioned as required. There was no impact on Unit 1. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE AT 1346 EDT ON 10/24/12 FROM PAGE KEMP TO S. SANDIN * * *

The licensee is updating their report to RETRACT the portion related to the after-the-fact entry into EAL SU6. At 0147 hours EDT on 10-24-12, a Unit 2 Pressurizer Power Operated Relief Valve, 2-RC-PCV-2455C, opened during automatic reactor trip. The valve indicated open for less than 1 second. 2-RC-PCV-2455C opened as designed in response to the plant trip and allowed a small amount of water to transfer to the Pressurizer Relief Tank, as designed. The Pressurizer Power Operated Relief Valve subsequently re-closed and remains available for automatic operation, if needed. Initially, this issue was reported to the NRC at 0240 hours on 10-24-12 as an After-The-Fact Unusual Event for EAL SU6.1. Subsequent review has determined that the Pressurizer Power Operated Relief Valve functioned as designed and the small amount of inventory was transferred to the Pressurizer Relief Tank as designed and therefore does not meet the criteria for an Unusual Event and this notification is being retracted. NEI 99-01, Rev. 5 provides additional guidance that relief valve normal operation should be excluded from this Initiating Condition. However, a relief valve that operates and fails to close per design should be considered applicable to this Initiating Condition if the relief valve cannot be isolated. In this case, the Pressurizer Power Operated Relief Valve operated as designed and returned to automatic operation. The licensee informed state and local agencies and the NRC Resident Inspector. Notified R2DO (Musser).

ENS 4778128 March 2012 17:14:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1503 hours EDT on March 28, 2012, with the unit in Mode 1 at 55% power, an automatic reactor trip occurred. The reactor trip was the result of a turbine trip from a 'B' Steam Generator Hi Level. All Reactor Coolant Pumps (RCP) remained running. The Auxiliary Feeedwater (AFW) System automatically actuated due to both main feedwater pump breakers opening from a valid feedwater isolation signal. Steam Generator Levels were then controlled by Auxiliary Feedwater pumps. Steam Generator Blowdown was automatically isolated with the AFW actuation. The RCS Code Safety valves, Pressurizer Power Operated Relief Valves (PORVs), Steam Generator PORVs or the Main Steam Safety valves (MSSVs) did not open during the event. All control rods fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently in Mode 3 and stable. There were no radiological consequences or releases as a result of this event. The cause of the Steam Generator Hi Level is under investigation. The Resident NRC Inspector has been informed.
ENS 4729326 September 2011 14:48:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1145 hours EDT on September 26, 2011, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. All Reactor Coolant Pumps (RCP) remained running. The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. The steam generator and pressurizer Power Operated Relief Valves (PORVs) and the Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3. The cause of the reactor trip is under investigation.
ENS 4720126 August 2011 16:23:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

On August 23, 2011 at 1351 hours, North Anna Power Station experienced a seismic activity event which resulted in a loss of offsite power and automatic reactor trip of both units. At 1403 hours, an Alert was declared, based on Shift Manager judgment, due to significant seismic activity on the site. Subsequent to the earthquake, both units were stabilized and offsite power was restored. Following the event, seismic data was retrieved from the installed monitoring system and shipped to the vendor to determine the response spectrum for the event. On August 26, 2011 at 1340 hours, initial reviews of the data determined that the seismic activity potentially exceeded the Design Basis Earthquake magnitude value above 5 Hz. Therefore, this is reportable per 10CFR50.72(b)(3)(ii) (B) for the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. North Anna Unit 1 is currently in Cold Shutdown with the Residual Heat Removal System providing core cooling. North Anna Unit 2 is currently in Hot Shutdown and will be taken to Cold Shutdown with the Residual Heat Removal System providing core cooling. No significant equipment damage to Safety Related system (including Class 1 Structures) has been identified through site walk-downs nor has equipment degradation been detected through plant performance and surveillance testing following the earthquake. Therefore, there is reasonable assurance that the Safety Related systems are fully functional. The Spent Fuel Pit cooling system also remains fully functional and the temperature of the Spent Fuel Pit remained unchanged during the event. The vendor will complete the analysis of the seismic data and this information will be utilized to address the long term actions following the earthquake. The licensee notified the NRC Resident Inspector.

* * * UPDATE FROM DON TAYLOR TO PETE SNYDER AT 1739 EDT ON 9/9/11 * * * 

This is an update to EN 47201 reported on 8/26/2011 where It was reported that North Anna potentially exceeded the Design Basis Earthquake (DBE) magnitude value above 5 Hz. The vibratory motion from the 5.8 magnitude earthquake were recorded in all three orientations at several locations in the plant using two types of instruments: the Engdahl scratch plates that record 12 discrete spectral accelerations between 2 and 25.4 Hz, and the Kinemetrics analog recorders that recorded time histories of the accelerations. Based on evaluation of recorded plant data, it is concluded that the Central Virginia earthquake of 8/23/2011 exceeded the spectral accelerations for the Operational Basis Earthquake (OBE) and DBE of North Anna Plant. Extensive actions are underway to inspect. evaluate, test, and repair if necessary. systems and components to ensure they are capable of performing their required functions. To date, no significant damage to safety related structures, systems or components (SSC) has been identified. The licensee notified the NRC Resident Inspector. Notified R2DO (Rich).

ENS 4718123 August 2011 14:24:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

At 1403 hrs. EDT, North Anna Power Station declared an Alert due to significant seismic activity onsite. The Alert was declared under EAL HA6.1. Both units experienced automatic reactor trips from 100% power and are currently stable in Mode 3. All offsite electrical power to the site was lost. All four emergency diesel generators (EDG) automatically started and loaded and provided power to the emergency buses. While operating, the 2H EDG developed a coolant leak and was shutdown. As a result, the licensee added EAL SA1.1 to their declaration. All control rods inserted into the core. Decay heat is being removed via the steam dumps to atmosphere. No personnel injuries were reported.

  • * * UPDATE FROM ROBERT RINK TO HOWIE CROUCH AT 1116 EDT ON 8/24/11 * * *

The licensee has downgraded the Alert to a Notification of Unusual Event based on equipment alignments and inspection results. The licensee notified R2 IRC. Notified IRD (Marshall), NRR (Thorp), FEMA (Hollis), DHS (Inzer), USDA (Ferezan), HHS (Willis) and DOE (Parsons).

  • * * UPDATE FROM ROBERT RINK TO HOWIE CROUCH AT 1317 EDT ON 8/24/11 * * *

The licensee has exited the Notification of Unusual Event at 1315 EDT. The exit criteria was that all inspections and walkdowns were completed and plant conditions no longer meet the criteria for a NOUE. Notified R2DO (Widmann), IRD (Marshall), NRR (Thorp), FEMA (Hollis), DHS (Inzer), USDA (Ferezan), HHS (Willis) and DOE (Jackson).

  • * * UPDATE FROM DON TAYLOR TO DONALD NORWOOD AT 1405 EDT ON 8/26/11 * * *

This notification is to report new information identified post event that a condition existed which met the emergency plan criteria but was not declared. On August 23 at 1403 EDT, North Anna Power Station declared an Alert due to seismic activity onsite. The Alert was declared under Emergency Action Level (EAL) HA6.1 "Other conditions existing which in the judgment of the SM warrant declaration of an alert. Initial review of seismic response data from the earthquake on 8/23/11 (1348 hours) indicates that design spectrum input assumptions (i.e. Design Basis Earthquake (DBE) limits) may have been exceeded above 5 HZ. This would have resulted in classification of an Alert under EAL HA1.1. No significant equipment damage to safety related systems (including class I structures) has been identified through site walk-downs nor has equipment degradation been detected through plant performance and surveillance testing following the earthquake. The licensee notified the NRC Resident Inspector. The licensee also plans on notifying the State Emergency Operations Center and the Louisa County County Administrator. Notified R2DO (Widmann) and NRR EO (Bahadur).

ENS 4676116 April 2011 19:24:00SurryAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

At 1849 hrs, Surry Power Station (SPS) Unit 1 and Unit 2 experienced an automatic Reactor Trip from a Loss of Offsite Power, as a result of a tornado touching down in the station's switchyard. Unit 1 reactor tripped as a result of a Loss of Coolant Flow > P-8 (35% power), and the Unit 2 reactor tripped as a result of a 500 kV Leads Differential Turbine-Generator trip. Both units responded as designed. Unit 1 electrical power is being provided by Number 1 Emergency Diesel Generator (EDG) to the 1H emergency bus, with the Station Blackout (SBO) diesel loaded on to the 1J emergency bus. Unit 2 electrical power is being supplied by the number 2 EDG to the 2H emergency bus, with the number 3 EDG loaded on to the 2J emergency bus. All Unit 1 control rods inserted on the reactor trip, and all Unit 2 control rods inserted on its respective reactor trip. The Low Level Intake Structure (LLIS) is without power. All three Emergency Service Water Pumps are running to supply the intake canal. Efforts are underway to restore Bus 7, which will give each unit an emergency bus powered by offsite power (Unit 1 1J, Unit 2 2H) and restore power to the LLIS. Decay heat is being removed by auxiliary feedwater on both units and atmospheric steam release via the steam generator PORVs. Both units are currently on natural circulation. All other system parameters are normal and stable. At 1855 hrs a NOUE was declared due to a loss of offsite power (applicable to U1 and U2). Additionally, due to an estimated 100 gallon fuel oil spill from an above ground storage tank near the station's garage, the Virginia State Department of Environmental Quality was notified at 2041 and the Surry County Local Emergency Planning Coordinator was notified at 2114. At 2334, the Virginia State Department of Environmental Quality was notified and the Surry County Local Emergency Planning Coordinator was notified at 2336, due to an estimated 200 gallon oil leak to the ground from a station switchyard transformer damaged during the tornado. The NRC Resident Inspector has been notified and is on-site. Notified DHS (Rickerson), FEMA (Boscoe), DOE (Turner), HHS (Hoskins), and USDA (Russell). See related EN #46762

  • * * UPDATE FROM TUCKER CARLSON TO CHARLES TEAL ON 4/19/11 AT 0756 * * *

The licensee exited the emergency condition at 0745 EDT on 4/19/11. Offsite power has been restored, and the plant is shutdown and cooled down. Notified R2DO (O'Donohue), NRR EO (Thorpe) and IRD (Gott). Informed the following Federal Agencies via Blast Dial: DHS, FEMA, USDA, HHS and DOE.

ENS 4674410 April 2011 06:40:00Beaver ValleyManual ScramNRC Region 1Westinghouse PWR 3-LoopOn April 09, 2011, at 2349 (EDT), Beaver Valley Power Station (BVPS) Unit No. 2 was operating at 15% power while preparing to synchronize the main unit generator to the grid. At that time, the 'A' Auxiliary Feedwater Injection Header was declared inoperable due to a water leak identified from a vent valve fillet weld between the inside and outside Containment Isolation Valves (outside of containment) for containment penetration X-79. In accordance with Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System, Condition D, at 0345 (EDT), April 10, 2011, BVPS Unit 2 commenced a Reactor Shutdown to Mode 3. Required action is to be in Mode 3 within 6 hours. This event is being reported as a Technical Specification required shutdown pursuant to 10CFR50.72(b)(2)(i), 4 hour notification. Repairs are in progress. The following additional shutdown actions may be required from the time the Injection Header/Containment Penetration was declared inoperable: Technical Specification 3.7.5, Condition D, Mode 4 in 18 hours and Technical Specification 3.6.1, Condition A, Mode 5 within 37 hours. This event is also being reported as a degraded condition for Containment pursuant to 10CFR50.72(b)(3)(ii)(A), 8 hour notification. Additionally, at 0357 (EDT), during the Reactor Shutdown, at 4.6% Reactor Power, the BVPS Unit 2 Reactor was manually tripped due to reaching a pre-established manual trip criteria of 25% Steam Generator Level for the 21A Steam Generator. This was conservative criteria set above automatic actuation setpoint of 20.5% level. This event is being reported as a RPS Actuation pursuant to 10CFR50.72(b)(2)(iv)(B), 4 hour notification. Control room personnel entered Emergency Operating Procedure E-0, 'Response to Reactor Trip and Safety Injection.' Safety systems and equipment functioned as designed following the manual reactor trip. Due to the cooldown and subsequent shrink of level in the 21A Steam Generator, an automatic start of the Steam Driven Auxiliary Feedwater Pump (2FWE-P22) occurred at 20.5%. This event is being reported as an Auxiliary Feedwater System Actuation pursuant to 10CFR50.72(b)(3)(iv)(A), 8 hour notification. All control rods fully inserted into the core. The plant electrical system is aligned to normal offsite power sources. Decay heat from the reactor coolant pumps is being directed to atmospheric dump valve. There is no primary to secondary leakage. There was no impact on Unit 1. The licensee notified the NRC Resident Inspector.
ENS 466606 March 2011 19:38:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-Loop

This is a 4-hr Non-Emergency notification to the NRCOC (Nuclear Regulatory Commission Operations Center) for an event that results in actuation of the Reactor Protection System (RPS) when the reactor is critical in accordance with 10CFR50.72(b)(2)(iv)(B). On 3/6/11 at approx 16:20 (EST), Steam Generator sodium concentrations started to rise and exceeded 3-ONOP-071.1 (Secondary Chemistry Deviation from limits) Action Level 3 criteria (250 ppb Sodium). The plant power was reduced to 25% per 3-ONOP-100, Fast Load Reduction, and a manual plant trip (was) initiated per procedure at 16:44 (EST). Unit (3) is stabilized in Mode 3, and (the licensee) is performing secondary clean-up.

All rods fully inserted. All safety systems functioned as required. The reactor trip was uncomplicated. Unit 4 was unaffected by this event.

The licensee has notified the NRC Resident Inspector.

ENS 465842 February 2011 08:52:00SurryAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopUnit 2 Reactor automatically tripped at 0533 EST. This was due to loss of coolant flow in the 'C' RCS Loop. The first indication of the reactor trip was the annunciator for 'Loss of Coolant Flow > P8.' The 'C' RCP is running with motor current indicating normal. 'C' LOOP RCS flow is approximately 25% on all three channels with 'A' & 'B' RCS LOOP FLOW approximately 104%. All three auxiliary feedwater (AFW) pumps automatically initiated as designed on low-low steam generator level following the trip. Currently, RCS temperature is being maintained stable at no load temperature. All systems responded as expected with the exception of the Intermediate Range Neutron Indication. N36 indication was undercompensated and Source Range indication did not automatically energize, but was subsequently manually energized. This notification is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for 4-hour notification of RPS Activation and 10 CFR 50.72(b)(3)(iv)(A) for 8-hour notification of actuation of AFW. Plant responded as expected. The NRC Resident Inspector has been notified of this event and is on site. There were no radiation releases due this event, nor were there any personnel injuries or contamination events. This was an uncomplicated reactor trip and all control rods fully inserted. The plant is in a normal electrical alignment. AFW has been secured and the steam generators are being feed from main feedwater. Decay heat is being released through the main condenser steam dumps. Estimated time for repair and re-start is not known.
ENS 4647110 December 2010 01:25:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-LoopAt 2200 (EST) 12/09/2010, Unit 4 had indication of a condenser tube leak. A power reduction was commenced in accordance with plant procedures to allow isolation of the leaking waterbox. Sodium levels in the Steam Generators increased and a unit shutdown was required. Power was reduced to 20% and a manual reactor trip was initiated at 2258 (EST) 12/09/10, in accordance with plant procedures. All systems operated as required and the unit is stable in mode 3. The reactor trip was not complicated. All control rods inserted fully and decay heat is being removed by the atmospheric steam dumps. There is no indications of primary to secondary leakage. Normal offsite power is available and Unit 3 is unaffected. 3 of 3 steam generators are affected by the increase in sodium levels. The licensee is in progress of conducting steam generator blowdowns. The licensee has notified the NRC Resident Inspector.
ENS 464678 December 2010 16:51:00HarrisManual ScramNRC Region 2Westinghouse PWR 3-LoopText provided by the licensee. Quotations omitted for readability. Report Type: This 60-day telephone notification is being made under 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1). Description: On October 28, 2010, during the performance of MST I0073, "Train 'B' 18 Month Manual Reactor Trip, Solid State Protection System Actuation Logic & Master Relay Test", two sequential errors resulted in the inappropriate activation of the 'B' ESW pump. During night shift on October 27/28th, the Master Relay Selector Switch was not returned to the required OFF position. This caused day shift to find one of the two general warning lights to be lit. During the troubleshooting for the light, a technician discovered the Master Relay Selector Switch out of the expected OFF position as required by the procedure. A technician moved the switch to the OFF position outside of procedural guidance, resulting in the partial activation of the Reactor Protection System, including the 'B' ESW pump. The plant was in Mode 6 due to refueling outage 16 during the event. Actual plant conditions and parameters did not exist that required an automatic start of the 'B' ESW Pump. Therefore, this actuation is classified as invalid. The system started and functioned successfully. This invalid actuation was entered into the corrective action program as NCR 430289. Cause: Poor performance of task by the individual. The licensee informed the NRC Resident Inspector.
ENS 4641915 November 2010 07:06:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-LoopOn 11/15/10 at 0603 (hrs. EST), the Unit 3 reactor was manually tripped after receiving a report that there was an overheating packing gland on the 3A2 Circulating Water Pump. The 3A1 Circulating Water Pump was already tagged out for maintenance. This required Unit 3 to be tripped due to only two Circulating Water Pumps in operation. The Unit 3 reactor was stabilized in Mode 3. Auxiliary Feedwater automatically actuated. A train of normal feedwater remains available to feed Steam Generators. During the Unit 3 reactor trip, the 3B Steam Dump to Atmosphere failed to close on operator demand per procedure. It was locally isolated with a manual isolation valve stopping the cooldown. The RCS was stabilized at 487.56?F and was borated as required by procedure. All boration systems operated as designed. The RCS returned to normal operating temperature and pressure. During the trip, all rods inserted into the core. Other than the steam dump to atmosphere valve lifting, no other relief valves lifted. There is no known primary to secondary leakage. The electrical grid is normal and the plant is in a normal shutdown electrical line-up. Decay heat is being removed via the 3A and 3C steam dumps to atmosphere since the Main Steam Isolation Valves were closed. The plant is heating up to normal operating temperature at which time auxiliary feedwater will be secured. The licensee has notified the NRC Resident Inspector.
ENS 4635222 October 2010 09:46:00North AnnaManual ScramNRC Region 2Westinghouse PWR 3-LoopOn 10/22/2010 at 0636 hours, North Anna Unit-1 reactor was manually tripped during physics testing and 1-E-0 was entered due to problems with the Rod Control In Hold Out Switch. The out direction of the switch was not functioning properly and the reactor was tripped to put the plant in a condition to perform maintenance. All control rods fully inserted into the reactor core. This was an uncomplicated reactor trip with no automatic ESF actuation required. Unit 1 is currently stable at normal operating temperature and pressure in MODE 3 (Hot Standby). The plant electrical line-up is normal. Decay heat removal is via the steam dumps. Notification will be made to the local county administrator's office. The NRC Resident Inspector has been notified.
ENS 463137 October 2010 04:06:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 0013 hours EDT on October 7, 2010, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. Reactor Coolant Pump (RCP) 'C' tripped during the event. The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. Steam generator and pressurizer Power Operated Relief Valves (PORVs) and Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3. The Main Turbine Lube Oil Deluge System actuated without a fire during the event. A fire hose station pipe ruptured after the deluge system actuation. The fire hose station was isolated. In addition, the 'B' Main Feedwater (MFW) Pump tripped during the event and the 'A' Main Feedwater pump subsequently tripped due to high steam generator level in the 'C' Steam Generator at about 0023 hours. There is currently indication of RCP 'C' Number 2 seal leakage of approximately 2.5 gpm. At approximately 0405 EDT the Auxiliary Feedwater (AFW) system actuated due to a trip of MFW Pump 'A' while attempting to start the pump in accordance with procedure GP-004, Post Trip Stabilization. The AFW system actuation signal caused motor-driven AFW Pump 'B' to start. The motor-driven AFW Pump 'A' was already in operation due to the post-trip condition. The cause of the MFW Pump 'A' trip is under investigation. This AFW system actuation is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). The cause of the reactor trip is under investigation. The licensee notified the NRC Resident Inspector.
ENS 4624610 September 2010 19:46:00Turkey PointManual ScramNRC Region 2Westinghouse PWR 3-LoopDuring a Unit 4 reactor startup on 9/9/2010 while the reactor was subcritical in Mode 3, the Control Bank C Group 1 Step Counter failed (due to a battery failure). The reactor trip breakers were opened in accordance with procedures as required by the Action of Technical Specification 3.1.3.3. All rods fully inserted. The unit remained in Mode 3. This was a manual actuation of the reactor protection system. This report is made in accordance with 10 CFR 50.72(b)(3)(iv)(A) - Valid actuation of the Reactor Protection System. This is a late report. Reportability was not recognized at the time of the event as a result of misleading guidance in 0-ADM-115 'Notification of Plant Events' (site specific) procedure. A condition report is being initiated (by the licensee) to evaluate procedure changes to avoid further occurrences. The licensee notified the NRC Resident Inspector.