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The query [[Category:ENS Notification]] [[Site.Company::Dominion]] [[Scram::+]] was answered by the SMWSQLStore3 in 1.3802 seconds.


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 Entered dateSiteScramRegionReactor typeEvent description
ENS 539083 March 2019 00:13:00North AnnaManual ScramNRC Region 2On March 2, 2019 at 2237 EST, North Anna Unit 2 reactor was manually tripped, while operating at approximately 12 percent power, due to degrading vacuum in the main condenser. The unit was in the process of a planned shutdown for refueling when condenser vacuum degraded to greater than 3.5 inches of mercury absolute. 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. The reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). There were no ESF system actuations. Decay heat is being removed by the Steam Generator Pressure Operated Relief valves. Unit 2 is in a normal shutdown electrical lineup. The NRC Resident Inspectors have been notified. The Louisa County Administrator will be notified."
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 5216911 August 2016 11:09:00MillstoneManual ScramNRC Region 1CEReactor operators manually tripped the reactor due to the loss of two out of four circulating water pumps which caused a drop in condenser vacuum. The trip was uncomplicated. The reactor is shutdown and stable with decay heat removal via steam dumps to the condenser. The cause of the circulating water pump trips is currently unknown, but initial indications are that the pumps tripped due to a lightning strike that caused an electrical perturbation. The reactor will remain shutdown while the licensee investigates the cause. Unit 3 was not affected. The licensee notified the NRC Resident Inspector and the State and Local governments.
ENS 5192915 May 2016 07:08:00MillstoneManual ScramNRC Region 1Westinghouse PWR 4-Loop

At 0638 EDT on 5/15/2016, an Unusual Event (EAL GU.2) was declared on Millstone Unit 3 due to a Main Generator hydrogen gas leak into the Turbine Building. At 0645 EDT, operators manually tripped the reactor. All rods inserted. All systems functioned as expected following the reactor trip. Operators are currently venting the remaining hydrogen from the generator through the normal vent path. There is no safety related equipment out-of-service. The plant is in a normal post-trip electrical line-up. All Emergency Diesel Generators are available. The licensee notified the State of Connecticut and the NRC Resident Inspector. Notified DHS SWO, FEMA Ops Center, and DHS NICC Watch Officer. Notified via E-mail FEMA National Watch Center and NuclearSSA.

  • * * UPDATE FROM MIKE CICCONE TO HOWIE CROUCH AT 1044 EDT ON 5/15/16 * * *

At 0949 EDT, the licensee terminated the Unusual Event. At 0645 EDT on 5/15/16, a manual reactor trip was initiated at Millstone Unit 3 due to a hydrogen leak from the main generator. As expected, Auxiliary Feedwater System (AFW) initiated on the reactor trip. The trip was uncomplicated and the plant is currently in Mode 3 with a normal electric lineup and decay heat is being removed via steam dumps to the condenser. The cause of the hydrogen leakage is under investigation. This is reportable under 10 CFR 50.72(b)(2)(iv)(B) - RPS actuation while critical, and 10 CFR 50.72 (b)(3)(iv)(A) - valid specified system actuation. The licensee has notified Waterford Township, the State of Connecticut and the NRC Resident Inspector. Notified R1DO (Burritt), NRR ET (McDermott), NRR EO (Morris), IRD (Grant), DHS SWO, FEMA Ops Center, DHS NICC, FEMA National Watch Center (email) and NuclearSSA (email).

ENS 5168225 January 2016 03:21:00MillstoneAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopAt 0147 EST on 1/25/16, the 'B' Reactor Coolant Pump (RCP) tripped offline. This caused a reactor trip on low coolant flow. As expected, the Aux Feedwater System (AFW) initiated on the reactor trip. The trip was uncomplicated and the plant is currently shutdown in Mode 3 with a normal electrical lineup and decay heat is being removed via steam dumps to the condenser. The cause of the 'B' RCP trip is under investigation. A containment entry is planned for dayshift on 1/25/16 to inspect and troubleshoot the 'B' RCP. The Licensee has notified the NRC Resident Inspector.
ENS 515218 November 2015 02:55:00MillstoneManual ScramNRC Region 1CEDuring power ascension following refueling outage, a decreasing oil level in the 'C' Reactor Coolant Pump was noted. When the oil level reached 69 percent, with the reactor at approximately 56 percent rated thermal power, per plant procedure, a rapid downpower was initiated which brought the plant to approximately 15 percent power and a manual reactor trip was initiated at that point. The reactor trip was uncomplicated and all plant equipment responded as expected. The licensee notified 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 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 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 495259 November 2013 16:09:00MillstoneAutomatic ScramNRC Region 1CEMillstone Unit 2 automatically tripped following a turbine trip due to a loss of condenser vacuum. The loss of vacuum was caused by the trip of the "C" circ water pump with the "D" circ water pump out of service. The licensee is still investigating the trip of the "C" circ water pump. The MSIVs are open with steam generators discharging steam to the main condenser. Auxiliary feedwater automatically started as expected following the reactor trip. All rods fully inserted and there were no complications following the reactor trip. All systems functioned as required and the unit is stable in Mode 3. There was no impact on Unit 3. The licensee has notified state and local authorities and the NRC Resident Inspector.
ENS 492609 August 2013 22:27:00MillstoneAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopThe loss of a non-vital 480v bus resulted in a feedwater transient that caused steam generator water level to lower below the automatic reactor trip setpoint. The reactor trip and plant response was uncomplicated with all rods being inserted into the core. The auxiliary feedwater system automatically actuated and is currently being to used to feed the steam generators. Decay heat is being removed via the steam dumps to the condenser. No relief valves or safeties lifted during the transient. The plant is currently stable in Mode 3 at normal operating pressure and temperature and is in its normal shutdown electrical lineup. There was no impact on Unit 2 which is currently operating at 100% power. The licensee has notified state and local authorities and the NRC Resident Inspector.
ENS 5200012 June 2013 23:51:00MillstoneManual ScramNRC Region 1Westinghouse PWR 4-Loop

At 2013 EDT on 6/12/16, Millstone Unit 3 commenced a Technical Specification (TS) required shutdown due to excessive Reactor Coolant System leakage from the "A" Reactor Coolant Pump (RCP) third stage seal. The leakage from the third stage seal was approximately two gpm which is greater than the Technical Specification (TS) limit of less than one gpm. During the shutdown, oscillations developed in the Main Feedwater which required the operator to initiate a manual reactor trip. Unit 3 is currently stable in Mode 3. Decay heat is being released via the Steam Dumps to the Main Condenser. Normal offsite power is available and the unit is in a normal shutdown electrical line-up. The cause of the Main Feedwater oscillations is being investigated. The licensee notified the NRC Resident Inspector. The licensee notified State and local government agencies.

  • * * UPDATE FROM WALTER ORF TO DONALD NORWOOD AT 0129 EDT ON 6/13/2016 * * *

The following clarifies Feedwater isolation vs. Feedwater oscillation: At 2337 EDT on 6/12/16, a manual reactor trip was initiated on Unit 3 following feedwater isolation. As expected, Aux Feedwater system (AFW) initiated on the reactor trip. The trip was uncomplicated and the plant is currently in Mode 3 with a normal electric line-up and decay heat is being removed via steam dumps to the condenser. This is reportable under 10 CFR 50.72(b)(2)(iv)(B) - RPS Actuation - Critical and 10 CFR 50.72(b)(3)(iv)(A) - Valid Specific System Actuation." The Feedwater isolation occurred due to high Steam Generator water level. The licensee notified the NRC Resident Inspector. Notified R1DO (Arner).

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 4903916 May 2013 10:20:00MillstoneManual ScramNRC Region 1Westinghouse PWR 4-Loop

On May 16, 2013, at 0256 EDT, while in Mode 3 (subcritical), operators manually opened the reactor trip breakers as directed by procedure. During rod drop testing, demand position and digital rod position indication did not agree within procedural limits. As directed by procedure, operators opened the reactor trip breakers. The cause of the discrepancy is under investigation. The plant responded as expected and all rods inserted as required. The plant remains stable in Mode 3. This condition is reportable pursuant to 10 CFR 50.72(b)(3)(iv). The NRC Senior Resident Inspector has been notified.

  • * * RETRACTION FROM TODD STRINGFELLOW TO DANIEL MILLS ON 7/10/2013 AT 1418 EDT * * *

The purpose of this call is to retract the report made on May 16, 2013, Event Number 49039. Operators manually opened the reactor trip breakers in MODE 3 during hot control rod drop testing as directed by procedure. The condition has been determined to be an invalid actuation of the RPS since boron concentration, and not control rods, were being relied upon for safety function during rod testing. The details of the engineering review will be provided to the NRC Senior Resident Inspector. Notified R1DO (Cook).

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 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 4746120 November 2011 20:37:00MillstoneManual ScramNRC Region 1Westinghouse PWR 4-LoopMillstone Unit 3 was manually tripped at low power during startup following the loss of condenser vacuum. The loss of vacuum was caused by the loss of the auxiliary steam boiler. The auxiliary steam boiler supplies gland seal steam to the turbine at low power and without the gland seal steam, the licensee could not maintain a condenser vacuum. All rods fully inserted and there were no complications during the trip. All systems functioned as required and the unit is stable in Mode 3. There was no impact on Unit 2. The licensee has notified state and local authorities and the NRC Resident Inspector.
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 4697120 June 2011 13:01:00MillstoneManual ScramNRC Region 1CEAt 1152 EDT on 6/20/11 while operating at 60% power, the "B" MFW Pump tripped for reasons unknown. There was no maintenance or I&C work on-going at the time involving this pump. Operators initiated a manual reactor trip, however, (they) are not certain whether the automatic reactor trip setpoint of 49.5% Steam Generator Water Level Narrow Range (SGWL NR) was reached first. SGWL decreased to the Auxiliary Feedwater (AFW) setpoint of 26.8% NR causing the initiation of both motor-driven AFW Pumps. All Control Rods fully inserted. Unit 2 is currently stable in Mode 3, Hot Standby, removing decay heat via the Main Steam line to the Condenser. Operators secured AFW and will initiate feed to the Steam Generators using the "A" MFW Pump. Unit 2 is in a normal post-trip electrical lineup with all sources of offsite power available. The licensee has the cause of the "B" MFW Pump trip under investigation. The licensee informed both state/local (Waterford Dispatch) and the NRC Resident Inspector
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 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 4644128 November 2010 17:27:00MillstoneAutomatic ScramNRC Region 1CELoss of two circulating water pumps in one condenser caused a high main condenser backpressure. The high main condenser back pressure caused an automatic main turbine and reactor trip while critical >15%. Low steam generator water level following the trip caused an automatic Auxiliary Feedwater actuation. The low water level condition has cleared. Normal post-trip response has been verified and the plant is stable. The licensee removed one circulating water pump from service in preparation of performing scheduled maintenance when the other circulator unexpectedly tripped. The cause of the pump trip is under investigation. During the trip, all rods inserted into the core. There were no primary safety valves that lifted during the transient but main steam safeties "chattered" during the transient and have fully reseated. There is no known primary to secondary leakage. The electrical grid is stable and supplying plant loads in a normal shutdown electrical lineup. The reactor is stable at normal operating pressure and temperature with decay heat being removed via the steam dumps to condenser. The licensee has notified the State of Connecticut Department of Environmental Protection, the town of Waterford, CT, and 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 4602016 June 2010 21:08:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopOn 6-16-2010 at 1920 hours, Unit 2 experienced an automatic reactor trip/turbine trip from 98% power. A severe lightning storm was in progress at the time of the trip and a lightning strike appears to be the cause of the event. The reactor trip was actuated from Channel 1 and Channel 2 Over Temperature Delta T. All control rods fully inserted into the core during the trip. The control room staff responded to the trip in accordance with plant procedures and the unit is stable in Mode 3. This event is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps started as designed following the reactor trip and steam generator inventory was restored to normal operating level. The Auxiliary Feedwater pumps have been secured and returned to automatic. This event is reportable per 10CFR50.72(b)(3)(iv)(A) due to the ESF actuation. Decay heat is being removed by the condenser steam dump system. The 'A' loop wide range hot and cold leg thermocouples remain failed high and the 'B' loop wide range cold leg thermocouple also failed high during the event. The plant is in a normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector and will notify the local authorities. See EN #41898 for similar occurrence.
ENS 459868 June 2010 12:00:00SurryAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 0948 hours (EDT) on 6/8/10, a Unit 1 vital AC bus was lost when the uninterruptible power supply inverter failed while the alternate AC source was out of service for scheduled maintenance. The loss of the vital bus inverter caused a loss of 120 VAC vital bus 1-III. The loss of this vital bus caused the 'A' main feed pump recirculation valve to fail open and also caused 2 of the 3 main feedwater regulating valves to fail to automatic-hold mode of operation. This combination of as designed failures resulted in a reduction in main feedwater flow and resulted in an automatic reactor trip due to a feed flow steam flow mismatch in conjunction with low steam generator level. The loss of vital bus 1-III also resulted in initiation of safety injection due to loss of vital bus 1-III instrumentation in conjunction with the expected momentary RCS cooldown below 543 DEG-F. The safety injection resulted from the high steam flow in conjunction with low RCS T-ave actuation signal. The safety injection actuation also resulted in automatic start of the #1 Emergency Diesel and the #3 Emergency Diesel Generators. Neither EDG was required to load since off-site power remained operable. The loss of vital bus 1-III also resulted in loss of numerous field inputs to the Plant Computer System (PCS) and resulted in non-functionality of the SPDS (Safety Parameter Display System). The PCS itself remains functional along with MCR (Main Control Room) annunciators and sufficient MCR instrumentation to monitor critical safety functions. All three auxiliary feedwater pumps automatically initiated as designed on low-low steam generator level following the trip. Currently, RCS temperature is being maintained stable at 547 degrees. All systems functioned as required following the reactor trip. During the post-trip transient, pressurizer PORV (Power Operated Relief Valve), PCV-1455C, cycled as required to maintain RCS pressure due to the safety injection and the loss of normal letdown. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events. This event is being reported in accordance with 10CFR50.72(b)(2)(iv)(B), 10CFR50.72(b)(2)(iv)(A), 10CFR50.72(b)(3)(iv)(A), and 10CFR50.72(b)(3)(xiii). The NRC Resident Inspector was notified of this event. During the trip, all rods inserted into the core. In addition to the pressurizer PORV lifting, a secondary main steam relief valve lifted. All relief valves properly reseated and there is no known primary to secondary leakage. The plant is in its normal shutdown electrical lineup. Main steam trip valves were isolated during the transient. Decay heat is being removed via main steam bypasses to the condenser and steam generator power operated relief valves.
ENS 4596028 May 2010 03:43:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

A Unit-2 reactor trip was initiated by a loss of the Unit-2 'B' station service bus. The loss of the 'B' station service bus caused a reactor trip due to the loss of flow on one-of-three loops due to the loss of the 'B' Reactor Coolant Pump. The Auxiliary Feed Water system actuated as expected due to the reactor trip. The plant was stabilized in Mode 3 using the appropriate emergency procedure. During the transient, the 'B' Reserve Station Service Transformer de-energized and the Unit-2 'H' Emergency Diesel Generator was previously tagged out for planned maintenance. This resulted in the Unit-2 'H' emergency bus being de-energized. The alternate AC diesel generator has been placed in service and is providing power to the Unit-2 'H' emergency bus. The automatic tap changer for the 'C' reserve station service transformer did not work in automatic and had to be manually adjusted to control voltage. Unit-2 'C' Reactor Coolant Pump remains in service. All control rods fully inserted on the trip and no relief valves lifted or safety valves lifted in either the primary or secondary systems. The turbine drive and 'B' motor driven Auxiliary Feed Water pumps automatically started and injected into the 'A' and 'B' steam generators on a low level signal. The 'A' motor driven Auxiliary Feed Water pump failed to start due to the loss of the 'H' emergency bus. The 'C' steam generator is being controlled with main feed water though the 'C' main feed regulating valve bypass valve. Decay heat removal is via the condenser steam dumps. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM MICHAEL WHALEN TO HOWIE CROUCH @ 1707 EDT ON 5/28/10 * * *

EN#45960 reported the RPS Actuation (50.72(b)(2)(iv)(B)) and AFW System Actuation (50.72(b)(3)(iv)(A). The event occurred at 0003 EDT on May 28, 2010. Technical Specification (TS) 3.0.3 was entered at 0004 hours on May 28, 2010, for inoperable offsite power sources with the 2H emergency diesel generator (EDG) being inoperable per TS 3.8.1. M. Update: At the time of the event, the station was experiencing a severe lightning storm. The Auxiliary Feedwater System was returned to auto standby at 0558 hours. At approximately 0942 hours, RCS cooldown to Mode 4 was started on Unit 2. Mode 4 was entered at 1245 hours. The 'A' and 'B' RCPs remain secured in Mode 4. Following repairs and post maintenance testing the 'C' reserve station service transformer (RSST) was declared operable at 1324 hours. This restored two (2) qualified offsite circuits for Unit 1 and one (1) qualified offsite circuit for Unit 2. TS 3.0.3 was cleared at this time on Unit 2. The 'B' RSST remains out of service (OOS) pending repairs and testing. The Unit 2 'B' station service bus remains OOS. The 2H EDG previously reported OOS for scheduled maintenance is expected to be returned to service on Monday, May 31, 2010. The alternate AC diesel generator continues to supply power to the 2H emergency bus. Limiting action remains for one (1) offsite circuit for Unit 2 being inoperable along with the 2H EDG OOS. The licensee will be notifying the NRC Resident Inspector. Notified R2DO (Haag).

ENS 4594522 May 2010 17:46:00MillstoneManual ScramNRC Region 1CEThe licensee experienced a feedwater transient which initiated the event. All safety systems are available. All control rods fully inserted. The electrical lineup is normal. The decay heat path is through the condenser steam dumps. No relief valves or safety valves lifted during the transient. Primary plant temperature is 533 degrees Fahrenheit and primary plant pressure is 2256 psia. The licensee is investigating the cause of the feed transient. The licensee notified the NRC Resident Inspector, the Waterford Dispatch, and the State Department of Environmental Protection. Earlier, the licensee was experiencing oscillations in the feedwater regulating valve (FRV) for the #2 steam generator when the valve was in automatic control. Troubleshooting planning was underway but no troubleshooting activities were in progress at the time of the trip. When the operator placed the #2 steam generator FRV in manual control, the steam generator water level began to increase and could not be recovered. The operator then manually tripped the reactor prior to reaching the high steam generator level trip setpoint. An Auxiliary Feed Water system actuation did occur during the transient. The trip and plant response was considered uncomplicated.
ENS 4593117 May 2010 12:27:00MillstoneAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopSystem Affected: RPS actuation and reactor trip. Actuation Initiation Signals: AFW auto initiation. Cause: Low level 'C' steam generator. Effect of event on Plant: Reactor and turbine trip. Actions taken or Planned: Standard post trip actions. Additional information: Plant stabilized at NOP/NOT. The licensee experienced a feedwater transient which initiated the event. All safety systems are available. All control rods fully inserted. The electrical lineup is normal. The decay heat path is through the condenser steam dumps. There were no relief valve or safety valve lifted during the transient. Primary plant temperature is 555 degrees Fahrenheit, and primary plant pressure is 2250 psig. The licensee is investigating the cause of the feed transient. The licensee notified the NRC Resident Inspector, the Waterford Dispatch, and the State Department of Environmental Protection,
ENS 4572926 February 2010 11:48:00MillstoneManual ScramNRC Region 1CEThe "C" Circulating Water Pump tripped on high delta pressure across the screens. Another Circulating Water Pump was out of service for maintenance. This condition required a shutdown, therefore the reactor was manually tripped. This caused the steam generator level to reach the low level setpoint. There was a loss of Main Feedwater Pumps when breaking condenser vacuum. Auxiliary Feedwater Pumps started automatically upon low level in the steam generators. The letdown system was isolated post trip and the charging and letdown systems are restored to normal lineup. The plant is in Mode 3. All rods fully inserted. Decay heat is being removed via atmospheric dump valves. There are no known primary to secondary leaks in the steam generator tubes. The electrical lineup is normal with offsite power. The licensee has notified the NRC Resident Inspector.
ENS 4558320 December 2009 00:06:00MillstoneAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopAn automatic reactor trip occurred due to a turbine trip caused by a generator electrical fault trip. The cause of the electrical fault is under investigation. All rods fully inserted into the reactor. The auxiliary feedwater pump started and is maintaining steam generator level. The reactor is NOP and NOT. The post trip electrical line-up is being back-fed from off-site power through the RSST transformer. All other post trip actions are standard and all systems are operating as expected. There was no affect on Unit-2. The licensee will contact the NRC Resident Inspector.
ENS 455569 December 2009 17:15:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1423 hours on 12/9/2009, electrical supply breaker L102 was inadvertently opened which caused electrical Bus 3 and the 'C' Reserve Station Service Transformer to de-energize. This caused the loss of 'F' Transfer Bus which resulted in a loss of power to the 1H and 2J Emergency Busses and an automatic start of the 1H and the 2J Emergency Diesel Generators. Both emergency diesel generators started and re-energized their associated emergency bus as designed. The Unit 2 'G' Bus, which supplies power to the Unit 2 Circulating Water Pumps, did not automatically transfer to the 'B' Reserve Station Service Transformer in a sufficiently short time to prevent the loss of the Unit 2 Circulating Water pumps. The loss of the Unit 2 Circulating Water pumps resulted in an automatic low vacuum turbine trip and a subsequent (Unit 2) reactor trip due to the turbine trip. The 2 'G' Bus did automatically transfer to the 'B' Reserve Station Service Transformer and is currently energized. The Unit 2 Auxiliary Feedwater pumps automatically started and provided flow to the steam generators. There were no issues with the Auxiliary Feedwater System operation. The Unit 2 'A' Charging Pump and the Unit 2 'A' Component Cooling Water pump automatically started as designed due to the loss of power. The Unit 1 'B' Charging Pump and the Unit 1 'B' Component Cooling Water pump automatically started as designed due to the loss of power. The Unit 2 'C' Station Service Bus was lost following the trip when the electrical system automatically transferred to the Reserve Station Service transformers. With the 'C' Reserve Station Service Transformer de-energized the 'C' Station Service Bus was unable to transfer to an energized transformer. This resulted in the loss of the Unit 2 'C' Reactor Coolant Pump. The 'A' and 'B' Reactor Coolant Pumps remain in service at this time. The reactor trip is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater system, Emergency Diesel Generator system, Charging system actuations are reportable per 10CFR50.72(b)(3)(iv)(A). The electrical system is being returned to a normal lineup. The condensate and feedwater system remained in service to provide flow to the steam generators. Steam Dump operation to the condenser is not available due to low condenser vacuum, therefore steam is being released to the atmosphere from the Steam Generator Power Operated Relief Valves. The licensee suspects that switchyard maintenance activities caused the L102 trip which initiated the chain of events. All rods inserted into the core during the trip. During the transient, some secondary relief valves lifted and properly reseated. There is no known primary to secondary leakage. During the event call, the licensee reported that the 'C' Reserve Station Service Transformer was returned to service. The licensee notified the NRC Resident Inspector and will be notifying the Louisa County Administrator.
ENS 4552829 November 2009 13:31:00SurryManual ScramNRC Region 2Westinghouse PWR 3-LoopDuring startup physics testing, the reactor operator identified a discrepancy between Group 1 & Group 2 Step Demand Counters for control bank 'B'. Research identified the issue to be associated with failure of a card in the rod control power cabinet. Failure of the card impacts group step counters for control bank 'B', control bank 'D' and shutdown bank 'B.' Inoperability of more than one group step counter per bank placed Unit 2 in a 6-hour clock to hot shutdown as of 0808 hrs 29 November 2009. Consequently, at 1045 hrs, the decision was made to trip Unit 2 reactor. The decision was based on time required to make repairs, complete post maintenance testing requirements, and maintaining proper control of reactivity. All systems functioned as required on the trip and initiation of any auxiliary or emergency systems was not required. Current heat removal is via normal plant alignment following refueling outage activities: steam generator blowdown, main feedwater, and steam generator PORV. All control rods fully inserted on the manual trip. The licensee notified the NRC Resident Inspector.
ENS 4456411 October 2008 23:35:00MillstoneAutomatic ScramNRC Region 1Westinghouse PWR 4-Loop

At 2236 on 10/11/08 while reducing power for a planned refueling outage, "C" Steam Generator water level decreased and the turbine tripped. The reactor tripped on the turbine trip. Auxiliary feedwater initiated as expected and decay heat is being removed via the condenser steam dumps. All control rods fully inserted. No significant safety equipment is out of service and all safety buses are being supplied by offsite power. Emergency Diesel generators are available if needed. No PORVs or primary/secondary relief valves lifted. The licensee is evaluating cause of steam generator level excursion. Unit 2 is unaffected.

The licensee notified the NRC Resident Inspector.
ENS 4432628 June 2008 12:54:00MillstoneManual ScramNRC Region 1CEAt 1146, the reactor was manually tripped due to a feedwater transient that was caused by a loss of both feedwater pumps. The exact cause of the loss of both feedwater pumps is still being investigated. Prior to the reactor trip, the licensee was performing main turbine stop valve testing and during that testing received an isolation of extraction steam. On the manual reactor trip, all control rods fully inserted. No RCS PORVs or reactor safety valves lifted. The steam generator atmospheric dump valves did lift and reseat. Decay heat is being removed to the condenser via the turbine dump valves. Auxiliary Feedwater (AFW) initiated and fed the steam generators after the loss of both main feedwater (MFW) pumps. MFW was restored and is controlling steam generator levels. AFW has been secured. The electric plant is in a normal shutdown lineup and the EDGs are available. There was no effect on Unit 3. The licensee has notified the NRC Resident Inspector, Connecticut Department of Environmental Protection, and all local emergency response organizations.
ENS 4423824 May 2008 10:03:00MillstoneAutomatic ScramNRC Region 1CE

The licensee entered an Unusual Event after an automatic reactor trip due to a loss of offsite power per EAL Designation PU-1. Currently the plant is stable in Mode 3 with power being supplied to the "A" & "B" Emergency Diesel Generators (EDG) running supplying power to the safety buses. Decay heat is being removed by the motor driven Auxiliary Feedwater Pump supplying water to the steam generators exhausting to atmosphere via atmospheric relief valves. There are no primary to secondary tube leaks. NRC Resident Inspector has been notified.

  • * * UPDATE FROM SKIP JORDAN TO JOHN KNOKE AT1252 EDT ON 05/25/08 * * *

At 1252 EDT the licensee terminated from their Unusual Event. After the loss of offsite power due to a failure of the Reserve Station Service Transformer (RSST) the licensee began to troubleshoot the problem as well as beginning to restore offsite power. At 1106 on 5/24/08 a cross-tie from Unit 3 was connected to Unit 2 and "A" EDG was secured. At 0950 on 5/25/08 the RSST was powered up and the non-vital busses were energized. At 1136 the "B" and "D" Reactor Coolant Pumps were started. At 1230 the vital bus was energized and the "A" EDG was secured. The licensee notified the NRC Resident Inspector. Notifications to: R1 (Marc Dupas, David Lew), R1DO (Mel Gray), NRR EO (Fred Brown), IRD MOC (William Gott), DHS (Fred Hill), FEMA (Dan Sullivan), USDA (Amanda Jimenez,), HHS (Mason Pyle), DOE (Julia Muse). The NRC Resident Inspector was notified.

ENS 4415221 April 2008 00:37:00SurryManual ScramNRC Region 2Westinghouse PWR 3-LoopWhile ramping the Unit 1 Turbine following a forced unit outage, vibrations on the number 4 bearing increased to 13.9 mils. The ramp was stopped and a rapid load reduction was initiated. Due to sustained vibrations (>14.1 mils) after ramping the turbine down, the Unit 1 reactor was manually tripped. Unit 1 has been stabilized at Hot Shutdown. All three auxiliary feedwater pumps automatically initiated as designed on low-low steam generator level following the trip. Currently RCS temperature is being maintained stable at 547 degrees. All systems functioned as required following the reactor trip. There were no radiation releases due to this event, nor were there any personnel injuries or contamination events. This event is being reported in accordance with 10 CFR 50.72 (b)(2)(iv) and 10 CFR 50.72 (b)(3)(iv). Upon exiting the refueling outage in 11/07 the main turbine had a vibration issue of about 11 or 12 mils. Decay heat is being removed via the steam dumps to the condenser. Unit 2 was not affected during this event. Offsite power was lined up normally. The licensee notified the NRC Resident Inspector.
ENS 4386625 December 2007 23:22:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopOn 12/25/07 at 2110 hours EST, Unit 2 tripped from 100% power due to a trip of the 'B' Reactor Coolant Pump. The reactor trip 1st out annunciator was 'Loss of flow, power >30%'. All control rods fully inserted. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater (AFW) pumps auto started due to the event and the steam driven AFW pump subsequently tripped on overspeed. The steam driven AFW pump was reset and placed in service. The ESF (Engineered Safety Function) actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). The unit is currently in mode 3 and (the licensee is) investigating the cause of the ground on the 'B' reactor coolant pump. The plant is at normal operating pressure and temperature. The electrical grid is stable and supplying plant loads through the startup transformer. Decay heat is being removed via the steam dumps to the condenser with feedwater being supplied via the normal path. The licensee has notified the NRC Resident Inspector.
ENS 430723 January 2007 19:32:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopA Unit-1 reactor trip was initiated by a Process Rock card failure that caused the 'B' main feed regulation valve to fail closed. The closure of the 'B' main feed regulation valve caused a reactor trip due to a steam flow - feed flow mismatch with a low steam generator water level. The auxiliary feed water system actuated as expected due to the reactor trip. The plant was stabilized with no other issues using the appropriate emergency procedures. All control rods fully inserted on the trip and no relief or safety valves lifted in either the primary or secondary systems. Auxiliary feed water pumps automatically started and injected into the steam generators on a low water level signal. The operators restored normal feedwater flow to the steam generators. Decay heat is via the condenser steam dumps. The plant is aligned to the normal shut down electrical alignment. Card replacement is expected tonight and reactor startup is expected tomorrow. The licensee notified the NRC Resident Inspector.
ENS 4298310 November 2006 17:33:00KewauneeAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopOn 11/10/2006 with a shutdown in progress to repair a degraded bearing on the turbine generator, an automatic reactor trip occurred due to a power range nuclear instrumentation (NI) low range - high flux trip. Reactor power had just been lowered to below 10% power (P-10) where the power range (NI) low range trips become active. The bistable for power range NI N-42 had been tripped due to an unrelated failure on 11/09/2006. When P-10 automatically unblocked, a power range NI low range high flux reactor trip was generated. At the time of the trip, reactor power was well below the trip setpoint of 24.5% power. Following the trip, Main Feedwater Regulating Valve, FW-7A, did not automatically close as required on the reactor trip coincident with Low Tave (554F). The Reactor Operator reported FW-7A was mid-position and attempted to manually close FW-7A. It did not respond. As a result, levels in steam generator A rose to greater than 67%, which initiated feedwater isolation. The feedwater isolation signal tripped the running feedwater pump. With no feedwater pumps running, both Auxiliary Feedwater Pump A and Auxiliary Feedwater Pump B automatically started as required. The High-High steam generator level also resulted in a second reactor trip initiation signal. The Reactor Operator manually controlled Auxiliary Feedwater flow to steam generator A to restore normal level. Following the feedwater isolation, FW-7A fully closed. Following the trip, MS-201B1, the steam supply to main steam reheater B1 was locally isolated to limit the RCS cool down. This was a previously discussed contingency action. Main steam isolation valves remained open and normal condenser heat sink remained available. Further investigation as to the cause of the trip is in progress. Recovery actions per normal operating procedures are in progress. The plant was being shut down at a rate of a half percent power per minute at the time of the trip. All control rods fully inserted on the reactor trip and no safety or relief valves lifted. The plant was aligned for the normal shutdown electrical lineup prior to the trip. The temperature on the generator bearing reached a maximum of 190F with trip guidance set at 225F. The licensee notified the NRC Resident Inspector.
ENS 4294730 October 2006 11:56:00KewauneeAutomatic ScramNRC Region 3Westinghouse PWR 2-Loop

During power operation, at 92% rated power, an automatic reactor trip occurred. The reactor protection signal that caused the reactor trip was steam generator 'B' steam flow greater than feedwater flow coincident with low water level on steam generator 'B.' The cause of the plant transient that led to the reactor trip was a loss of Instrument Bus 1 (Red Channel). Instrument Bus 1 unexpectedly deenergized during the performance of maintenance on the inverter (BRA-111) that feeds Instrument Bus 1. Following the reactor trip, the auxiliary feedwater pumps automatically started, as designed, due to a low level in the steam generators. After the trip, non-safety related 4160 Volt AC Bus 4 de-energized and secondary plant feedwater heater 15B relief valve lifted. The cause of the loss of Bus 4 and 158 feedwater heater relief lifting Is under Investigation. The plant is currently stable and in the hot shutdown (HSD) mode. Power was restored to Instrument Bus 1 at 1018. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical because of the automatic reactor trip and under 10 CFR 50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any of the systems listed below because of the actions of the RPS and the automatic start of the AFW pumps. All control rods fully inserted. Decay heat is being removed by feeding the steam generators with AFW and steaming to the Condenser Dump. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM T. BUNKELMAN TO W. GOTT AT 1332 EST ON 10/30/06 * * *

Due to the loss of Bus 4, the running Circulating Water Pump was lost resulting in a loss of normal heat sink to the condenser. The standby Circulating Water Pump was started at 1002 CST and the condenser heat sink was restored. Until the condenser steam dump was restored, the plant was steaming through the steam generator PORVs (atmospheric steam dumps). There is no steam generator tube leakage. The licensee notified the NRC Resident Inspector. Notified R3DO (M. Phillips)

ENS 428887 October 2006 18:27:00SurryManual ScramNRC Region 2Westinghouse PWR 3-Loop

The main turbine cross-under safety relief valves lifted for no known reason and blew siding off the side of the Unit 2 Turbine building. This siding hit the feeder lines to the A & C Reserve Station Service Transformers (RSSTs). The operator manually scrammed the plant due to swings in steam generator level and unusual noise coming from the turbine building . Unit 2 shutdown currently de-energized A & C Reserve Station Transformers, which effects D & E transfer buses. This also effects 1J bus, which is de-energized, and 1H & 2J buses which are energized with #1 diesel and #3 diesel. Decay heat removal is being performed thru the SG PORV's and auxiliary feedwater system, with forced cooling from the "B" RCP. Safety related systems are available if required. Notified USDA (A. Jimenez) in addition to the other agencies already identified. The licensee notified the NRC Resident Inspector, as well as State and local agencies.

  • * * UPDATE ON 10/8/2006 AT 05:45 FROM MIKE CHRIS TO ABRAMOVITZ * * *

The site terminated the Alert at 05:40 due to having the "A" RSST in service with bus 1J being powered from its normal power supply. No damage was found from the displaced siding with the exception of the "C" RSST (which should be repaired around noon). The "C" RSST is currently tagged out for maintenance. The licensee notified the NRC Resident Inspector, state, and local governments. Notified: R2DO (Decker), R4DO (Pick), NRR (Dyer, Weber, Quay), IRD (Blount, Wilson), R2 (McCree), DHS (Gray), FEMA (Dunker), DOE (Steve Bailey), EPA (Allison), USDA (Dean Giles), and HHS (Lt. Smith).

  • * * RETRACTION ON 10/16/06 AT 1724 FROM L. WHEELER TO M. ABRAMOVITZ * * *

At 1827 hours on 10/07/06, an Emergency Notification System (ENS) notification was made for an Alert declaration at Surry Power station. The steam discharge from the turbine system safety valves that had lifted caused pieces of siding from the turbine building to dislodge and come in contact with two phases of the overhead bus for the 'A' and 'C' reserve station service transformers (RSST). The basis for the declaration was the Emergency Action Level (EAL) Tab K-11: 'Notification of missile impact causing damage to safety-related equipment or structures'. Upon further review, the RSSTs were determined not to be safely-related equipment. Therefore, the conditions for an Alert emergency did not exist and the notification is being retracted. A notification will be made to the Virginia Department of Emergency Management. This is being reported In accordance with 10 CFR50.72, (b) (2) (xi). As noted in EN# 42890, conditions for a Notification of Unusual Event (NOUE) did exist at the time of the Alert notification. The licensee notified the NRC Resident Inspector. Notified the R2DO (Henson).

ENS 4253026 April 2006 22:39:00KewauneeManual ScramNRC Region 3Westinghouse PWR 2-Loop

At 2043 (CDT) on 4-26-06, during a plant shutdown with the reactor at approximately 35% power, the operating crew manually initiated a reactor trip. The operating crew had just stopped one of the two condensate pumps and then the remaining feedwater pump tripped unexpectedly. The operating crew recognized the turbine did not trip, as it is expected to automatically trip when no feedwater pumps are running. The automatic turbine trip would have automatically tripped the reactor. Therefore, the operating crew manually initiated a reactor trip. Because the reactor did not automatically trip (i.e., failure of RPS to initiate and complete a reactor trip), the Shift Manager declared an Alert, at 2049, based on Chart F of Table 2-1 EPIP-AD-02. Therefore, this is a one-hour notification in accordance with 10CFR50.72(a)(1)(i) 'The declaration of any of the emergency classes specified in the licensee's approved Emergency Plan.' The manual reactor trip is reportable (4-hour) in accordance with 10CFR50.72(b)(2)(iv)(B) 'Any event or condition that results in an 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.' All systems functioned as expected following the manual reactor trip. Service Water Train B is inoperable because of a one-gallon per minute leak. All rods inserted fully. Decay heat is being removed with the steam dump and secondary PORVs. The condenser is losing vacuum due to the turbine trip. Auxiliary Feedwater Pumps started as required. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM JERRY RISTE TO JOHN KNOKE AT 01:45 EDT ON 04/27/06 * * *

At 2049 on April 26, 2006, Kewaunee Power Station staff declared an Alert emergency classification (reference EN# 42530). The Kewaunee Power Station staff has assessed this event. There was no affect on the health and safety of the general public and no release of radiation. No plant personnel were injured and the only plant equipment problem was with the failure of a trip of both feedwater pumps to cause the main turbine to trip. The Kewaunee Power Plant staff has conducted a preliminary investigation of the control room indications and sequential events recorder, which indicates that before the manual reactor trip there was no automatic reactor trip signal present and a failure of the reactor trip breakers did not occur. The Alert was terminated at 0024 CDT (on 04/27/06). The unit is currently in the Hot Shutdown Mode with plans to cool the plant to less than 350 degrees Fahrenheit. The licensee notified the NRC Resident Inspector and will be notifying State and local government and issuing a press release. Notified NRR EO (MJ Ross-Lee), IRD Mgr (P. Wilson), R3DO (H. Peterson), DHS (Holz ), FEMA (Steindurf), NRC/EPA (Crews), DOE (Wyatt), USDA (Timmons), HHS (Peagler).

  • * * UPDATE ON 04/27/06 AT 1718 EDT FROM JERRY RISTE TO ARLON COSTA * * *

When performing a review of the event reported on April 26, 2006 (EN# 42530), the Kewaunee Power Station staff determined another reporting criterion was met. An eight-hour report is required to be made per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of 10 CFR 50.72 except when the actuation results from and is apart of a pre-planned sequence during testing or reactor operation.' 10 CFR 50.72(b)(3)(iv)(B)(6) is PWR auxiliary or emergency feedwater system. As described in EN# 42530, a manual reactor trip of the Kewaunee Power Station was initiated at 2043 on April 26, 2006. The manual reactor trip was initiated when the plant experienced a loss of both feedwater pumps. With a loss of both feedwater pumps and a manual reactor trip, the narrow range water level in both steam generators decreased to the actuation setpoint value for starting the Auxiliary Feedwater Pumps, causing all three Auxiliary Feedwater Pumps to start as designed. Because the steam generator water level was below the actuation setpoint, this was a valid actuation of the auxiliary feedwater system. As a valid actuation of the auxiliary feedwater system, this condition is reportable under 10 CFR 50.72(b)(3)(iv). The untimeliness of the report has been entered into the Kewaunee Power Station's corrective action program. The licensee notified the NRC Resident Inspector. Notified R3DO (H. Peterson).

ENS 424838 April 2006 01:11:00North AnnaManual ScramNRC Region 2Westinghouse PWR 3-LoopWhile in Mode 3 at 547 degrees and 2235 psig in the Reactor Coolant System, during Rod Control System rod drop testing, the group 2, 'A' shutdown bank step counter failed. The step counter is required to be operable in Modes 3, 4 and 5 per (Technical Requirements Manual requirement) 3.1.3 or the reactor trip breakers must be opened within 15 minutes. The step counter failed at 2217 and at 2231 the reactor trip breakers were opened. This was considered a valid actuation of the (Reactor Protection System) due to the (Technical Requirements Manual) requirements and due to the equipment malfunction. Shutdown margin is adequate and all emergency buses are on offsite power. Emergency Diesel Generators are available. The licensee notified the NRC Resident Inspector.
ENS 4236723 February 2006 11:16:00MillstoneManual ScramNRC Region 1CELoss of main feedwater occurred due to an instrument air line failure during a maintenance activity. The reactor was manually tripped and all control rods inserted fully. Auxiliary feedwater received an auto start signal and is providing feedwater to the steam generators. No safety relief valves or PORVs lifted. Decay heat removal is via the turbine bypass valves to the condenser. The plant is in a normal shutdown plant electrical lineup and there was no effect to Unit 3. The licensee notified the State of Connecticut and the city of Waterford. A media press release will be made at a later time. The licensee notified the NRC Resident Inspector.
ENS 421801 December 2005 15:50:00MillstoneAutomatic ScramNRC Region 1Westinghouse PWR 4-LoopWhile reducing power in order to enter containment, and following a manual main turbine trip due to high vibration, an automatic reactor trip on low steam generator level was received as a result of the turbine trip transient. Containment was being entered to investigate the source of an RCS identified leakage, which was less than the Technical Specification limit. In conjunction with the reactor trip, an automatic actuation of the Auxiliary Feedwater System was received as expected. All control rods fully inserted on the automatic trip. The current decay heat removal path is via the steam dumps to the main condenser. No primary or secondary relief valves lifted during the transient. There are no known primary to secondary leaks. All safety related buses are powered from offsite power. With the exception of one diesel out of service for planned maintenance, all emergency diesel generators are available and in standby. Unit 2 was not affected. The licensee notified State and local agencies and the NRC Resident Inspector.
ENS 4217329 November 2005 01:24:00KewauneeAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopAt 22:19 CST, Main Feedwater Pump B tripped on over current. A secondary plant runback from 100% power was automatically initiated. During the secondary plant runback, the reactor automatically tripped on Steam Generator B low-low level at 22:20 CST. All three Auxiliary Feedwater pumps automatically started due to low-low Steam Generator level. The plant has been stabilized at Hot Shutdown (RCS temperature approximately 547 degrees F, RCS pressure approximately 2235 psig). Investigation into the cause of the trip is on-going. This event is being reported under 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system (RPS) when the reactor is critical and 10CFR50.72(b)(3)(iv)(A) for valid actuation of the Auxiliary Feedwater System. All control rods fully inserted on the automatic trip. Steam generator water levels have recovered to indicate in the narrow range. The current decay heat removal path is auxiliary feedwater to the steam generators steaming through the power operated relief valves. There are no known primary to secondary leaks. All safety related buses are powered from offsite power. Emergency diesel generators are available and in standby. The licensee notified the NRC Resident Inspector.
ENS 4202429 September 2005 14:49:00MillstoneManual ScramNRC Region 1Westinghouse PWR 4-LoopThe licensee reported that high wind and wave action at the site has resulted in sea weed and related debris buildup at the Unit 3 intake structure. The traveling screens were unable to keep pace with the debris buildup and two out of six circulating water pumps tripped on high differential pressure across the traveling screens. Based on procedural requirements, the licensee is required to manually trip the plant due to the loss of the two circ water pumps. The reactor trip was characterized as uncomplicated with all systems functioning as required. All rods fully inserted. No primary or secondary relief valves lifted. Auxiliary feedwater automatically started as expected and is supplying cooling water to the steam generators. Decay heat is being discharged to the condenser via the turbine bypass valves. The plant is stable in hot standby at no-load temperature and pressure. The plant trip had no impact on Unit 2. The debris buildup on the Unit 2 intake is being monitored but is less of a problem due to the orientation of the Unit 2 intake structure. The NRC Resident Inspector has been notified by the licensee. The licensee has also notified State and local authorities and has made a press release.
ENS 418986 August 2005 00:20:00North AnnaAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopThe Unit 2 reactor automatically tripped due to an overpower-delta temperature (OpDeltaT) trip signal. The licensee states that an actual OpDeltaT condition did not exist at the time of the trip. The trip signal is believed to have been generated by lightning strikes from an electrical storm that was passing through the area at the time. The trip was uncomplicated and all systems functioned as required. All control rods fully inserted; no safety relief valves lifted; decay heat is being discharged to the main condenser using normal feedwater to supply the steam generators; the reactor temperature and pressure are at normal hot standby range. No obvious grid disturbance was seen during the trip and Unit 1 was not impacted (Unit 1 was being ramped offline at the time for secondary side maintenance work). The licensee noted that Auxiliary Feedwater did auto-start as expected due to a trip from full power and was subsequently secured. The licensee is still investigating the cause of the OpDeltaT trip signal but noted that other Unit 2 instrumentation was found failed after the transient including the Unit 2, A-loop, wide range T-hot and T-cold indications and the Unit 2, B-loop, T-cold indication. The licensee plans to remain in Mode 3 until the investigation is complete and instrument repairs completed. The licensee notified the NRC Resident Inspector.