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 Entered dateSiteRegionScramReactor typeEvent description
ENS 539083 March 2019 00:13:00North AnnaNRC Region 2Manual ScramOn 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 509462 April 2015 06:55:00North AnnaNRC Region 2Manual ScramWestinghouse 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 AnnaNRC Region 2Automatic ScramWestinghouse 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 497842 February 2014 11:01:00North AnnaNRC Region 2Manual ScramWestinghouse 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 4907528 May 2013 18:09:00North AnnaNRC Region 2Manual ScramWestinghouse 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 4902010 May 2013 08:20:00North AnnaNRC Region 2Manual ScramWestinghouse 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 AnnaNRC Region 2Automatic ScramWestinghouse 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 4720126 August 2011 16:23:00North AnnaNRC Region 2Automatic ScramWestinghouse 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 AnnaNRC Region 2Automatic ScramWestinghouse 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 4635222 October 2010 09:46:00North AnnaNRC Region 2Manual ScramWestinghouse 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 AnnaNRC Region 2Automatic ScramWestinghouse 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 4596028 May 2010 03:43:00North AnnaNRC Region 2Automatic ScramWestinghouse 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 455569 December 2009 17:15:00North AnnaNRC Region 2Automatic ScramWestinghouse 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 4386625 December 2007 23:22:00North AnnaNRC Region 2Automatic ScramWestinghouse 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 AnnaNRC Region 2Automatic ScramWestinghouse 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 424838 April 2006 01:11:00North AnnaNRC Region 2Manual ScramWestinghouse 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 418986 August 2005 00:20:00North AnnaNRC Region 2Automatic ScramWestinghouse 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.
ENS 4080410 June 2004 16:11:00North AnnaNRC Region 2Automatic ScramWestinghouse PWR 3-Loop

A Unit 2 automatic reactor trip occurred while the licensee was performing planned periodic testing on train "A" solid state protection. All control rods fully inserted into the reactor core. The Auxiliary Feedwater Pumps automatically started as expected immediately following the reactor trip due to low-low level in the steam generators. The unit is being maintained stable in mode 3 and heat sink is being performed via steam dump to the condensers. All other systems functioned as required. The cause of the reactor trip is under investigation. The licensee notified the NRC Resident Inspector.

      • UPDATE ON 6/11/04 AT 12:23 EDT FROM B. BROWN TO A. COSTA * * *

This is an update to event notification 40804. At 1313 hours on June 10, 2004, North Anna Unit 2 experienced an automatic trip from 100 percent during the performance of 2-PT-36.1A (Train 'A' Reactor Protection and ESF Logic Actuation Logic Test). The cause of the reactor trip, was determined to be an incorrect configuration of the cell switch (52h contract) on 'A' Reactor Bypass Breaker, 2-EP-BKR-BYA. The incorrect cell switch configuration resulted in a turbine trip signal being generated during testing which resulted in a reactor trip signal being generated in the 'B' train Reactor Protection System. The Auxiliary Feedwater System actuated in response to the event. Control room personnel responded to the event in accordance with emergency procedure E-0, Reactor Trip or Safety Injection. The control room team stabilized the plant using ES-0.1 Reactor Trip recovery. The lowest Reactor Coolant System (RCS) pressure during the event was 1988 psig and the lowest RCS temperature was 549 degrees. No human performance issues were identified during this event. A non-emergency four-hour report was made to the NRC operations center at 1611 hours pursuant to 10CFR50.72(b)(2)(iv)(B) for an actuation of the Reactor Protection System while critical. An eight-hour report was also made to the NRC in accordance with 10CFR 50.72(b)(3)(iv)(A) due to the Auxiliary Feedwater Pump starts (Engineering Safety Features Actuation). The Reactor Protection System, AMSAC (ATWAS Mitigating System Actuation Circuit), and the Auxiliary Feedwater System operated properly in response to the event. During the Unit 2 reactor trip, a blown output fuse on a logic card (that feeds the permissive for arming the Steam Dumps from loss of load) prevented the Main Steam Dump Valves from opening in Tavg Mode as expected. The Steam Generator Power Operated Relief Valves (PORVs) lifted and operated to control RCS temperature until transferring Steam Dump control to the Steam Pressure Mode. The fuse was replaced. A post trip review was conducted at 1500 hours on June 10, 2004. The cell switches on the Reactor Trip Bypass breakers have been repaired and post maintenance testing has been completed. Management approval was granted to start-up Unit 2. North Anna Unit 2 is currently in Mode 1 and is preparing to be placed on-line. The licensee notified the NRC Resident Inspector. Notified R2DO (Lesser) and NRR EO (Bateman).