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The query [[Category:ENS Notification]] [[Site.Company::Duke Energy]] [[Scram::+]] was answered by the SMWSQLStore3 in 2.4398 seconds.


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 Entered dateSiteScramRegionReactor typeEvent description
ENS 5421211 August 2019 12:14:00RobinsonAutomatic ScramNRC Region 2At 0840 EDT, on August 11, 2019, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip. The trip was not complex, with all systems responding normally post trip. Because of the reactor trip, the Auxiliary Feedwater (AFW) System actuated as expected due to low water levels in the steam generators. The AFW pumps started as designed when the valid system actuation was received. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a 4-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). This event is also being reported as an 8-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS and AFW. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The site remains in a normal electrical lineup.
ENS 540473 May 2019 19:00:00McGuireAutomatic ScramNRC Region 2At 1554 EDT on 5/3/19, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped on Over Temperature Delta Temperature following a pressure transient in the Reactor Coolant System. The trip was uncomplicated with all systems responding normally post trip. Operations manually started the motor driven auxiliary feedwater pumps and has stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected. Due to Reactor Protection System actuation while critical and actuation of the motor driven auxiliary feedwater pumps, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Unit 1 is in a normal electrical lineup. Prior to the automatic trip, the backup pressurizer heaters were in service as is normal during power ascension. The pressure transient started when the backup heaters were in the process of being removed from service. The licensee notified the NRC Resident Inspector.
ENS 5401622 April 2019 01:51:00BrunswickAutomatic ScramNRC Region 2

At 2307 EDT on April 21, 2019, in Mode 1 at approximately 100 percent reactor power, Unit 1 automatically tripped due to a Main Turbine Trip. The Main Turbine Trip was a result of two out of three level instruments sensing a false high reactor water level. All control rods inserted as expected during the scram. Safety Relief Valves G and K lifted per design. The same level instruments that failed also tripped both Reactor Feed Pumps. As a result, reactor water level dropped below the Low Level 1 and 2 actuation setpoints. Per design, the Low Level 1 signal resulted in Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The Low Level 2 signals resulted in Group 3 (i.e. Reactor Water Cleanup) isolation, a secondary containment isolation signal, and an auto start of Standby Gas Treatment and Control Room Emergency Ventilation. Also, the Low Level 2 resulted in (high pressure coolant injection) HPCI and (reactor core isolation cooling system) RCIC automatically starting and injecting into the vessel. All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Decay heat is currently being removed via the turbine bypass valves. Condensate and feed water are maintaining water level. The reactor is still at saturation temperature and 475 psi, lowering slowly. The reactor is still in a normal electrical lineup. There was no impact to Unit 2 as a result of this event.

  • * * UPDATE ON 04/22/19 AT 0220 EDT FROM ALAN SCHULTZ TO JEFFREY WHITED * * *

The licensee updated the event report to include a 4-Hr Non-Emergency Notification in accordance with 10 CFR 50.72(b)(2)(iv)(A) for Emergency Core Cooling System, HPCI, Discharge to the Reactor Coolant System. Notified R2DO (Dickson), NRR EO (Miller) and IR MOC (Gott).

ENS 5396630 March 2019 21:06:00BrunswickManual ScramNRC Region 2At 17:47 Eastern Daylight Time (EDT) on March 30, 2019, with Unit 2 in Mode 1 at approximately 23 percent reactor power and main turbine startup in progress coming out of a refuel outage, a high temperature was sensed at main turbine bearing #9. As a result of and to arrest the high temperature condition, the main control room inserted a manual reactor scram. All control rods inserted as expected during the scram. When the scram was inserted, reactor water level dropped below the Low Level 1 actuation setpoint. Per design, the Low Level 1 signal resulted in Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The main control room manually closed all Main Steam Isolation Valves (MSIVs), in anticipation of a low vacuum prior to the Group 1 automatic closure signal being received. High Pressure Coolant Injection (HPCI) was aligned for pressure control and Reactor Coolant Isolation System (RCIC) was aligned for level control. The Reactor Coolant Sample Line Isolation valves closed as expected on low main condenser vacuum. All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. At the time of notification, decay heat was being removed by the condenser through one open MSIV and a feedwater pump running.
ENS 5367720 October 2018 01:13:00OconeeAutomatic Scram
Manual Scram
NRC Region 2On 10/19/18 at 2202 EDT, at 19 (percent) Reactor power, a malfunction of (the) Turbine Steam Seal Header pressure control caused a loss of Condenser vacuum, resulting in an automatic trip of the Main Turbine and a manual reactor trip (RPS Actuation). Just prior to the reactor trip, Emergency Feedwater was manually initiated to mitigate the potential loss of Main Feedwater. Condenser vacuum was recovered after the reactor trip and Main Feedwater remained in operation. Due to the RPS actuation while critical, this event is being reported as a 4-hour non-emergency per 10CFR50.72(b)(2). Also, due to the manual initiation of Emergency Feedwater, this event is also being reported as an 8-hour non-emergency per 10CFR50.72(b)(3). Following the reactor trip, all systems responded as expected with no complications. Emergency feedwater was secured at 2300. Unit 1 is in Mode 3 and stable, continuing to cooldown for a refueling outage. The NRC Resident Inspector has been notified.
ENS 5332913 April 2018 06:07:00OconeeManual ScramNRC Region 2B&W-L-LPOn 4/13/2018 at 0227 (EDT), the Oconee Unit 1 Reactor was manually tripped from 24 percent power due to the inability to control main feedwater flow through the Main Feedwater Control Valves using the Integrated Control System. Due to the RPS actuation while critical, this event is being reported as a 4-hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B). Following the reactor trip, multiple Main Steam Relief Valves failed to reseat at the expected pressure. Using procedure guidance, Main Steam Pressure was lowered by 115 psig, resulting in the closing of all Main Steam Relief Valves. All other post-trip conditions are normal and all other systems performed as expected. Unit 1 is currently in Mode 3 and stable. Decay heat is being removed by the steam generators discharging steam to the main condenser using the turbine bypass valves. Units 2 and 3 are not affected by the Unit 1 reactor trip. The licensee notified the NRC Resident Inspector.
ENS 533197 April 2018 12:10:00BrunswickAutomatic ScramNRC Region 2GE-4

On April 7, 2018, at 0836 EDT, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped during testing of the stator cooling system. The trip was uncomplicated with all systems responding normally. No safety-related equipment was inoperable at the time of the event. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B).

Operations responded using Emergency Operating Procedures and stabilized the plant in Mode 3. Reactor water level being maintained via normal feedwater system. Decay heat is being removed through the bypass valves.

Reactor water level reached low level 1 (LL1) as a result of the reactor trip. The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the Primary Containment Isolation System (PCIS) actuation, this event is also being reported as an eight-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PCIS. Unit 2 was not affected. There was no impact on the health and safety of the public or plant personnel. The safety significance of this event is minimal. The automatic reactor trip was not complicated and all safety-related systems operated as designed. Investigation of the cause of the Reactor Protection System actuation is in progress. The licensee notified the NRC Resident Inspector.

ENS 5321716 February 2018 13:58:00McGuireAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 1014 (EST) hours on 2/16/18, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped when the Reactor Trip Breakers opened during Train B Solid State Protection System (SSPS) testing. The trip was uncomplicated with all systems responding normally post-trip. Operations manually started the motor driven auxiliary feedwater pumps. The turbine driven auxiliary feedwater pump (TDCAP) auto-started on low steam generator level. A Feedwater Isolation occurred as designed due to the Reactor Trip and Lo Tave condition. Operations stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected. Due to the Reactor Protection System actuation while critical, actuation of the TDCAP and motor driven auxiliary feedwater pumps along with the Feedwater Isolation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5287024 July 2017 19:25:00OconeeAutomatic ScramNRC Region 2B&W-L-LPAt 1638 (EDT) on 7/24/2017, Oconee Unit 3 experienced an automatic reactor trip due to a load rejection when the generator output breakers both tripped open unexpectedly while 525kV switchyard maintenance was being performed. The trip was uncomplicated with all systems responding normally post-trip. Due to the RPS (Reactor Protection System) actuation while critical, this event is being reported as a 4-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The plant responded normally to the reactor trip, and there was no impact on the health and safety of the public or plant personnel. Operations responded using the Emergency Operating Procedure and stabilized Unit 3 in MODE 3. The NRC Senior Resident Inspector has been notified. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 and 2 are not affected.
ENS 522908 October 2016 13:44:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

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

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

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

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

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

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

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

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

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

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

ENS 517157 February 2016 13:46:00BrunswickManual ScramNRC Region 2GE-4

At 1346 EST the licensee reported that at 1326, Brunswick Unit 1 declared an Alert under EAL HA 2.1 due to an explosion/fire in the Balance of Plant 4 kV switchgear bus area. Prior to the Alert declaration, the operators initiated a manual SCRAM due to an unexpected power decrease from 88% to 40%. The licensee has visually verified that there is no ongoing fire and is investigating the initial cause of the event. Offsite power is available to the site, but EDGs 1 and 2 are running and supplying Unit 1 loads. The MSIVs shut and HPCI/RCIC are being used to maintain vessel level. At 1412 EST, NRC decided to remain in Normal Mode. At 1704 EST the licensee reported the following: At 1313 hours Eastern Standard Time (EST) a manual reactor scram was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. At this time, a Startup Auxiliary Transformer (SAT) experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. Emergency Diesel Generators (EDGs) 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The power interruption resulted in closure of the Main Steam Isolation Valves, per design. The manual scram also resulted in closure of Group 2, 6, and 6 Containment Isolation Valves. The Reactor Core Isolation Cooling (RCIC) system was manually started and is being used to control reactor water level. The High Pressure Coolant Injection (HPCI) system was manually started and is being used for pressure control. The Plant response to the event was per design. Unit 2 is not directly affected by the event, however, due to the shared electrical distribution system is in a Technical Specification Action statement due to the Inoperable Unit 1 SAT. The public health and safety is not impacted by this event. At 1751 EST, the licensee reported that the emergency declaration had been downgraded to an Unusual Event at 1730 because the plant no longer meets the criteria for an Alert, but does meet the criteria for an Unusual Event due to a "loss of all offsite power to Emergency 4 kV buses E1 (E3) and E2 (E4) for greater than or equal to 15 minutes." The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).

  • * * UPDATE FROM MARTY IRWIN TO DANIEL MILLS AT 1825 ON 2/07/16 * * *

At 1814 EST the emergency declaration was terminated because offsite power was restored. The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified R2DO (Musser), NRR EO (Morris), IRD MOC (Stapleton), R2RA (Haney), NRR ET (Lubinski), NRR ET (Dean), DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).

ENS 5078131 January 2015 18:00:00OconeeAutomatic Scram
Manual Scram
NRC Region 2B&W-L-LPAt 1431 EST on 1/31/15, Oconee Unit 3 was manually tripped due to oscillations in the feedwater system in anticipation of an automatic trip. At 1427 EST, Unit 3 began experiencing small feedwater oscillations. Specifically, 3FDW-32, the 3A main feedwater control valve, appeared to be oscillating with corresponding feedwater flow oscillation. Feedwater oscillations continued to grow in magnitude and at 1431 EST a manual trip was directed by the Unit 3 control room supervisor. The shutdown was orderly and the unit is currently stable and in Mode 3 (Hot Standby). Units 1 and 2 were unaffected by the trip and are currently 100% power (Mode 1). Due to the RPS actuation while critical, this event is being reported as a 4-hour non-emergency per 10CFR50.72(b)(2)(iv)(B). Following the reactor trip, a main steam relief valve (MSRV) failed to reseat as expected. Emergency Operating procedure guidance was utilized to reduce main steam system pressure by approximately 80 psig to reseat the valve (valve reseated at 1506 EST). All of the main steam relief valves are now seated. In addition, the 3B condensate booster pump experienced a mechanical seal leak (approximately 4-5 gpm). The pump was subsequently secured at 1447 EST. All other post trip conditions were normal and all other systems performed as expected. Unit 3 is currently in Mode 3 and stable. All rods fully inserted. Main Feedwater is feeding the steam generators and decay heat is being removed to the Main Condenser. The cause of the trip is under investigation. There is no known primary to secondary leakage. The NRC Resident Inspector has been informed.
ENS 4974218 January 2014 10:47:00HarrisManual ScramNRC Region 2Westinghouse PWR 3-Loop

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

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

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

ENS 4970810 January 2014 00:27:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 2234 hours EST on 01/09/2014, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. At the time of the event, Steam Generator Water level Protection Channel Testing was in progress. While testing was in progress with the 'C' Steam Generator Channel 1 Water Level Protection channel in trip for testing, a Turbine Trip occurred. The cause of the Turbine Trip is under investigation. The (Turbine Driven and Motor Driven) Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. The pressurizer Power Operated Relief Valves (PORVs) and the Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. Currently, the unit is stable in Hot Standby (Mode 3). This event is being reported as a 4 hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B) due to the valid Reactor Protection System Actuation. This event is being reported as an 8 hour Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A) due to the valid actuation of Auxiliary Feedwater System. At no time during this occurrence was the public or plant staff at risk as a result of this event. The NRC Resident Inspector has been notified. State and local authorities will be notified. Estimated restart date is 1/12/2014
ENS 4953814 November 2013 16:03:00McGuireManual ScramNRC Region 2Westinghouse PWR 4-Loop

On November 14, 2013, at 1313 Eastern Standard Time, Unit 1 was manually tripped from 100% power due to indications of (four) dropped control rods. This manual reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The cause of the dropped rods is not confirmed at this time, but may be related to maintenance in a Rod Control Power Cabinet ongoing at the time of the event. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip and all plant systems operated as designed. The Auxiliary Feedwater (AFW) system (1A and 1B motor-driven pumps) was manually started for steam generator level control following reactor trip. The start of the AFW system is reportable per 10 CFR 50.72 (b)(3)(iv)(A) for a valid system actuation. Decay heat is being removed via the steam generators (via steam dumps to the main condenser). This event does not impact public health and safety. Unit 2 was not affected by this event. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM WARREN MOORE TO DANIEL MILLS ON 11/18/13 AT 0950 EST * * *
A subsequent licensee evaluation determined that there were ten dropped control rods. 

Notified the R1DO (Desai).

ENS 495065 November 2013 21:31:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1800 hours EST on 11/05/2013, with the unit in Mode 1 at 19% power, an automatic reactor trip occurred. Operators were transferring loads from the Startup Transformer to the Unit Auxiliary Transformer in accordance with normal operating procedures. When breaker 52/7, Unit Aux to 4KV Bus 1 Breaker, was taken to the close position, indication on the Reactor Turbine Generator Board (RTGB) went from 'Open' to 'No' indication. Breaker 52/12, Incoming Line Startup Transformer No. 2, cycled open and then re-closed. This resulted in a momentary loss of power to 4KV Bus 2 and 4KV Bus 1. The reactor trip signal was based on a loss of 4KV bus voltage to 2 of the 3 required 4KV buses. The cause of the loss of 480V bus E-1 was a result of loss of power to 4KV Bus 2. As a result of the loss of 480V bus E-1, the 'A' Emergency Diesel Generator (EDG) auto started. The required loads sequenced onto the 'A' EDG with the exception of the 'A' Service Water (SW) pump. The cause of the failure of the 'A' SW pump is under investigation. The one running Main Feedwater Pump ('A' Pump) tripped on the resulting under voltage of 4 kV Bus 1. By design, this condition resulted in an automatic start of Auxiliary Feedwater due to both Main Feed pump breakers being opened. Both 'A' and 'B' Motor-Driven Auxiliary Feedwater (AFW) Pumps started as designed. Steam generator water levels were maintained in the normal operating band. Currently, the unit is stable in Hot Standby (Mode 3). This event is being reported as a 4 hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B) due to the valid Reactor Protection System Actuation This event is being reported as an 8 hour Non-Emergency per 10 CFR 50.72(b)(3)(iv)(A) due to the valid actuation of AFW and EDG auto-start and subsequent starting of required under voltage loads. At no time during this occurrence was the public or plant staff at risk as a result of this event. The (NRC ) Resident Inspector has been notified. The reactor trip was uncomplicated and the is plant is stable in mode 3 with decay heat being released to the main condenser. Normal offsite power is available with the exception of the 480V bus E-1 being supplied by the "A" Emergency Diesel Generator.
ENS 4947124 October 2013 09:45:00OconeeManual Scram
Automatic Scram
NRC Region 2B&W-L-LP

At 0553 EDT on 10/24/2013, Oconee Unit 3 was manually tripped due to oscillations in the feedwater system in anticipation of an automatic reactor trip. At 0549 EDT, Unit 3 began experiencing small feedwater oscillations. The feedwater control system was placed in manual in an attempt to stabilize feedwater flows. Feedwater oscillations continued to grow in magnitude and at 0553 EDT, a manual trip was directed to prevent an automatic reactor trip. Due to an RPS actuation, this event is being reported as a 4 and 8 hour Non-Emergency per 10 CFR 50.72 (b)(2)(iv)(B) and 10 CFR 50.72 (b)(3) Following the reactor trip, four main steam relief valves failed to reseat. Procedure guidance was utilized to reduce main steam system pressure by approximately 30 psig to reseat the main steam relief valves. All main stream relief valves are now reseated. All other post trip conditions were normal and all other systems performed as expected. Unit 3 is currently in Mode 3 and stable. Operations have been stabilized on Unit 3. A post-trip investigation is in progress, per site procedures and directives. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM BOB MEIXELL TO DONALD NORWOOD AT 1439 ON 9/10/14 * * *

Duke Energy reviewed NRC Event Number 49471 against NUREG 1022, Rev 3, section 3.2.6, "System Actuation" and determined this event should have been reported only per 10 CFR 50.72 (b)(2)(iv)(B), RPS Actuation (while critical). Thus, Duke Energy is revising NRC Event Number 49471 to remove the 8-hour report criteria 10 CFR 50.72 (b)(2)(iv)(A). The NRC Resident Inspector was notified of this revised report. This update has no effect on safety significance. Notified R2DO (Shaeffer).

ENS 4877521 February 2013 12:49:00McGuireAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 0957 EST on 02/21/13, Unit 1 Reactor automatically tripped from 100% power, due to a turbine trip. The turbine trip was caused by a loss of both main feedwater (MFW) pumps. The 1A motor driven auxiliary feedwater (AFW) pump auto started to feed the "A" and "B" Steam Generators (S/G). The 1B motor driven AFW pump was unavailable due to planned maintenance, so the turbine driven AFW pump was manually started to feed the "C" and "D" S/Gs. The reactor trip was uncomplicated. All control rods fully inserted. Decay heat removal is to the main condenser via the turbine bypass valves. There was no primary to secondary leakage. Electrical buses are being supplied via offsite power. Steam generator levels are being returned to normal and MFW has been reset and is available. All other plant systems functioned as designed during and after the reactor trip. There is no impact on Unit 2. There is no impact on the health and safety of the public. The loss of the MFW pumps is still under investigation. The licensee has informed the NRC Resident Inspector.
ENS 4778128 March 2012 17:14:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1503 hours EDT on March 28, 2012, with the unit in Mode 1 at 55% power, an automatic reactor trip occurred. The reactor trip was the result of a turbine trip from a 'B' Steam Generator Hi Level. All Reactor Coolant Pumps (RCP) remained running. The Auxiliary Feeedwater (AFW) System automatically actuated due to both main feedwater pump breakers opening from a valid feedwater isolation signal. Steam Generator Levels were then controlled by Auxiliary Feedwater pumps. Steam Generator Blowdown was automatically isolated with the AFW actuation. The RCS Code Safety valves, Pressurizer Power Operated Relief Valves (PORVs), Steam Generator PORVs or the Main Steam Safety valves (MSSVs) did not open during the event. All control rods fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently in Mode 3 and stable. There were no radiological consequences or releases as a result of this event. The cause of the Steam Generator Hi Level is under investigation. The Resident NRC Inspector has been informed.
ENS 4769023 February 2012 02:55:00BrunswickManual ScramNRC Region 2GE-4

At 2319 hours EST, a manual Reactor Protection System (RPS) actuation was inserted on Unit 1 in anticipation of a loss of condenser vacuum. Shortly before the manual RPS actuation, Circulating Water Intake Pump (CWIP) 1B tripped due to high delta-pressure across the intake traveling screen. This caused the trip of the remaining pumps. Previously, at 1859 hours, balance of plant (BOP) bus Common C unexpectedly de-energized. This caused loss of power to the CWIP traveling screen motors which, in turn, lead to the high delta-pressure across the traveling screen(s). All control rods inserted properly. As a result of the scram, reactor water level reached the Low Level 1 actuation set point and Primary Containment (i.e., Group 6) isolation occurred. All systems functioned as designed. The High Pressure Coolant Injection (HPCI) system is being used, as needed, for pressure control. The Reactor Core Isolation Cooling (RCIC) system is being used, as needed, for level control. No Safety/Relief Valves (SRVs) actuated as a result of the manual RPS actuation. The manual RPS actuation is reportable in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). The actuation of the HPCI and RCIC systems and the Group 6 isolation are reportable in accordance with 10CFR50.72(b)(3)(iv)(A). The unit is currently in Mode 3 with a cooldown in progress. The licensee notified the NRC Resident Inspector. Notified R2DO (Ernstes).

  • * * UPDATE FROM STEWART BYRD TO CHARLES TEAL AT 0741 EST ON 2/23/12 * * *

At 2319 hours EST, a loss of all Circulating Water Intake Pumps caused a lowering vacuum on Unit 1. As previously reported (i.e. Event Notification 47690), a manual Reactor Protection System (RPS) actuation was inserted on Unit 1 at this time. In addition, a valid actuation of the RPS, High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), and a Group 6 isolation was reported in accordance with 10CFR50.72(b)(3)(iv)(A). At 2342, Main Condenser vacuum was 15 in. Hg and lowering. All Main Steam Isolation Valves were slow closed in anticipation of Group 1 isolation at this time. This follow-up notification is being made to report the manual actuation of the Group 1 isolation valves in accordance with 10 CFR 50.72(b)(3)(iv)(A). The Group 1 isolation was discussed with the NRC during initial notification of EN 47690, and this follow-up is providing written notification of the MSIV closure. The NRC Resident Inspector has been notified. Notified R2DO (Ernstes).

ENS 4744416 November 2011 03:33:00BrunswickManual ScramNRC Region 2GE-4

On 11/16/11 at 0208 EST, Brunswick Nuclear Plant, Unit 2 calculated a drywall floor drain 42 minute leak rate of 5.88 gpm, following several hours of gradually rising floor drain leakage during a plant startup. Tech Spec 3.4.4 A was entered, requiring floor drain leakage to be restored below 5 gpm within 8 hours. At 0253 EST, a 45 minute leak rate of 10.11 gpm was calculated. At 0301 EST, Unusual Event SU 6.1 was declared for unidentified leakage exceeding 10 gpm, and at 0309 EST, a manual reactor scram was inserted from approximately 7% power (10 CFR 50.72(b)(2)(iv)(B)). Following the scram, the reactor was depressurized at a maximum cooldown rate of 92.5 deg F/hr, and the unidentified leak rate fell less than 10 gpm within 1 hour and less than 5 gpm within 2 hours. Leak rate at 0614 EST on 11/16/11 is 3.82 gpm with reactor pressure at 228 psig. The exact nature of the leak is unknown at this time. The current plan is to continue to depressurize and cool down the reactor to Mode 4, such that a full drywall inspection can commence. At present, Brunswick has not terminated the Unusual Event. Level control is currently being maintained with control rod drives (CRD). The MSIVs were manually closed to control cooldown. The maximum cooldown was observed to be 92.5 F/hour. The plant plans to reopen MISIVs and depressurized to condensate booster pump injection pressure of 350 psig. The plan is to achieve Mode 4 for a leak inspection. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM DAVID FASHCHER TO CHARLES TEAL AT 0550 EST ON 11/16/11 * * *

The leakage rate is currently 3.73 gpm. The decrease is due to lower pressure which is currently at 258 psig. There are no additional changes. The leakage source is not identified at this time. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

  • * * UPDATE FROM DUSTIN LUPTON TO JOHN SHOEMAKER AT 0648 EST ON 11/16/11 * * *

The leakage rate is currently below the T.S. limit due to lower pressure which is currently at 210 psig. There are no additional changes. The plant will remain in an Unusual Event (UE) until further notice. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

  • * * UPDATE FROM DUSTIN LUPTON TO CHARLES TEAL AT 0749 EST ON 11/16/11 * * *

The leakage rate is stable. The leak rate is calculated at 3.04 gpm at 183 psig at 0708 EST. The current reactor pressure is 162 psig. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

  • * * UPDATE FROM DUSTIN LUPTON TO JOHN KNOKE AT 0832 EST ON 11/16/11 * * *

The licensee terminated from their Unusual Event at 0815 EST. The leakage is still unidentified. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

ENS 4729326 September 2011 14:48:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 1145 hours EDT on September 26, 2011, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. All Reactor Coolant Pumps (RCP) remained running. The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. The steam generator and pressurizer Power Operated Relief Valves (PORVs) and the Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3. The cause of the reactor trip is under investigation.
ENS 4655920 January 2011 16:12:00McGuireManual ScramNRC Region 2Westinghouse PWR 4-LoopUnit 1 performed a manual reactor trip (from 28% power) due to the trip of the 1B main feedwater pump trip concurrent with the 1A main feedwater pump having been previously tripped per the shutdown procedure. Reactor power was below the setpoint for an automatic reactor trip from a turbine trip. All rods fully inserted, heat removal is from auxiliary feedwater and condenser steam dumps, and the electrical system is in a normal alignment. The plant has stabilized at normal RCS pressure and temperature. Both Units 1 and 2 are being shutdown as per Tech Spec 3.0.3 associated with an inoperable Nuclear Service Water System. Unit 2 is at 40% and decreasing power due to the Tech Spec action and was not effected by the Unit 1 reactor trip. Both units will be placed in Mode 5. The licensee has notified the NRC Resident Inspector.
ENS 464678 December 2010 16:51:00HarrisManual ScramNRC Region 2Westinghouse PWR 3-LoopText provided by the licensee. Quotations omitted for readability. Report Type: This 60-day telephone notification is being made under 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1). Description: On October 28, 2010, during the performance of MST I0073, "Train 'B' 18 Month Manual Reactor Trip, Solid State Protection System Actuation Logic & Master Relay Test", two sequential errors resulted in the inappropriate activation of the 'B' ESW pump. During night shift on October 27/28th, the Master Relay Selector Switch was not returned to the required OFF position. This caused day shift to find one of the two general warning lights to be lit. During the troubleshooting for the light, a technician discovered the Master Relay Selector Switch out of the expected OFF position as required by the procedure. A technician moved the switch to the OFF position outside of procedural guidance, resulting in the partial activation of the Reactor Protection System, including the 'B' ESW pump. The plant was in Mode 6 due to refueling outage 16 during the event. Actual plant conditions and parameters did not exist that required an automatic start of the 'B' ESW Pump. Therefore, this actuation is classified as invalid. The system started and functioned successfully. This invalid actuation was entered into the corrective action program as NCR 430289. Cause: Poor performance of task by the individual. The licensee informed the NRC Resident Inspector.
ENS 463137 October 2010 04:06:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 0013 hours EDT on October 7, 2010, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the One Loop Low Flow reactor protection function. Reactor Coolant Pump (RCP) 'C' tripped during the event. The Auxiliary Feedwater System automatically actuated, as expected due to low steam generator water level, and provided feedwater to the steam generators. Steam generator and pressurizer Power Operated Relief Valves (PORVs) and Main Steam Safety Valves did not open during the event. All control rods indicated fully inserted following the reactor trip. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3. The Main Turbine Lube Oil Deluge System actuated without a fire during the event. A fire hose station pipe ruptured after the deluge system actuation. The fire hose station was isolated. In addition, the 'B' Main Feedwater (MFW) Pump tripped during the event and the 'A' Main Feedwater pump subsequently tripped due to high steam generator level in the 'C' Steam Generator at about 0023 hours. There is currently indication of RCP 'C' Number 2 seal leakage of approximately 2.5 gpm. At approximately 0405 EDT the Auxiliary Feedwater (AFW) system actuated due to a trip of MFW Pump 'A' while attempting to start the pump in accordance with procedure GP-004, Post Trip Stabilization. The AFW system actuation signal caused motor-driven AFW Pump 'B' to start. The motor-driven AFW Pump 'A' was already in operation due to the post-trip condition. The cause of the MFW Pump 'A' trip is under investigation. This AFW system actuation is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). The cause of the reactor trip is under investigation. The licensee notified the NRC Resident Inspector.
ENS 462389 September 2010 18:04:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

REACTOR TRIP DUE TO OVERTEMPERATURE DELTA-T SIGNAL

At 1437 hours EDT on September 9, 2010, with the unit in Mode 1 at 100% power, an automatic reactor trip occurred. The reactor trip signal was based on the Overtemperature Delta-T reactor protection function.

During the event, the steam generator power operated relief valves (PORVs) and one pressurizer PORV briefly opened and re-closed, in response to pressure changes in the steam generators and pressurizer due to the plant transient condition. The Auxiliary Feedwater System automatically actuated, as expected, and provided feedwater to the steam generators. The main steam safety valves did not open during the event. All control rods indicated fully inserted following the reactor trip.

The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable. Decay heat is being removed via the condenser. The reactor is currently stable in Mode 3.

There was an indication of an approximate 0.65 gpm leak to the pressurizer relief tank following the reactor trip. The isolation valve to the pressurizer PORV that opened during the reactor trip was closed and the leak indication stopped. The indicated leakage was within Technical Specification leakage rate limits.

The cause of the reactor trip and indication of pressurizer PORV leakage is under investigation.

The licensee notified the NRC Resident Inspector.

ENS 461597 August 2010 18:47:00OconeeManual ScramNRC Region 2B&W-L-LPAt 1451 on 8/7/2010, Oconee Unit 1 initiated a manual reactor trip from approximately 17% power due to indicated vibrations on 1A1 and 1A2 reactor coolant pumps (RCP) reaching the high vibration trip criteria. All systems responded normally following the reactor trip. Unit 1 is currently stable in MODE 3. An investigation is in progress to determine the cause of the elevated reactor coolant pump vibrations. All control rods fully inserted on the trip. Decay heat is being removed via the turbine bypass valves to the main condenser. Steam generator water level is being maintained with main feedwater. There is no evidence that the PORVs or safety valves lifted. The plant is in its normal shutdown electrical lineup. There were no indications on the loose parts monitor except for the RCP high vibration. There was no affect on units 2 or 3. The licensee notified the NRC Resident Inspector.
ENS 4600312 June 2010 09:14:00McGuireManual ScramNRC Region 2Westinghouse PWR 4-LoopMcGuire Unit 1 was operating at 44% power due to a previously dropped control rod. Indication was received of a second control rod drop and the reactor was manually tripped in accordance with abnormal operating procedure guidance. All control rods fully inserted. The auxiliary feedwater system was manually started due to an approaching autostart setpoint. The unit is stable in Mode 3 at normal operating temperature and pressure. Normal containment air release remains in progress. The steam generators are being fed through the auxiliary feedwater system and will transition to the normal feedwater system. Decay heat removal is to the condenser through the steam dumps. There was no impact on Unit 2. The licensee is investigating the cause of the control rod drop indication. The licensee notified the NRC Resident Inspector.
ENS 459025 May 2010 15:11:00BrunswickAutomatic ScramNRC Region 2GE-4On May 5, 2010, at 1144 hours Eastern Daylight Time (EDT), an automatic reactor scram occurred on Unit 1 following a trip of the 1B Reactor Feed Pump (RFP). Following the 1B RFP trip, the reactor recirculation pumps did not run back as expected. The resulting water level shrink caused level in the Reactor Pressure Vessel (RPV) to drop to Low Level 1, causing the activation of the Reactor Protection System (RPS) and the Primary Containment Isolation System (PCIS). All control rods properly inserted. PCIS Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 8 (i.e., RHR Shutdown Cooling) isolation signals were received on Low Level 1. Actuations of the Primary Containment Isolation Valves (PCIVs) were completed and the affected equipment responded as designed. Due to the expected RPV level reduction following a reactor scram, water level in the RPV momentarily reached Low Level 2. This initiated the High Pressure Coolant Injection (HPCI) System, the Reactor Core Isolation Cooling (RCIC) System, and a partial Group 3 PCIS (i.e., RWCU) isolation. The HPCI and RCIC systems did not inject. The 1-G31-F001 isolated (i.e., inboard isolation) but 1-G31-F004 (i.e., outboard isolation) did not automatically isolate. Based on a preliminary assessment, this response appears to be in accordance with plant design. Further assessments of plant response are on-going to validate plant response. The licensee has notified the NRC Resident Inspector. The scram was uncomplicated. No SRVs lifted. Decay heat removal is via the 'A' feed water pump via the turbine bypass valves to the condenser. The electrical line-up of Unit 1 is normal. Brunswick Unit 2 was not affected.
ENS 4579928 March 2010 22:47:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-Loop

At approximately 1852 hours Eastern Daylight Time (EDT), on March 28, 2010, with the unit operating at approximately 99.5% power, the H. B. Robinson Steam Electric Plant, Unit No. 2, reactor protection system actuated resulting in an automatic trip of the reactor. At about the time of the reactor trip, there was indication of a loss of the power to the Train 'B' emergency bus. The Train 'B' Emergency Diesel Generator started and provided power to the Train 'B' emergency bus. The Reactor Coolant System pressure response after the reactor trip resulted in the actuation of the Safety Injection System. The reduction in Reactor Coolant System pressure allowed the safety injection system to provide flow to the reactor coolant system, although, there was no indication of conditions that would require the safety injection system to provide flow to the reactor coolant system. Specifically, diagnosis of the event determined that a loss of coolant event, steam generator tube rupture, or secondary system break were not occurring. The safety systems that actuated for this event included the Reactor Protection System, the Safety Injection System, both Emergency Diesel Generators started and the 'B' Emergency Diesel Generator provided power to the 'B' Train Emergency Bus, the Auxiliary Feedwater System actuated, and the three main steam isolation valves closed. The Steam Generator Power Operated Relief Valves are being used for decay heat removal because of the closure of the main steam isolation valves. The Startup Transformer is energized and providing power to the 'A' Train Emergency Bus. Investigation of the cause of the reactor trip and associated emergency system actuations is in progress. At the time of the event, it was noted that there was evidence of a fire condition at 4KV bus number 5 located in the Turbine Building. It is currently believed that this is the cause of the transient condition that resulted in the reactor trip and other emergency system actuations. The reactor is currently being maintained in MODE 3, Hot Standby, conditions. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) for ECCS discharge into the RCS and (B) for RPS actuation, and 10 CFR 50.72(b)(3)(iv)(A) for specified system actuation as described. Additionally, based on an inquiry from the local news media reporter, information regarding this event was discussed with the local media. This contact with the local news media is being reported in accordance with 10 CFR 50.72(b)(2)(xi). All control rods fully inserted on the trip. Volume Control Tank (VCT) level was not dropping consistent with no reactor coolant system leakage. Pressurizer level and pressure drop was indicative of shrink associated with the Reactor Coolant Pump stoppage and cooldown in one loop. The licensee observed no indication of reactor coolant system leakage from the containment sump water level monitors. The licensee informed the NRC Resident Inspector.

  • * * UPDATE AT 2340 EDT ON 03/28/10 FROM MIKE DONITHAN TO PETE SNYDER * * *

The original Non-Emergency classification was reclassified as an ALERT based on the following: A fire on 4kV Bus 5 affected 4kV Bus 4 which caused a loss of 'B' Reactor Coolant Pump which caused a Reactor trip and Turbine Trip. That fire was extinguished without a required event declaration, but a subsequent fire on 4kV Bus 4 required declaration of an ALERT at 2300 (EDT) based on the fire affecting the safety-related 'A' and 'B' DC Buses. The fire was out at 2301 (EDT). The primary systems are available and control of the Reactor Coolant System has been maintained. The Safety Injection termination Emergency Operating Procedure has been performed and the general procedure for post-trip stabilization is being performed. The licensee informed state/local agencies and the NRC Resident Inspector. Notified R2DO(Guthrie), IRD(Gott), EO(Blount), ET(Leeds), and R2RA(Reyes).

  • * * UPDATE AT 0134 EDT ON 03/29/10 FROM BRYAN C. WALDSMITH TO CHARLES TEAL * * *

Event Termination Notice for Event 45799. ALERT declaration is no longer required. Event Termination Notice: Current plant Conditions are stable and the conditions that required declaration of the ALERT are no longer present. The fire causing the ALERT classification was extinguished at 2301 (EDT). There was no explosion or steam line break. The last safety-related DC bus ground was identified and cleared at 0001 (EDT). The licensee informed state/local agencies and the NRC Resident Inspector. Notified R2DO(Guthrie), IRD(Gott), EO(Blount), DHS(Moore), FEMA(Casto), DOE(Moorone), USDA(Hovey), and HHS(Nunn).

ENS 4549916 November 2009 02:24:00HarrisManual ScramNRC Region 2Westinghouse PWR 3-LoopAt 2242 (EST) on 11/15/09, the reactor was manually scrammed from 100% power due to a large oil leak on the main generator seal oil system. Condenser vacuum was broken immediately following the reactor trip, and the main turbine stopped rotating at 2324 (EST). Following the reactor trip, the 'B' steam generator Main Steam Isolation Valve (MSIV) failed to fully close on demand, but was closed due to field actions at 2303 (EST). The reactor remained stable at NOP/NOT following the reactor trip. Offsite power remained available throughout the event. This condition is being reported as actuation of the reactor protection system in accordance with 10CFR 50.72(b)(2)(iv)(B). All control rods fully inserted and decay heat is being removed through the S/G relief valves to the atmospheric dumps. No known primary to secondary leakage exists. The plant remains stable in Mode 3. The licensee notified the NRC Resident Inspector.
ENS 454836 November 2009 23:36:00RobinsonManual ScramNRC Region 2Westinghouse PWR 3-LoopA Manual Reactor Trip was initiated due to closure of Main Feedwater Regulating Valve 'A' with Steam Generator 'A' Level at 35% narrow range and lowering with a Steam Flow / Feed Flow mismatch present. Both Motor Driven Auxiliary Feedwater Pumps (MDAFW) and the Steam Driven Auxiliary Feedwater Pump (SDAFW) auto-started as required based on low Steam Generator Water Levels. All systems responded normally and plant operators have stabilized the unit in Mode 3. There were no complications. All rods inserted during the trip. Decay heat is being removed via steam dumps to condensers. The licensee notified the NRC Resident Inspector.
ENS 4528624 August 2009 12:48:00Crystal RiverManual ScramNRC Region 2B&W-L-LPFollowing completion of surveillance testing for the electrical checks of the Control Rod Drive power train, the operating crew observed that Group 7 regulating control rods unexpectedly lost power and inserted into the core. The Operators manually tripped the reactor prior to exceeding any RPS trip set point. There were no other safety system actuations and the plant is stable at normal post-trip temperature and pressure. There are seven rods in Group 7. Operators tripped the unit within 7 seconds. All rods fully inserted on the trip. The plant is removing decay heat through the main condenser and feeding generators with auxiliary feed. The licensee notified the NRC Resident Inspector.
ENS 4480727 January 2009 12:41:00Crystal RiverManual ScramNRC Region 2B&W-L-LPCalibrations of the 'A' Unit 4160V Switchgear metering were in progress when the 'A' Unit 4160V Bus tripped. This resulted in the loss of the 'A' Feedwater Booster Pump (FWBP) and 'A' Condensate Pump (CDP). The Operating crew identified the loss of the 'A' FWBP with increasing RCS pressure and manually tripped the reactor. There were no other safety system actuations and the plant is stable at normal post-trip temperature and pressure. All rods inserted during the trip. Decay heat is being removed via steam dumps to the condenser. The electrical grid is stable with plant loads being supplied by offsite power via the startup transformer. Both vital busses are being powered from offsite. During the transient, main steam relief valves did lift but have been reseated. The NRC Resident Inspector has been notified.
ENS 4468526 November 2008 15:12:00BrunswickAutomatic ScramNRC Region 2GE-4At 12:00 hours EST, an apparent Electro-Hydraulic Control (EHC) system malfunction while synchronizing the Main Generator to the grid resulted in a Group 1 Main Steam Isolation Valve (MSIVs) closure on low reactor pressure and a subsequent automatic reactor scram. Preliminary investigation of the automatic scram signal indicates that Main Steam Line Low Pressure Instruments (B21-PT-N015 A thru D) sensed low steam line pressure after the Main Generator was paralleled to the grid. This resulted in the closure of all MSIVs. Closure of MSIVs in Mode 1 results in an automatic reactor scram. All control rods fully inserted. With the exception of the EHC system, all systems responded as designed. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to the Reactor Protection System (RPS) actuation and 10CFR 50,72(b)(3)(iv)(A) due to the Primary Containment Isolation System (PCIS) Groups 1, 2, and 6 actuations. Unit 2 was not affected by this event and remains at 100% power. Reactor Pressure is 844 psig, reactor temperature is 526 degrees Fahrenheit. MSIV's remain shut. Decay heat was removed via MSIV bypass lines. RCIC actuated for a period of time for level control but has been secured. CRD pumps are providing makeup water. No SRV's or relief valves lifted. The NRC Resident Inspector has been notified.
ENS 4466017 November 2008 09:21:00RobinsonManual ScramNRC Region 2Westinghouse PWR 3-LoopAt 0551 hours EST, on November 17, 2008, the H. B. Robinson Steam Electric Plant, Unit No. 2, reactor was manually tripped from approximately 78% power due to high vibrations detected on the main turbine. The reactor trip was initiated in accordance with Abnormal Operating Procedure, AOP-006, "Turbine Eccentricity / Vibration." At approximately 0230 hours EST, it was noticed that turbine vibration on the No. 9 bearing was at approximately 10.9 mils and increasing. At 0516 hours EST, the No. 9 bearing was at about 13.5 mils and still increasing. A power reduction from 100% power was commenced in accordance with Operating Procedure, OP-105, "Maneuvering the Plant when Greater than 25% Power." At approximately 0551 hours EST, with power level at approximately 78%, the No. 9 bearing vibrations reached the trip criterion of 14 mils and the reactor was tripped in accordance with AOP-006. Path-1, which is the flow chart based on the Westinghouse Owners Group Emergency Response Guideline E-0 for reactor trip response, and Procedures EPP-4, "Reactor Trip Response," and GP-004, "Post Trip Stabilization," were used after initiation of the reactor trip. The auxiliary feedwater system started automatically, as expected, in response to steam generator level changes after the reactor trip. It was also noted that the "B" Main Feedwater Pump had tripped. The cause of the main feedwater pump trip has not been determined and is under investigation. The primary system and steam generator power operated relief valves and safety valves did not actuate during this event. The main feedwater system, main steam system, and condenser remained available during the event and are currently being used for decay heat removal. The normal post-trip electrical configuration is providing power to the required buses and the offsite electrical system is stable at this time. The emergency diesel generators are operable. The unit is currently stable in MODE 3. An estimated restart date has not yet been established. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) for reactor protection system actuation and 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the auxiliary feedwater system. The licensee notified the NRC Resident Inspector.
ENS 446479 November 2008 12:08:00BrunswickManual ScramNRC Region 2GE-4One hour reportable event based on a safety relief valve (SRV) failure to close (NUREG 0626 and NUREG 0660). On 11/09/08 at approximately 11:08 with Unit 2 at 100% steady state power SRV 2-B21-F013H spuriously failed full open with no operator action or testing in progress. The valve's control switch was cycled as required by Abnormal Operating Procedure AOP-30 with no success. At 1113 the valve was successfully closed by pulling the associated fuses. At 1117, a manual reactor scram was inserted based on a Torus temperature of 109.8 degrees F (Technical Specifications require a scram to be inserted at 110 degrees F). All control rods (fully) inserted from the manual scram signal. Reactor water level lowered to Low Level 2 resulting in Primary Containment Isolation System (PCIS) isolations of Groups 2, 3, 6, and 8. In addition, this resulted in a Reactor Core Isolation Cooling (RCIC) system actuation and injection into the reactor. The High Pressure Coolant Injection (HPCI) system actuated but did not inject because reactor water level recovered. An Alternate Rod Insertion signal was received, the Standby Gas Treatment (SBGT) system initiated, and the Reactor Recirculation Pumps tripped as designed. Plant safety systems responded as designed to the transient. (Licensee) investigations are underway to determine the cause of the SRV failure. Reactor decay heat is being removed through the main turbine bypass valves to the condenser. Reactor make-up is being maintained by the normal feedwater system. The plant is in its normal shutdown electrical lineup supplied by offsite power. The diesel generators are available for service to the plant The Licensee notified the NRC Resident Inspector.
ENS 446387 November 2008 12:05:00OconeeAutomatic ScramNRC Region 2B&W-L-LPAt 0834 on 11-7-08, Oconee Nuclear Station (ONS) Unit 3 experienced an automatic reactor trip due to a reactor protective systems (RPS) actuation. ONS Unit 3 post trip parameters are normal. Main Feedwater remained in service following the event and Emergency Feedwater was not required and remained available. Electrical power automatically transferred to the Startup power source from the switchyard and emergency AC power sources were not required and remained available. The cause of the RPS actuation and automatic reactor trip are under investigation. ONS Unit 1 remained at 100% power with no issues following the ONS Unit 3 reactor trip. ONS Unit 2 remained in No-Mode. No other safety systems have actuated or exhibited abnormal behavior. Therefore, the safety significance of this condition is low. All control rods fully inserted during this event. The unit is removing decay heat to the main condenser. No primary relief valves lifted and Main Steam relief valves cycled following the trip. There are no Steam Generator tube leaks. The licensee notified the NRC Resident Inspector.
ENS 446243 November 2008 04:35:00McGuireManual ScramNRC Region 2Westinghouse PWR 4-Loop

Manually opened reactor trip breakers in Mode 5 to insert control bank 'B' due to blown fuse in rod control cabinet. Licensee was moving control rod bank "B" following I & C work, and the control rod bank failed to move as expected. All other control rod banks were inserted into the core at the time of the event. EDG's and offsite power sources are OPERABLE, and there is no increase in plant risk. The licensee will inform the NRC Resident Inspector.

* * * RETRACTION ON 12/31/08 AT 1326 FROM RICK ABBOTT TO PETE SNYDER * * * 

Regarding the NRC Event Number 44624 conveyed November 3, 2008, McGuire Nuclear Station has determined that manually opening the reactor trip breakers was not reportable and hereby retracts this notification. Upon further consideration it was determined that manually opening the reactor trip breakers was a conservative decision to fully insert control rods based on the failure mechanism causing a single rod to drop to the fully inserted position. Manual actuation of the reactor trip breakers was not required by abnormal procedures and was performed only after consultation between operations, engineering and senior station management agreed that this was the preferred option. Therefore, the decision to manually open the reactor trip breakers is considered to be a preplanned actuation of the reactor protection system and is not reportable. The licensee informed the NRC Resident Inspector. Notified R2DO (M. Lesser).

ENS 4461831 October 2008 13:45:00McGuireManual ScramNRC Region 2Westinghouse PWR 4-LoopInitial conditions: Low Power Physics Testing after outage. Performing Dynamic Rod Worth measurements. Single dropped rod (K-2) on Control Bank B occurred after movement disagreement between rod groups. Entered AP-14, 'Rod Control Malfunction'. Performed a manual reactor trip and manual start of Auxiliary Feedwater as directed by Normal Operating Procedures for unit shutdown. Feedwater isolation occurred due to Low T-Ave concurrent with reactor trip. All control rods were fully inserted. Licensee has notified NRC Resident Inspector.
ENS 4445330 August 2008 17:47:00BrunswickManual Scram
Automatic Scram
NRC Region 2GE-4At 1503 hours EDT, an Electro-Hydraulic Control (EHC) system malfunction caused the Unit 2 Main Turbine bypass valves (BPV) to start cycling. Initially, BPV 1 partially opened and closed followed shortly thereafter by four BPVs going full open. At that time the order was given to insert a manual scram. An automatic scram signal occurred just as the operator was beginning to insert the manual scram. Preliminary investigation of the automatic scram signal indicates that it was initiated by low Relay Emergency Trip Supply (RETS) pressure to the main turbine control valves due to the EHC malfunction. Reactor water level momentarily dropped below Low Level during the response. This resulted in Primary Containment Isolation System (PCIS) Group 2 and Group 6 isolations, as expected. All control rods fully inserted. All systems responded as designed. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to the Reactor Protection System actuation and 10 CFR 50.72(b)(3)(iv)(A) due to the PCIS Group 2 and Group 6 actuations. Unit 1 was not affected by this event and remains at 100% power. The NRC resident inspector has been notified.
ENS 4443824 August 2008 18:03:00Crystal RiverManual ScramNRC Region 2B&W-L-LPThe 'A' Condensate Pump became uncoupled, lowering Condensate flow. Operators began to manually lower reactor power to maintain deaerator level. Reactor power was lowered to approximately 62 percent. At this power, (feedwater) flow oscillations began and were excessive. With these flow oscillations increasing the decision was made to manually trip the Reactor. The Reactor was manually tripped at 1557 hours. There were no safety system actuations other than RPS (Manual). The plant is stable in a normal post trip configuration. All control rods inserted into the core during the reactor trip. Offsite power is available and powering safety loads. The steam generator safeties lifted during the transient and reseated. There is no known primary to secondary leakage. Decay heat is being removed via steam dumps to the condenser using normal feedwater to the steam generator. The emergency feedwater system was not initiated during the reactor trip. The licensee notified the NRC Resident Inspector.
ENS 4442319 August 2008 15:07:00HarrisManual ScramNRC Region 2Westinghouse PWR 3-LoopOn August 19, 2008, with the Unit shut down in Mode 3, post maintenance testing was being performed for the Digital Rod Position Indication System. While performing this test, a 'Rod Control Urgent Failure' alarm was received upon initial withdrawal of Control Bank-C. All other control and shutdown banks remained fully inserted in the core. Local inspection revealed a phase failure on movable gripper coils in a power cabinet. In accordance with plant procedures, at 0905 hours, a manual reactor trip was initiated by operators opening the reactor trip breakers. All Safety Systems functioned as designed and Rod Control System repairs are in progress. This event posed no significant safety implications because the reactor was subcritical when the reactor trip breakers were opened. Compliance with all Technical Specification requirements was maintained. The health and safety of the public were not affected by this event. The NRC Resident Inspector was notified.
ENS 4440411 August 2008 04:21:00HarrisManual ScramNRC Region 2Westinghouse PWR 3-LoopAt 0049 EDT on 8/11/2008, the Harris Nuclear Plant was manually scrammed from 21% power due to indications of degrading condenser vacuum. At the time, a reactor shutdown was in progress with indications of a degraded condenser boot seal. The unit was stabilized in Mode 3 with no additional equipment failures or other complications. The reactor is currently at normal operating pressure and temperature. The highest vacuum observed was 8". All rods inserted into the core after the manual trip. Auxiliary feed water did not start as a result of the trip. Steam generator level is being maintained via normal feed water flow path. Decay heat is being removed via the steam dumps to atmosphere. There is no known primary to secondary leakage. No power-operated or manual reliefs lifted during the transient. The grid is stable and loads are being supplied via the station start-up transformer. The licensee has notified the NRC Resident Inspector.
ENS 4410931 March 2008 17:11:00OconeeAutomatic ScramNRC Region 2B&W-L-LP

Event: At 1352 hours, Unit 2 experienced a Reactor Trip due to a turbine trip. The indicated cause was low condenser vacuum. The exact cause is still under investigation but is believed to be related to a maintenance procedure in progress on the condenser vacuum instrumentation. Post-trip response was normal. Auxiliary power transferred to the start-up source (switchyard) as expected. Main Feedwater was not affected so there was no demand for Emergency Feedwater. A second High Pressure Injection pump was manually started per procedure to maintain Pressurizer level indication on scale. This is a routine action to compensate for post-trip RCS temperature and volume changes. In an apparently unrelated event, Keowee Hydro Units (KHU) 1 and 2 were shutdown from commercial operation at approximately 1425 hours. During the shutdown, the KHU 1 output breaker failed to open as expected and KHU 1 was manually locked out. The lockout removed both the Overhead and Underground Power Paths from service, making on-site emergency power unavailable to all three Oconee units. Per Tech Specs, a gas turbine unit at Lee Steam Station was started and used to energize the Oconee Standby Bus at 1518 hours. Initial Safety Significance: There is little or no safety significance to the Unit trip. The subsequent KHU 1 lockout removed both the Overhead and Underground Power Paths from service, making on-site emergency power unavailable until the Lee gas turbine was aligned. There was no demand for the Keowee emergency power function during either event. Corrective Action(s): Investigations are in progress as to the cause of the Reactor Trip and the KHU 1 lock out. Alignment of KHU 2 to the underground is currently on hold, pending evaluation of the problem which led to the KHU 1 lockout. All control rods fully inserted with decay heat being removed via the turbine bypass valves. The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 1055 EDT ON 04/02/08 FROM COREY GRAY TO HOWIE CROUCH * * *

The licensee is retracting a portion of the original report identified under 10CFR50.72(b)(3)(v)(D) based on the following: Event: UPDATE: At 1352 hours, Unit 2 experienced a Reactor Trip due to a turbine trip. The indicated cause was low condenser vacuum. The exact cause is still under investigation but is believed to be related to a maintenance procedure in progress on the condenser vacuum instrumentation. Post-trip response was normal. Auxiliary power transferred to the start-up source (switchyard) as expected. Main Feedwater was not affected so there was no demand for Emergency Feedwater. A second High Pressure Injection pump was manually started per procedure to maintain Pressurizer level indication on scale. This is a routine action to compensate for post-trip RCS temperature and volume changes. In an unrelated event, Keowee Hydro Units (KHU) 1 and 2 were shutdown from commercial operation at approximately 1425 hours. During the shutdown, the KHU 1 output breaker, ACB-1, failed to open as expected and KHU 1 was manually locked out. The lockout removed KHU 1 and the Overhead Power Path from service due to the failed ACB. Because KHU 1 was the unit aligned to the Underground Power Path, Operations declared that path inoperable also. This condition was initially reported as making on-site emergency power unavailable to all three Oconee units. Per Tech Specs, a gas turbine unit at Lee Steam Station was started and used to energize the Oconee Standby Bus at 1518 hours. In this alignment, the Lee gas turbine provides the on-site emergency power function. A design feature allows a KHU to automatically align to the Underground Path when the unit originally aligned to that path has been locked out, the overhead path is locked out and an emergency start signal exists. Keowee Operations confirmed that during the lockout event the required lockout signals were present on KHU 1 such that KHU 2 would have aligned to the Underground Power Path if an emergency start demand had occurred. As a result, the Underground Power Path remained available during this event and there was no loss of safety function. Therefore the portion of the event related to 10CFR50.72(b)(3)(v)(D) is RETRACTED. The Underground Power Path was administratively inoperable because a surveillance (SR 3.8.1.3) to verify operability of KHU 2 to the underground path, required per TS 3.8.1 Condition C.1, could not be performed. The surveillance procedure utilizes a normal start signal, which was inhibited by the lockout on the overhead path. The surveillance procedure does not include provisions for using an emergency start signal. At 1906 hours, after the overhead lockout had been reset, KHU 2 successfully completed an Operability test aligned to the Underground Power Path. Investigation determined that the failure of ACB-1, the output breaker for KHU 1, was due to a terminal strip sliding link in the trip circuit being in an intermediate position. It was damaged during repair so the entire sliding link block was replaced. The unit was successfully tested connected to each power path and was declared Operable on 4-1-08 at approximately 0600 hours. At 0930 hours on 4-1-08 the standby bus was disconnected from Lee. Initial Safety Significance: There is little or no safety significance to the Unit trip. The subsequent KHU 1 lockout removed the Overhead Power Path from service. The Lee gas turbine was aligned to energize the standby bus. Operations and Engineering subsequently confirmed that on-site emergency power remained available via KHU 2 and the Underground Power Path. There was no demand for the Keowee emergency power function during either event. Corrective Action(s): Investigation as to the cause of the Reactor Trip continues. Unit 2 is now on-line at low power and is in the process of returning to full power. ACB-1, the output breaker for KHU 1, terminal strip sliding link has been repaired/replaced; both KHUs have been tested and declared operable; the Lee gas turbine has been secured. The licensee informed the NRC Resident Inspector. Notified R2DO (Evans).

ENS 4367629 September 2007 02:19:00HarrisAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopOn September 28, 2007, while reducing power for a planned refueling outage with the reactor at approximately 30 percent power in MODE 1, an unplanned actuation of the reactor protection system occurred. At 2232 a fault pressure trip signal was received on the A Startup Transformer (SUT), causing a loss of power to Aux Buses D, A & C electrical buses as well as the A-SA safety bus. The loss of A & C buses initiated the RCP underfrequency trip which tripped the Reactor and all three RCPs as designed. The A Diesel Generator automatically started and reenergized bus A-SA as designed. The auxiliary feedwater system actuated as expected due to undervoltage on the A-SA safety bus and loss of the main feedwater pumps. All control rods inserted on the reactor trip. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system temperature and pressure following the reactor trip. Steam generator water levels are being maintained using auxiliary feedwater. All emergency core cooling system equipment is available. The plant electrical system is being restored at this time. The A SUT remains out of service. The cause of the loss of power from A SUT is under investigation. This condition is being reported as an unplanned reactor protection system actuation and specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector was notified of this event by the licensee. The plant was in natural circulation for approximately 1 hour. The Main Steam Isolation Valves (MSIVs) were manually isolated per procedure due to loss of EHC indication. Presently the B RCP has been restored to service, MSIVs are still closed, and the motor driven auxiliary feedwater pumps are feeding the Main Steam Generators. There are not any leaking steam generator tubes. The A EDG will be secured after backfeeding of the deenergized buses have been established.
ENS 4336415 May 2007 14:46:00RobinsonAutomatic ScramNRC Region 2Westinghouse PWR 3-LoopAt 11:16 am Eastern Daylight Time, a reactor trip occurred at the H. B. Robinson Steam Electric Plant, Unit No. 2. The unit was at approximately 82% reactor power and power level was being increased after restart from a refueling outage that had ended on May 13, 2007. The reactor protection system actuation was identified as a turbine trip signal that caused the reactor trip. The turbine trip signal appears to have been caused by the generator differential protection circuitry. The reactor is currently stable in MODE 3. All control rods indicated fully inserted following the reactor trip. The Auxiliary Feedwater (AFW) System actuated as expected in response to plant conditions, except the 'A' motor-driven AFW pump did not start. The plant operators manually started the 'A' motor-driven AFW pump. The cause of the 'A' AFW pump failure and the cause of the reactor trip are being investigated. The primary system and steam generator power operated relief valves and safety valves did not actuate during this event. The normal post-trip electrical lineup is providing power to the plant and the electrical system is stable at this time. Decay heat is currently being removed by use of the normal feedwater system and the condenser steam dumps. The 'B' Main Feedwater Pump also tripped during this event. The 'A' Main Feedwater Pump continued to operate and is being used to supply main feedwater to the steam generators. The licensee informed the NRC Resident Inspector.
ENS 4317921 February 2007 22:11:00Crystal RiverAutomatic ScramNRC Region 2B&W-L-LPAt 1915 on February 21, 2007, Crystal River Unit 3 was at reduced power (71%) for planned maintenance on one of four condenser waterboxes when the Integrated Control System (ICS) for Main Feedwater became erratic causing a feedwater transient that underfed the once thru steam generators. The reactor protection system (RPS) actuated on high Reactor Coolant System pressure causing a reactor trip. CR3 also received an Emergency Feedwater Actuation (EFIC) on low steam generator levels. This event is reportable as a 4-Hour Non-Emergency Notification per 10CFR50.72 (b)(2)(iv)(B) for Reactor Protection System Actuation and as an 8-Hour Non-Emergency Notification per 10CFR50.72 (b)(3)(iv)(A) for Emergency Feedwater Actuation and for Reactor Protection System actuation. All systems responded as designed and the plant remains stable with Emergency Feedwater in MODE 3. The licensee informed the NRC Resident Inspector.
ENS 4306225 December 2006 09:27:00BrunswickAutomatic ScramNRC Region 2GE-4On 12/25/06 at approximately 05:39 an automatic reactor scram occurred on Brunswick Unit 2. The Reactor Protection System (RPS) actuated on Neutron Monitoring System (APRM/OPRM) trip for APRM 2 and 4. All control rods properly inserted when the scram occurred from the RPS signal. Reactor water level reached low level 1 (LL1) and low level 2 (LL2) as a result of the scram. The LL1 signal causes a Group 2 (floor and equipment drain isolation valves), Group 6 (monitoring and sampling isolation valves) and Group 8 (shutdown cooling isolation valves) isolation signal. The LL1 isolations occurred as designed. The LL2 (signal) causes a Reactor Core Isolation Cooling (RCIC) system actuation, High Pressure Coolant Injection (HPCI) system actuation, Group 3 (reactor water cleanup valves) isolation signal, a secondary containment isolation signal, a Standby Gas Treatment (SBGT) initiation signal, a Control Room Emergency Ventilation (CREV) initiation signal, Reactor Recirculation Pump trip and an Alternate Rod Insertion (ARI) actuation signal. The low level 2 condition was reached momentarily and did not affect all instruments due to calibration differences. Initial assessment concludes that the appropriate LL2 isolations and actuations occurred as designed. Further evaluation of LL2 isolation and actuations will be conducted. The RCIC system actuation resulted in injection into the reactor as designed. The HPCI system actuated but did not inject because reactor water level was recovered. The plant is in a stable condition. An investigation is in progress to determine the cause of the Neutron Monitoring System trip. RCIC started momentarily and then was secured. Reactor water level being maintained via normal feedwater system. Decay heat being removed through the bypass valves. Normal electrical lineup for shutdown. EDGs available. Unit 1 not affected by this transient. The licensee notified the NRC Resident Inspector.
ENS 4298611 November 2006 16:31:00BrunswickManual ScramNRC Region 2GE-4At 1243 EST, during startup activities, a manual reactor scram was inserted as a result of high conductivity in the condenser. It is believed that the high conductivity was the result of a condenser tube leak. Upon receipt of the conductivity excursion alarm, abnormal operating procedures were consulted and the manual scram was inserted. Unit 2 was at approximately 1 percent of rated thermal power and reactor pressure was approximately 100 psi. At the time of the conductivity excursion, the condensate system was not in service and, as such, reactor water chemistry was not adversely affected. All safety systems operated per design. No emergency core cooling systems (ECCS) actuated. Unit 2 will be taken to mode 4 and the necessary repairs will be completed. All control rods inserted as expected. The licensee believes there is no spread of high conductivity to adjacent systems (e.g. CRD and the CST). Confirmatory samples are in progress. Decay heat is being removed by RCIC in the pressure control mode with the intention of placing shutdown cooling in-service. The electrical system is in a normal shutdown lineup. The licensee notified the NRC Resident Inspector.