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| Entered date | Site | Region | Scram | Reactor type | Event description | |
|---|---|---|---|---|---|---|
| ENS 56585 | 22 June 2023 13:50:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 1035, on June 22, 2023, with Unit 2 in Mode 1 at 100% power, the reactor automatically tripped due to `A' train reactor trip breaker and `B' train reactor trip bypass breaker opening during testing. The trip was not complex, with all systems responding normally post-trip. MST-021 (Reactor Protection Logic Train `B' At Power) testing was in progress at the time of trip. 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 actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). As a result of the reactor trip, emergency feedwater actuated; therefore, this event is being reported as an eight-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 54212 | 11 August 2019 12:14:00 | Robinson | NRC Region 2 | Automatic Scram | At 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 52290 | 8 October 2016 13:44:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse 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).
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.
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).
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).
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 49708 | 10 January 2014 00:27:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | At 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 49506 | 5 November 2013 21:31:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | At 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 47781 | 28 March 2012 17:14:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | At 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 47293 | 26 September 2011 14:48:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | At 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 46313 | 7 October 2010 04:06:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | At 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 46238 | 9 September 2010 18:04:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | 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 45799 | 28 March 2010 22:47:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse 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.
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).
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 45483 | 6 November 2009 23:36:00 | Robinson | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | A 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 44660 | 17 November 2008 09:21:00 | Robinson | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | At 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 43364 | 15 May 2007 14:46:00 | Robinson | NRC Region 2 | Automatic Scram | Westinghouse PWR 3-Loop | At 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 42933 | 25 October 2006 06:32:00 | Robinson | NRC Region 2 | Manual Scram | Westinghouse PWR 3-Loop | On October 25, 2006, at approximately 0247 hours (EDT), Main Control Room operators observed indications of a 100% load rejection and manually tripped the reactor at 0248 hours (EDT). During the event, Main Control Room operators received an alarm indicating that one or both pressurizer PORVs lifted briefly. The PORV(s) reclosed as designed. The reactor is currently stable in Mode 3. During the event, the steam generator PORVs briefly opened and re-closed, as expected. All controls 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 at this time. Decay heat is currently being removed by use of the normal feedwater system and the condenser steam dumps. The cause of this event is under investigation. Follow-up notifications will be provided if important new information is obtained or plant conditions significantly change. The licensee reported no steam generator tube leakage i.e. no radionuclide release during the brief lifting of the steam generator PORVs. The licensee notified the NRC Resident Inspector.
Subsequent to the initial notification, it has been determined that the automatic initiation of the Auxiliary Feedwater (AFW) System is also reportable under 10 CFR 50.72(b)(3)(iv)(A), Specified System Actuation. The AFW System automatically actuated as expected during the event in response to the secondary plant transient. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ernestes). |