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 Entered dateSiteRegionReactor typeEvent description
ENS 5421211 August 2019 12:14:00RobinsonNRC 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 5374519 November 2018 23:04:00RobinsonNRC Region 2On 11/19/2018, at 1916 EST, with unit 2 in Mode 5 at 0 percent power, an actuation of the 'B' (Emergency Diesel Generator) EDG occurred during troubleshooting activities with the opposite train protected. The reason for the 'B' EDG auto-start was low voltage on the E-2 bus due to its supply breaker opening. The 'B' EDG automatically started as designed when the low voltage signal was received. Following the EDG start, required loads sequenced on as designed including the 'B' (Motor Driven Auxiliary Feedwater Pump) MDAFW Pump. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Emergency AC Electrical Power System (Emergency Diesel Generator) and Auxiliary Feedwater System (Motor Driven Auxiliary Feedwater Pump). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
ENS 5308121 November 2017 11:26:00RobinsonNRC Region 2Westinghouse PWR 3-LoopA non-licensed contract employee had a confirmed positive for illegal drugs during a random fitness-for-duty test. The licensee notified the NRC Resident Inspector.
ENS 526624 April 2017 02:33:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt 2155 hours EDT on 04/03/2017, with the unit in Mode 3 at 0 (percent) power, an automatic actuation of the Auxiliary Feedwater (AFW) System occurred during surveillance testing. The cause of the AFW system auto-start was an improperly performed procedure step to bypass the auto-start logic of the AFW pumps during performance of the surveillance test. The 'A' and 'B' AFW pumps automatically started as designed when the feedwater isolation signal was received. Due to the valid actuation of the AFW system, this event requires an 8-hour non-emergency notification under 10 CFR 50.72(b)(3)(iv)(A)(B)(6). At no time did this occurrence pose undue risk to the health and safety of the public. The NRC Resident Inspector has been notified.
ENS 5233229 October 2016 15:31:00RobinsonNRC Region 2Westinghouse PWR 3-LoopThis is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as the discovered condition affects the functionality of an emergency response facility. A deficiency with the TSC and EOF ventilation system was discovered on October 29, 2016 at 1032 (EDT). The deficiency involved a fire alarm unable to be reset impeding the ability to place the TSC and EOF in recirculation mode in the event of a radiological release. The fire alarm was reset at 1204 (EDT) on October 29, 2016, restoring capability to perform full emergency assessment to the TSC and EOF. If an emergency had been declared while the fire alarm was unable to be reset requiring TSC or EOF activation during this period, the TSC and EOF would be staffed and activated using existing emergency planning procedures unless the TSC or EOF became uninhabitable due to ambient temperature. radiological, or other conditions. If relocation of the TSC or EOF became necessary, the Emergency Response Manager would relocate the TSC and/or EOF staff to an alternate location in accordance with applicable site procedures. This condition had no impact on the health and safety of the public. The NRC Resident Inspector has been notified. A malfunctioning smoke detector was removed from the system. Compensatory measures are being employed.
ENS 522908 October 2016 13:44:00RobinsonNRC 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 5219824 August 2016 16:55:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt 1338 (EDT) on 08/24/2016, a turbine trip, and a subsequent reactor trip occurred. The cause 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. As of 1643 (EDT), the Plant is in Mode 3. The current RCS (Reactor Coolant System) Temperature is 548 degrees F (Average), and the Steam Generator Levels are in the range of 40 to 45% (normal range) with levels controlled by the Main Feedwater System. All systems and equipment operated as expected. 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 NRC Resident Inspector has been notified.
ENS 5217411 August 2016 18:41:00RobinsonNRC Region 2Westinghouse PWR 3-LoopDuring testing it was discovered that the H. B. Robinson Steam Electric Plant, Unit 2 (RNP) Lake Robinson Dam Tainter Gates A and B were in degraded condition. The Tainter Gates are non-technical specification features of design that control Lake Robinson water level. RNP was not able to achieve full opening of the Tainter Gates. Subsequent analysis has concluded that the degraded condition of the Tainter Gates adversely impacts their UFSAR credited safety function. On 8/11/2016, at 1505 EDT, it was determined this places RNP in an unanalyzed condition that significantly degrades plant safety (per) 10 CFR 50.72 (b)(3)(ii)(B). The Tainter Gates have been repaired, tested and rendered complaint with their original licensing basis function. The current plant status does not pose undue risk to the health and safety of the public. The NRC Senior Resident Inspector has been notified of this condition.
ENS 5186413 April 2016 17:57:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt approximately 1430 hours EDT on 4/13/2016, it was determined that the source document for the mass and energy release parameters used to determine the containment pressure and temperature response to a Main Steam Line Break (MSLB) does not adequately account for all possible single active failure scenarios in the steam or feedwater line isolation provisions. The source document addresses the active failure of the Main Feedwater Control valves to close in the faulted steam generator feed line, but not the failure of a feedwater bypass valves to close in the faulted steam generator feed line. An active failure of a feedwater regulating bypass valve whereby the valve fails to close will increase the secondary mass available for release to the containment structure. This can result in a higher peak containment pressure that could challenge the containment design pressure. This condition is only a concern when the feedwater regulating bypass valves are in the open position in modes 1, 2 or 3, and they fail to close on a engineered safeguards actuation signal. The feedwater regulating bypass valves are currently closed. This determination is being reported as an unanalyzed condition in accordance with 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
ENS 5077828 January 2015 20:55:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt 1957 EST on 01/28/2015, with the unit in Mode 1 at 100 percent power, it was discovered that a modification installed during the fall 2013 refueling outage at H. B. Robinson inadvertently cross connected both trains of reactor protection. This cross connection resulted in both trains of safety injection being required to actuate in order to produce a reactor protection reactor trip. This is reportable pursuant to 10 CFR 50.36(c)(1)(ii)(A) since it was determined that an automatic safety system does not function as required. The cross connection of the reactor protection trains presented an unanalyzed condition and is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B). An additional unanalyzed condition was identified in which the 'A' and 'B' DC Train systems were cross connected by the same condition and is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(B). This condition is also being reported as an eight hour non-emergency under 10 CFR 50.72(b)(3)(v)(D) condition that could prevent fulfillment of safety functions. 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. This condition did place the unit in Technical Specification 3.0.3 for the reactor protection system and the DC vital buses, but the condition causing the issue was cleared at 2048 EST prior to any lowering of reactor power. The licensee will be notifying appropriate State, local and other government agencies as required.
ENS 501786 June 2014 16:40:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 1622 EDT, the licensee declared a notification of unusual event under EAL HU2.1 for a fire alarm on the third level inside the containment building. The alarm was received at 1607 EDT. Upon receipt, the licensee dispatched their fire brigade which determined that no fire existed. The fire brigade is conducting a containment walkdown to confirm that there are no issues inside of containment. The licensee did not request offsite assistance. The plant continues to operate at 100% power and is stable. The licensee has notified the NRC Resident Inspector, state and local authorities. Notified DHS, FEMA, and NICC. FEMA NWC and Nuclear SSA were notified via email.

  • * * UPDATE FROM COREY PAGE TO DONG PARK AT 1700 EDT ON 6/6/14 * * *

At 1648 EDT, the licensee terminated from their notification of unusual event. The basis for termination was visual confirmation that no fire existed inside containment. The licensee has notified the NRC Resident Inspector, the State of South Carolina, and the counties of Chesterfield, Darlington and Lee. Notified R2DO (Freeman), NRR EO (Skeen), IRD MOC (Grant), DHS, FEMA, and NICC. FEMA NWC and Nuclear SSA were notified via email.

ENS 4970810 January 2014 00:27:00RobinsonNRC 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 495065 November 2013 21:31:00RobinsonNRC 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 495025 November 2013 05:00:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt 0041 hours (EST) on 11/05/2013, while placing the Condensate Polishers into service, a secondary side perturbation occurred, resulting in a loss of the one running Main Feedwater Pump ('A' Pump) on low suction pressure. At the time, the plant was in Mode 2 with Startup Low Power Physics Testing in progress. By design, this condition resulted in an automatic start of Auxiliary Feedwater. Both 'A' and 'B' Motor-Driven Auxiliary Feedwater (AFW) Pumps started as designed. Steam generator water levels were maintained by the Auxiliary Feedwater flow. The 'B' Motor Driven AFW pump was secured following its automatic start to stabilize steam generator water levels and reactor coolant system temperature. Plant conditions including steam generator water levels have been stabilized. At 0250 hours (EST) 'A' Main Feedwater Pump was restarted and at 0252 hours (EST), the 'A' Motor-Driven AFW pump was secured. At this time, the cause of the secondary side perturbation is being investigated. The plant remains in Mode 2 and Startup Low Power Physics Testing has resumed. Due to the valid actuation of AFW, this event is being reported as an 8-hour non-emergency per 10CFR50.72(b)(3)(iv)(A). 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.
ENS 4946923 October 2013 15:25:00RobinsonNRC Region 2Westinghouse PWR 3-LoopThis is a non-emergency event notification. On 10/23/2013, a supplemental worker (i.e., a contract individual) suffered an apparent heart attack while servicing portable lighting equipment in a site perimeter parking lot. This parking lot is used by plant workers and is located outside of the Owner Controlled Area. Medical personnel who responded to this location found the worker unresponsive and transported the individual by ambulance to a local hospital for treatment. Following the evaluation by emergency room personnel, the individual was pronounced deceased at 1016 EDT. OSHA was being notified pursuant to the requirements of 29 CFR 1904.39 based on a fatality that could be work related. As such, this ENS report was made to the NRC in accordance with 10 CFR 50.72(b)(2)(xi). There was no radioactive contamination involved in this event. The health and safety of the public was not affected. The NRC Resident Inspector has been notified.
ENS 4945821 October 2013 18:21:00RobinsonNRC Region 2Westinghouse PWR 3-LoopThis is a non-emergency event notification. During planned outage work, H. B. Robinson Steam Electric Plant (HBRSEP), Unit #2 personnel encountered soils with a petroleum odor in an area excavated for inspection of Emergency Diesel Generator fuel oil piping. The inspections were being performed as part of the Buried Piping and Tanks Inspection Program. Pressure testing of the fuel oil underground piping on 10/14/13 was satisfactorily completed. Additionally, ultrasonic testing (UT) performed on the exposed diesel fuel oil pipes confirmed piping integrity. These fuel oil lines are typically pressure tested on a two year frequency and were previously tested in 2012 with similar results. The release was determined to be historical and transient in nature. The soil that was removed was replaced with uncontaminated soil. Any future actions will be coordinated with SC DHEC. HBRSEP, Unit #2 personnel are notifying SC DHEC in accordance with reporting guidance regarding the release of petroleum from an unregulated source. This event is reportable per 10 CFR 50.72(b)(2)(xi) as described in NUREG-1022, based on an event related to protection of the environment for which a notification to other government agencies has been made. The health and safety of the public was not affected. The NRC Resident Inspector has been notified.
ENS 485667 December 2012 09:30:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. On December 7, 2012, the EOF/TSC air handler chiller unit was removed from service to perform planned maintenance. This maintenance activity will not affect the air filtration portion of the system and these facilities remain available for use during an emergency. This maintenance activity will be performed in a manner to minimize the time that the air handler chiller is out of service. This maintenance activity impacts the ability to maintain ambient air temperature in the facilities. The estimated duration of this activity is planned to be 12 hours.

If an emergency condition occurs that requires activation of the emergency response facilities, the EOF and TSC will be utilized. The Emergency Response Organization team members have the ability to relocate to alternate locations in accordance with emergency implementing procedures based on conditions. Alternate emergency response facilities will remain available in the event that relocation is necessary. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. An update message will be provided when the emergency response facilities are restored. The licensee has notified the NRC Resident Inspector, State and local authorities.

* * * UPDATE FROM STEVE HEBLER TO PETE SNYDER AT 1948 EST ON 12/7/12 * * * 

The licensee completed maintenance and returned the EOF/TSC air handler chiller unit to service as of 1920 EST on 12/7/12. The licensee notified the NRC Resident Inspector. Notified R2DO (Blamey).

ENS 485585 December 2012 08:00:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

This is a non-emergency eight hour notification for a loss of emergency assessment capability. On December 5, 2012, Radiation Monitor R-9, Letdown Line Radiation Monitor, was bypassed at 0031 hours as part of planned maintenance on the Chemical Volume and Control System and is not available for monitoring letdown line radiation. This radiation monitor is used for accident assessment and is credited for Emergency Action Level (EAL) classification for an Unusual Event in the Robinson Nuclear Plant Emergency Plan. Additionally, this monitor is one of multiple indicators used to detect the loss of a fission product barrier. Inability to classify an EAL due to an out of service monitor is considered a loss of accident assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2. Actions are in place to restore the monitor to functional status. The NRC Resident Inspector will be notified.

  • * * UPDATE AT 2028 EST ON 12/5/12 FROM GEORGE CURTIS TO S. SANDIN * * *

Radiation Monitor R-9 was returned to service at 2022 EST. The licensee informed the NRC Resident Inspector. Notified R2DO (Blamey).

ENS 4850814 November 2012 16:13:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. On November 14, 2012, Radiation Monitor R-9, Letdown Line Radiation Monitor was removed from service at 0927 (EST) hours following failure of the internal and external source checks during performance of radiation monitor source checks. Both source checks failed high. This radiation monitor is used for accident assessment and is credited for Emergency Action Level (EAL) classification for an Unusual Event in the Robinson Nuclear Plant Emergency Plan. Additionally, this monitor is one of multiple indicators used to detect the loss of a fission product barrier. Inability to classify an EAL due to an out of service monitor is considered a loss of accident assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2. Actions are in place to restore the monitor to functional status. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 11/17/12 AT 0812 EST FROM GEORGE CURTIS TO DONG PARK * * *

At 1717 EST on 11/16/12, the Letdown Line Radiation Monitor was placed back in service and is functional. The licensee has notified the NRC Resident Inspector. Notified R2DO (Widmann).

ENS 4835428 September 2012 09:00:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. On September 28, 2012, the EOF/TSC air handler chiller unit was removed from service to perform planned maintenance. This maintenance activity will not affect the air filtration portion of the system and these facilities remain available for use during an emergency. This maintenance activity will be performed in a manner to minimize the time that the air handler chiller is out of service. This maintenance activity impacts the ability to maintain ambient air temperature in the facilities. The (estimated) duration of this activity is planned to be 4 hours. If an emergency condition occurs that requires activation of the emergency response facilities, the EOF and TSC will be utilized. The Emergency Response Organization team members have the ability to relocate to alternate locations in accordance with emergency implementing procedures based on conditions. Alternate emergency response facilities will remain available in the event that relocation is necessary. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. An update message will be provided when the emergency response facilities are restored. The licensee notified the NRC Resident Inspector, the State of South Carolina and the local counties of Lee, Chesterfield and Darlington.

  • * * UPDATE FROM ARNOLD TO KLCO ON 9/28/12 AT 1125 EDT * * *

The EOF/TSC Chiller is back in service as of 1102 (EDT) on 9/28/12. The ability to maintain ambient air temperature in the EOF/TSC facilities has been restored. The licensee notified the NRC Resident Inspector. Notified the R2DO (Nease).

ENS 482888 September 2012 08:16:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the work activity affects the functionality of an emergency response facility. Planned maintenance activities are being performed today to the Emergency Offsite Facility (EOF)/Technical Support Center (TSC) HVAC. The work entails replacement of a pressure switch. The filtration portion of the system will not be affected by this work. This work activity is planned to be performed and completed expeditiously within about 3.5 hours including establishing and removing the clearances and performing post maintenance testing; however, restoration time required during the maintenance could exceed the time required to activate the TSC.

If an emergency condition occurs that requires activation of the EOF and TSC, plans are to utilize the EOF and TSC during the time this work activity is being performed as long as habitability conditions allow. The Emergency Response Organization team members will be relocated to alternate locations if required by habitability conditions in accordance with emergency implementing procedures. Alternate emergency response facilities will remain available in the event that relocation is necessary." The licensee has notified the NRC Resident Inspector. Licensee has also notified state and local agencies.

  • * * UPDATE FROM GEORGE CURTIS TO DONALD NORWOOD AT 1025 EDT ON 9/8/2012 * * *

The maintenance work was completed. The TSC and EOF were declared operable as of 1025 EDT. The licensee will notify the NRC Resident Inspector. Notified R2DO (Lesser).

ENS 4819917 August 2012 06:32:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 0552 EDT Control Room Operators experienced a Loss of Annunciators. At 0607 EDT, the licensee declared an Unusual Event per EAL SU4.1, " Unplanned loss of most or all Annunciators or Indicators associated with Safety Systems on the RTGB (Remote Turbine Gage Board) Sections A and B, Primary System Annunciators and Indicators > 15 minutes." During an alarm test, the annunciators lit up as expected but did not clear at the completion of the test. Exit criteria for the Unusual Event will be restoration of the annunciator functions. The licensee has compensatory measures in effect. ERFIS (Emergency Response Facility Information System) and the normal RTGB instruments are available. The licensee notified the state and informed the NRC Resident Inspector. Notified Other FEDS (FEMA, DHS and NNSA via email).

  • * * UPDATE FROM HILL TO KLCO ON 8/17/2012 AT 0954 EDT * * *

On August 17, 2012 at 0940 EDT, the licensee terminated the Unusual Event based on restoring annunciator capability. The licensee notified the NRC Resident Inspector and has terminated ERDS transmission. Notified NRR-EO(Thomas), R2DO(Guthrie), IRD(Gott) and other FEDS (FEMA, DHS and NNSA via email)

ENS 4818011 August 2012 07:58:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt 0730 hours EDT on Saturday, August 11, 2012, the H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, Technical Support Center (TSC)/Emergency Response Facility (EOF) air conditioning system will be removed from service to facilitate the replacement of the HVAC chiller system. Unavailability of the primary chiller system will last greater than 8 hours. The duration of work is expected to be approximately 31 hours. The filtration portion of the system will not be affected by this work. In order to ensure habitable conditions in the TSC/EOF, as a compensatory measure, a temporary chiller system will be onsite to maintain habitability in the event of emergency facility activation and will remain onsite until such time that the TSC/EOF ventilation system has been returned to service. Emergency Responders assigned to these facilities and all offsite agencies have been informed of this scheduled maintenance and will be informed of any updates. Alternate emergency response facilities will remain available in the event that relocation is necessary. TSC/EOF ventilation system maintenance and post maintenance testing is scheduled to be completed by 1400 hours EDT on Sunday August 12, 2012. The NRC Resident Inspector has been informed.
ENS 4801111 June 2012 10:03:00RobinsonNRC Region 2Westinghouse PWR 3-LoopOn May 4, 2011, at approximately 1325 Eastern Daylight Time (EDT), with H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, in Mode 1 at 100% power, it was determined that over the last 40 years, HBRSEP, Unit No. 2 periodically performed cleanup of the Refueling Water Storage Tank (RWST) by aligning the non-seismically qualified refueling water purification system to the safety related and seismically qualified RWST without recognizing that the action rendered the RWST inoperable. As a result, on multiple occasions, the RWST was inoperable for a period longer than allowed by Technical Specification (TS) Limiting Condition for Operation 3.5.4, Emergency Core Cooling Systems Refueling Water Storage Tank. 'The cause of this event was that regulatory requirements for the separation of seismically qualified and non-qualified systems, structures, and components were not adequately incorporated into the Design Basis Document (DBD) and Updated Final Safety Analysis Report (UFSAR). A clearance order has been placed in effect to ensure restrictions for piping that could affect the operability of the RWST remain in place. This event was described in Licensee Event Report 2011-001-00 and was initially reported in accordance with 10 CFR 50.73(a)(2)(i)(B), any operation or condition which was prohibited by the plant's Technical Specifications. However, further reviews determined at 0800 on June 11, 2012, that the event is also reportable under 10 CFR 50.72(b)(3)(v)(D), Event or Condition that could have prevented fulfillment of a safety function. During the three year period prior to May 4, 2011, it is estimated that the purification loop was in service aligned to the RWST while on-line nine times, totaling 297 days. The licensee has notified the NRC Resident Inspector.
ENS 4782812 April 2012 09:42:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At approximately 0930 hours EDT on Thursday, April 12, 2012, the H. B. Robinson Steam Electric Plant, Unit No. 2, Technical Support Center (TSC)/Emergency Response Facility (EOF) air conditioning and charcoal filtration systems will be removed from service to facilitate the replacement of the charcoal filtration media. The duration of work is expected to be approximately 11 hours. Since the unavailability will last greater than 8 hours, this is considered a Loss of Emergency Assessment Capability, and reportable under 10 CFR 50.72(b)(3)(xiii). Due to the inability of the TSC/EOF ventilation system to maintain a habitable atmosphere, as a compensatory measure, Emergency Responders assigned to these facilities have been informed to report to the alternate facilities until such time that the TSC/EOF ventilation system has been returned to service. TSC/EOF ventilation system maintenance and post maintenance testing is scheduled to be completed by 2030 hours EDT on Thursday April 12, 2012. The NRC Resident Inspector has been informed.

  • * * UPDATE FROM WARREN WONKA TO HOWIE CROUCH AT 0435 EDT ON 4/13/12 * * *

At 1814 EDT on 4/12/12, maintenance on the TSC/EOF ventilation system was completed and the TSC/EOF was returned to service. Notified R2DO (O'Donohue).

ENS 4778128 March 2012 17:14:00RobinsonNRC 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 4774816 March 2012 10:56:00RobinsonNRC Region 2Westinghouse PWR 3-LoopThe Refueling Water Storage Tank (RWST) was placed on purification in accordance with (site procedure) OP-913, Refueling Water Purification Pump Operation, as directed from OP-301-1, Chemical and Volume Control System (Infrequent Operation), at 0400 (EDT) on 3/16/2012 to support make up of level to the RWST. This condition, connection of the purification loop, is not currently allowed based on unresolved seismic concerns with purification piping to the RWST. This was later discovered during a log review at 0545, and operators were immediately directed to remove the RWST from purification. ITS 3.5.4 was applied from 0400 based on when it was determined that this condition had been entered. ITS 3.5.4 was exited at 0622 when the RWST was removed from purification. This is being reported pursuant to 50.72(b)(3)(v), Event or Condition That Could Have Prevented Fulfillment of a Safety Function. The licensee notified the NRC Resident Inspector.
ENS 4766314 February 2012 12:55:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 1240 EST, on February 14, 2012, power was removed to a major portion of the Emergency Response Facility Information System (ERFIS) to perform a planned modification on Power Panel - 8. This work will install a new breaker in PP-8 requiring that the panel be de-energized for the maintenance. The expected duration of ERFIS inoperability is approximately 6 hours. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above-described functions. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still be made, if required, during the time that the ERFIS computer system is inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. An additional message will be provided when the ERFIS is restored. It should also be noted that during the period of ERFIS inoperability, it is likely that the system could be restored within one hour to support Emergency Response Facility activation. This report is provided to conservatively cover the possibility that restoration within one hour may not be able to be accomplished if facility activation were to occur. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM KEN BOYD TO DONALD NORWOOD AT 1516 EST ON 12/14/2012 * * *

At 1240 EST, on February 14, 2012, the Emergency Response Facility Information System (ERFIS) computer system became inoperable. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, and the Inadequate Core Cooling Monitor (ICCM). Actions were completed to restore the ERFIS computer system to an operable status at 1458 EST on February 14, 2012. Alternate methods, as described in plant procedures, were available for the above-described functions during the time that the ERFIS computer system was inoperable. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still have been made, if required, during the time that the ERFIS Computer system was inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. The licensee notified the NRC Resident Inspector. Notified R2DO(Desai).

ENS 4761823 January 2012 18:51:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt 1804 EST on 01/23/12, Unit 2 declared an Unusual Event due to an electrical fire in the control cabinet on the Polar Crane. The fire brigade responded, de-energized the equipment and confirmed that the fire was extinguished. There were no injuries as a result of this incident. At 1837 EST, the license terminated the Unusual Event classification. Unit 2 is currently in mode 5 (Cold Shutdown) for a Refueling Outage. The licensee informed state/local agencies and the NRC Resident Inspector.
ENS 4760017 January 2012 22:35:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt approximately 1814 hours (EST) on January 17, 2012, H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, personnel confirmed that the Station Battery Performance Capacity Test (Five Year Interval), for station battery 'B' was not completed within its required periodicity plus grace period. Based upon this information, plant Operations personnel declared the 'B' battery inoperable and entered the required action for Improved Technical Specifications (ITS) 3.8.4, Condition 'A', which requires that the DC electrical power subsystem be restored to operable status in 2 hours. At 2014 hours (EST), Operations personnel entered the required action for ITS 3.8.4, Condition 'B' due to required actions and associated completion time for Condition 'A' not met. Condition 'B' requires the unit to be in Mode 3 within 6 hours AND be in Mode 5 within 36 hours. This report is being made in accordance with 10 CFR 50.72(b)(2)(i) based on the initiation of a plant shutdown required by the Technical Specifications. Following the shutdown, the unit will commence a refueling outage that was originally scheduled to begin at 0000 hours (EST) on January 21, 2012. The licensee informed the NRC Resident Inspector.
ENS 4755123 December 2011 16:54:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

The licensee declared an alert condition based on the inadvertent discharge of Halon gas into their Emergency Bus Room during system testing. Plant operation was unaffected and the unit continued to operate at full power. There were no injuries or equipment damage due to the event. The licensee observed no fire or flames in the area. The licensee evacuated the area and provided ventilation to disperse the Halon. The licensee notified the NRC Resident Inspector. Also notified HHS (Jewel Wright).

* * * UPDATE FROM BRIAN WALDSMITH TO DONALD NORWOOD ON 12/23/11 AT 1822 EST * * * 

At 1810 EST on 12/23/11, Robinson terminated the Alert condition based on a return to a habitable atmosphere in the Emergency Bus Room. "The cause for the Halon System actuation appears to have been a Human Performance Error. The test/inhibit switch was placed in the incorrect position." There were no radiological releases for this event. The licensee notified state, local and other governmental agencies concerning this event. Notified R2DO (McCoy), NRR EO (Giitter), IR (Gott), DHS SWO (Arnold), FEMA (O'Connell), DOE (Jackson), USDA (Walters), and HHS(Wright).

ENS 475016 December 2011 13:54:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 0900 hours EST, on December 6, 2011, Emergency Preparedness personnel determined that the Unusual Event and Alert EAL criteria for the Liquid Waste Disposal Effluent could not be achieved due to the monitor range capability. Specifically, the R-18 effluent monitor instrument range has a maximum range of 1.0E+06 with the alarm set at 1.0E+06. The UE and Alert criteria, which are 2 times the alarm and 200 times the alarm respectively, both exceed the instrument range. The NRC Resident Inspector has been notified. This condition has existed for some time (at least three years) and was discovered during a review of the Operating Experience database involving a similar condition reported by Crystal River-3. The licensee maintains the ability to perform grab samples of the liquid effluent for assessment.

  • * * UPDATE ON 12/08/11 AT 2330 EST FROM RAY BUZARD TO JOHN KNOKE * * *

As a follow-up to the condition reported above, the licensee inspected other discharge and effluent radiation monitors for similar conditions and provided the following update: At 2250 hours EST, on December 8, 2011, it was determined that the Alert EAL classification criteria for the Steam Generator Blowdown radiation monitors and the Condensate Polisher Sump discharge radiation monitor could not be achieved due to the monitors range capability. Specifically, the R-19 A, B, C and the R-37 monitors instrument range have a maximum range of 1.6E+06 which is less than the Alert EAL classification criteria. This information was validated as a follow-up to a similar event which was reported on December 6, 2011 in EN# 47501. The NRC Resident Inspector has been notified. R2DO (Musser) notified.

ENS 4730229 September 2011 06:56:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

During plant startup, a single train fire alarm was received in containment at 0540 EDT on 9/29/2011. An Unusual Event was declared at 0602 EDT based on the containment not being accessible within 15 minutes. An inspection in containment revealed no fire or smoke or the cause for receipt of the alarm. The licensee has notified the State, Counties and the NRC Resident Inspector.

  • * *UPDATE FROM JOE PENNINGTON TO VINCE KLCO ON 9/29/2011 AT 0744* * *

The Unusual Event was terminated at 0714 EDT on 9/29/2011. The licensee has notified the State, Counties and the NRC Resident Inspector. Notified the R2DO (Nease), NRR (Thomas), IRD (Gott), FEMA (Casto) and DHS (Rickerson).

ENS 4729326 September 2011 14:48:00RobinsonNRC 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 468287 May 2011 08:33:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 0827 hours EDT, on May 7, 2010, the Emergency Response Facility Information System (ERFIS) computer system was removed from service to perform a planned modification of the ERFIS. This modification will upgrade the electrical infrastructure. The expected duration of ERFIS inoperability is approximately 4 hours. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above-described functions. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still be made, if required, during the time that the ERFIS computer system is inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. An additional message will be provided when the ERFIS is restored. It should also be noted that during the period of ERFIS inoperability, it is likely that the system could be restored within one hour to support Emergency Response Facility activation. This report is provided to conservatively cover the possibility that restoration within one hour may not be able to be accomplished if facility activation were to occur. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM WARREN WONKA TO JOHN KNOKE AT 1516 EDT ON 5/7/11 * * *

The licensee reported that the Emergency Response Facility Information System (ERFIS) computer system was returned to service. The NRC Resident Inspector has been notified. Notified R2DO (David Ayres)

ENS 463177 October 2010 18:08:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt about 1315 hours EDT on October 7, 2010, with the unit in Mode 3 and operators performing recovery actions following a reactor trip that had occurred at 0013 hours (see EN#46313), it was discovered that actions that had been performed to restore the main feedwater system had inadvertently resulted in disabling the feedwater isolation function. The feedwater isolation function, as described in Technical Specifications Section 3.3.2, Table 3.3.2 1, Function 5, requires that the feedwater isolation function be operable in Modes 1, 2, and 3, when the feedwater system is not isolated by the main feedwater isolation valves, main feedwater regulating valves, and bypass valves or by a closed manual valve. At approximately 1018 hours, during actions being taken to restore operation of the main feedwater system, the feedwater isolation key switches for the three steam generators, A, B, and C, were placed in the override/reset position. Although it was not realized at that time, this action was contrary to the Technical Specifications Section 3.3.2 operability requirements for the feedwater isolation function. This inoperability of the feedwater isolation function would have prevented the automatic feedwater isolation function described in the basis of Technical Specifications Section 3.3.2, which states that the primary function of the feedwater isolation signal is to stop excessive flow of feedwater into the steam generators. It also states that this function is necessary to mitigate the effects of overfeeding the steam generators, which could result in overcooling of the primary system. This function is actuated by a safety injection signal. There is no Technical Specifications allowed condition for both trains of the feedwater isolation function to be inoperable. Therefore, Technical Specifications Limiting Condition for Operation (LCO) 3.0.3 was applicable from the time the feedwater isolation switches were placed in the override/reset position, until the feedwater isolation function operability was restored at approximately 1329 hours. The LCO 3.0.3 completion time to be in Mode 4 within 13 hours was not exceeded. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(v)(D), any event or condition that at the time of discovery could have prevented the fulfillment of a safety function. The licensee notified the NRC Resident Inspector.
ENS 463137 October 2010 04:06:00RobinsonNRC 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 18:04:00RobinsonNRC 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 4620425 August 2010 08:47:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 0845 hours EDT, on August 25, 2010, the Emergency Response Facility Information System (ERFIS) computer system was removed from service to perform a planned modification of the ERFIS. This modification will remove obsolete software and hardware related to the Emergency Response Data System, which has been replaced by an upgraded system. The expected duration of ERFIS inoperability is approximately 3 hours. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above-described functions. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still be made, if required, during the time that the ERFIS Computer system is inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As previously stated, alternate means remained available to assess plant conditions, make notifications, and accomplish required communications, as necessary. An additional message will be provided when the ERFIS is restored. It should also be noted that during the period of ERFIS inoperability, it is likely that the system could be restored within one hour to support Emergency Response Facility activation. This report is provided to conservatively cover the possibility that restoration within one hour may not be able to be accomplished if facility activation were to occur. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM DON GRANT TO JOE O'HARA AT 1454 ON 8/25/10 * * *

At 1430, the ERFIS system was returned to service. The NRC Resident Inspector has been notified. Notified the R2DO(Franke)

ENS 4612726 July 2010 14:18:00RobinsonNRC Region 2Westinghouse PWR 3-LoopAt 1314 hours (EDT) on May 25, 2010, with the unit defueled during Refueling Outage 26, an inadvertent Safety Injection (SI) Signal occurred. The signal was received when Safeguards Train 'A' breaker DP-A-20 was closed during the performance of procedure PIC-020, 'Time Delay Relay Calibration Safeguards Train 'A'.' The SI Signal resulted in a containment phase 'A' isolation, containment ventilation isolation, control room ventilation transfer to emergency pressurization mode, and Radiation Monitors R-11 and R-12 (Containment Vessel Airborne Particulate and Gas Monitors) isolation. Emergency Bus 1 and Emergency Diesel Generator 'A' were under clearance, including the Train 'A' sequencer. Therefore, automatic loading of the 'A' Train Sequencer did not occur and no actual injection into the reactor vessel occurred. The affected systems actuated as expected. The cause of the inadvertent SI Signal resulted from SI Initiation Latching Relay SIA1 when power was restored to Safeguards Rack 51. Subsequent investigation determined the SIA1 relay was in an unexpected position (i.e., latched). The SIA1 relay likely became inadvertently latched while performing cable pulls in Safeguards Rack 51 for work associated with Water Cooled Condensing Unit 1A. Corrective actions include a revision to a procedure to address additional steps to require reset of SI Initiation Latching Relays SIA1 and SIA2 after cleaning and lubrication and development of a planning tool to better assess the risk associated with work being performed (on) plant equipment. In addition, a procedure will be developed for Operations to manually reset safeguards SI Initiation Latching Relays SIA1 and SIA2 prior to restoring system power. These actions are expected to be completed before or on December 16, 2010. The licensee has notified the NRC Resident Inspector.
ENS 4604524 June 2010 21:34:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 14:34 hours on June 24, 2010, with the unit in MODE 5, Cold Shutdown, with approximately 50% pressurizer level, during Refueling Outage 26, Instrument Buses 3 and 8 unexpectedly de-energized during performance of testing in accordance with procedure OST-163, 'Safety Injection Test and Emergency Diesel Generator Auto Start on Loss of Power and Safety Injection.' The loss of Instrument Buses 3 and 8 occurred during the loss of power and Safety Injection testing of the 'A' Train. Instrument Buses 3 and 8 are normally powered from Inverter 'B' which is normally supplied by the Train 'B' DC Bus. During the test, it was noted that the power supply to Instrument Buses 3 and 8 had tripped. The cause of the failure of Inverter 'B' is not currently known. The failure of inverter 'B' caused the closure of the Residual Heat Removal (RHR) Heat Exchanger discharge valve (HCV-758) and the RHR Heat Exchanger bypass valve (FCV-605). Both trains of RHR continued to operate and reactor coolant system temperature remained in the range of approximately 93 to 96 degrees Fahrenheit. Abnormal Operating Procedure AOP-020, 'Loss of Residual Heat Removal (Shutdown Cooling)' was entered. Power was restored to Instrument Buses 3 and 8 by use of the alternate power supply at 14:49 hours. Normal configuration of the RHR system was restored and AOP-020 was exited at 14:51 hours. Currently Instrument Buses 3 and 8 are being powered from the alternate power supply which causes the associated 'B' EDG to be inoperable due to the inoperability of the automatic load sequencer that starts the associated Service Water and Component Cooling Water pumps. The 'A' EDG is inoperable due to the need to complete required post-maintenance testing. Therefore, both EDGs are currently inoperable. Both EDGs are currently considered available and are aligned for automatic starting. Both EDGs would be expected to automatically supply their respective buses if a loss of offsite power were to occur. Manual action would be required to start the required loads on the 'B' Train due to the current alignment of the Instrument Buses 3 and 8 on the alternate power supply. It is expected that the 'B' EDG will be restored to operable status when Inverter 'B' is restored to operable status and realigned to supply Instrument Buses 3 and 8. The Technical Specifications (TS) Action Statement currently in effect for loss of Inverter 'B' (TS 3.8.8 Condition A) requires initiation of action to restore AC instrument bus sources to OPERABLE status immediately. The actions to restore Inverter 'B' were initiated immediately and are continuing. This report is being made in accordance with 10 CFR 50.72(b)(3)(v)(D), for any event or condition that at the time of discovery could have prevented the fulfillment of structures or systems that are needed to mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE AT 1408 ON 6/27/2010 FROM ASHLEY VALONE TO MARK ABRAMOVITZ * * *

This is a follow-up notification to Event Notification EN #46045 regarding Instrument Buses 3 and 8 that unexpectedly de-energized during performance of testing in accordance with procedure OST-163, 'Safety Injection Test and Emergency Diesel Generator Auto Start on Loss of Power and Safety Injection.' Power was restored to Instrument Buses 3 and 8 by use of the normal power supply at 08:03 hours on June 27, 2010. Inverter 'B' has been realigned to supply Instrument Buses 3 and 8. The restoration of inverter 'B' has returned the associated 'B' EDG to operable status at 11:09 hours with the return of the automatic load sequencer that starts the associated Service Water and Component Cooling Water pumps. The 'A' EDG continues to be inoperable due to the need to complete required post-maintenance testing. Currently 'A' EDG is considered available and aligned for automatic starting. The Technical Specifications (TS) Action Statement for loss of Inverter 'B' (TS 3.8.8 Condition A) that requires initiation of action to restore AC instrument bus sources to OPERABLE status immediately was exited at 11:09 hours on June 27, 2010. TS Action Statement for loss of 'B' EDG (TS 3.8.2 Condition B) that requires initiation of action to restore required DG to OPERABLE status immediately was also exited at 11:09 hours on June 27, 2010. The licensee is still investigating the cause of the failure. The licensee notified the NRC Resident Inspector. Notified the R2DO (Shaeffer).

ENS 4579928 March 2010 22:47:00RobinsonNRC 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 454836 November 2009 23:36:00RobinsonNRC 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 4466017 November 2008 09:21:00RobinsonNRC 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 436883 October 2007 18:43:00RobinsonNRC Region 2Westinghouse PWR 3-LoopOn October 3, 2007, a non-work related fatality occurred at the H. B. Robinson Steam Electric Plant, Unit No. 2. Specifically, at 1502 hours EDT, first responders were dispatched for a Progress Energy employee who suffered a medical condition that required immediate attention. Medical assistance was provided on-site by H. B. Robinson Plant first responders, and the employee was subsequently transported by ambulance to a nearby medical facility. At 1540 hours EDT, site personnel were notified that attempts to revive the employee were not successful and the employee had died. The employee was working in a non-radiologically controlled area of the plant and no radioactive material or contamination was involved. Notification of other governmental agencies, specifically the South Carolina Occupational Safety and Health Administration (SCOSHA), was completed at 1830 hours EDT on October 3, 2007. This report is being made in accordance with 10 CFR 50.72(b)(2)(xi), which states that a report shall be made as soon as practical and in all cases within four hours of any event or situation, related to the health and safety of the public or on-site personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an on-site fatality. The resident inspector has been notified.
ENS 4345027 June 2007 12:56:00RobinsonNRC Region 2Westinghouse PWR 3-LoopThis telephone notification to report an invalid actuation is provided in accordance with 10 CFR 50.73(a)(1), which states, 'In the case of an invalid actuation reported under Sec. 50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER.' The specific reporting requirement in 10 CFR 50.73(a)(2)(iv)(A), states, 'Any event or condition that resulted in manual or automatic actuation of any of the systems listed in paragraph (a)(2)(iv)(B).' For this report, the affected system was the emergency AC electrical power systems as listed in 10 CFR 50.73(a)(2)(iv)(B)(8). Specifically, the Train A emergency diesel generator (EDG) automatically started due to an invalid system actuation. On May 1, 2007, at approximately 2339 hours (EDT), with H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2, in MODE 5 with reactor coolant system temperature at approximately 84 degrees F, during activities associated with surveillance testing as part of Refueling Outage 24, the Train A EDG automatically started from the standby condition. The Train A EDG did not automatically connect to the associated emergency bus (E-1), because no E-1 bus undervoltage (UV) signal was present. The automatic start was caused by the inadvertent grounding of a test connection lead that was being used to monitor the Train A EDG during testing. The grounding of the test lead resulted in a drop in voltage on the starting circuitry that caused the actuation of the 4A diesel start relay. The Train A EDG successfully started and achieved the required speed and voltage. As previously stated, the EDG did not load because no UV signal was present on the associated E-1 bus. No failures or abnormalities were noted. The Train B EDG remained operable during this event and the Train A EDG was able to provide power to E-1 emergency bus, if necessary. The invalid EDG start event was entered into the corrective action program for HBRSEP, Unit No. 2. The investigation of this event has been completed. The results of the investigation determined that the EDG start was caused by improper isolation of a safety significant power source from the field installed equipment. The Plant Nuclear Safety Committee has reviewed the investigation associated with this reportable event and concurred with the investigation. Corrective actions for this event are in progress and are being tracked via the corrective action program." The licensee informed the NRC Resident Inspector.
ENS 4336415 May 2007 14:46:00RobinsonNRC 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 4313430 January 2007 10:31:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 0543 hours EST, on January 30, 2007, the Emergency Response Facility Information System (ERFIS) computer system became inoperable due to multiplexer (MUX) failure. At this point, it appears that the malfunction is due to a failure of the MUX power supply. Technical support personnel are currently working to restore ERFIS functionality. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above described functions. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still be made, if required, during the time that the ERFIS computer system is inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As previously stated, alternate means remain available to assess plant conditions, make notifications, and accomplish required communications, as necessary. An update to this message will be provided when the ERFIS is restored. The licensee notified the NRC Resident Inspector.

  • * * UPDATE ON 01/30/07 AT 1729 EST BY HENRY CURRY TO MACKINNON * * *

ERFIS returned to service at 1700 EST. R2DO (M. Ernstes) notified. The NRC Resident Inspector will be notified of this update by the licensee.

ENS 4312727 January 2007 09:10:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 0905 hours EST, on January 27, 2007, the Emergency Response Facility Information System (ERFIS) computer system was removed from service to complete a planned modification of the ERFIS power supply system. The expected duration of ERFIS inoperability is approximately 2 hours. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above described functions. Therefore, it is expected that appropriate assessment or plant conditions, notifications, and communications could still be made, if required, during the time that the ERFIS computer system is inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As previously stated, alternate means remain available to assess plant conditions, make notifications, and accomplish required communications, as necessary. An update to this message will be provided when the ERFIS is restored. This message is similar to the January 6, 2007 event notification number 43081. At that time the ERFIS was being placed on an alternate power supply to allow modification of the normal power supply. This interruption of the ERFIS is due to the completion of that planned modification, which allows restoration of the ERFIS to the normal power supply. The licensee notified the NRC Resident Inspector.

  • * * UPDATE ON 01/27/07 AT 1347 EST FROM LEONARD WIEGAND TO MACKINNON * * *

Emergency Response Facility Information System (ERFIS) returned to service at 1150 EST. The NRC Resident Inspector will be informed of this update by the licensee. R2DO (Brian Bonser) notified

ENS 430816 January 2007 10:00:00RobinsonNRC Region 2Westinghouse PWR 3-Loop

At 0904 hours EST, on January 6, 2007, the Emergency Response Facility Information System (ERFIS) computer system was removed from service to perform a planned modification of the ERFIS power supply system. The expected duration of ERFIS inoperability is approximately 2 hours. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS), Safety Parameter Display System (SPDS), Meteorological Data link system, and the Inadequate Core Cooling Monitor (ICCM). The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above-described functions. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still be made, if required, during the time that the ERFIS Computer system is inoperable. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As previously stated, alternate means remained available to assess plant conditions, make notifications, and accomplish required communications, as necessary. An additional message will be provided when the ERFIS is restored. The licensee will notify the NRC Resident Inspector.

  • * * UPDATE PROVIDED BY HENRY CURRY TO JEFF ROTTON AT 1158 EST ON 01/06/07 * * *

ERFIS computer has been returned to operable status at 1146 EST. Licensee notified NRC Resident Inspector. Notified R2DO(Payne)