Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56585 | 22 June 2023 14:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Occured During Protection System Testing | 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. | Feedwater Reactor Protection System Main Condenser | |
| ENS 55099 | 9 February 2021 13:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to an Inadvertent Actuation of Emergency Sirens | At approximately 0800 on February 9, 2021, thirty-one (31) H.B. Robinson Nuclear Plant Offsite Emergency Notification sirens in Darlington County, SC were inadvertently actuated. The Darlington County Emergency Services and South Carolina Emergency Management Division were promptly notified. The actuation lasted for three (3) minutes at full volume. The cause of the actuation is under investigation at this time. Capability to notify the public was never degraded during the inadvertent actuation. All Emergency Notification sirens remain in service. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi), Offsite Notification, as a four (4) hour report. The NRC Resident Inspector has been notified. A local news agency did report about the alarms sounding and reported that there was no concern at the site. | ||
| ENS 55079 | 1 December 2020 14:46:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-DAY Optional Telephonic Notification of an Invalid Specified System Actuation | This 60-day optional telephone notification is being made in lieu of an LER submittal, as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At 0946 hrs on December 1, 2020, with unit 2 in Mode 5 at 0% power, an invalid actuation of the Emergency Diesel Generators (EDG) 'A' and 'B', 'A' Residual Heat Removal (RHR) Pump, 'A' Service Water Booster Pump (SWBP), and Auxiliary Feed Water (AFW) Pumps 'A' and 'B' occurred. The actuation was caused by a Safety Injection (SI) signal while installing simulations to support Reactor Safeguards testing. The SI signal occurred when two out of three logic was met for Low Pressurizer Pressure, which was caused by a high resistance connection to a test point from a loose test lead. All aligned equipment, 'A' and 'B' EDGs, 'A' RHR Pump, 'A' SWBP and 'A' and 'B' AFW Pumps, responded properly to the auto-start signal and the actuation was complete. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified. | Service water Emergency Diesel Generator Auxiliary Feedwater Residual Heat Removal | |
| ENS 54796 | 21 July 2020 12:51:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown | At 0851 EDT on July 21, 2020, a Technical Specification required shutdown was initiated at Robinson Unit 2. Technical Specification LCO 3.0.3 was entered due to LCO 3.1.7 not being met as a result of indication loss on Control Rod positions with more than one position indication inoperable for a group. LCO 3.0.3 was entered at 0752 EDT to initiate action within 1 hour to place the unit in MODE 3 within 7 hours. Since a Technical Specification required shutdown was initiated, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). Technical Specification LCO 3.0.3 was exited at 1003 EDT on July 21, 2020. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Shutdown was initiated and power was reduced approximately 3 percent. Reactor power was back to 98.5 percent at the time of notification. | Control Rod | |
| ENS 54768 | 10 July 2020 13:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Common Emergency Operations Facility Maintenance | At 0900 EDT hours on 7/10/2020 Duke Energy will undertake planned maintenance activities on the common Emergency Operations Facility (EOF) for Brunswick, Catawba, Harris, McGuire, Oconee, and Robinson nuclear sites. The work includes performance of upgrades to the emergency AC power system and requires the removal of both normal and emergency power to the facility. The work duration is approximately ten (10) days. If a declared emergency were to occur at Robinson, the Alternate EOF would be set up in the Catawba Alternate Technical Support Center (TSC) location as described in implementing procedures. The Emergency Response Organization has been notified that the primary EOF will be unavailable during the upgrade project and to report to the alternate location, if activated. This is an eight-hour, non-emergency 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. The NRC Resident Inspector has been notified. | ||
| ENS 54212 | 11 August 2019 12:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | 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. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 53745 | 19 November 2018 05:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Actuation Due to Low Voltage Signal | On 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. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 53081 | 20 November 2017 20:48:00 | 10 CFR 26.719, FFD Reporting requirements | Fitness-For-Duty Report | A 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 52662 | 4 April 2017 01:55:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Auxiliary Feedwater System During Surveillance Testing | At 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. | Feedwater Auxiliary Feedwater | 05000261/LER-2017-001 |
| ENS 52332 | 29 October 2016 14:32:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center (Tsc) and Emergency Operations Facility (Eof) Ventilation System Out of Service | 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) 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 52290 | 8 October 2016 17:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to Loss of Offsite Power | 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). | Steam Generator Service water Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Decay Heat Removal | 05000261/LER-2016-005 |
| ENS 52198 | 24 August 2016 17:38:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due Automatic Turbine Trip from 100 Percent Power | At 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. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | 05000261/LER-2016-004 |
| ENS 52174 | 11 August 2016 19:05:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Robinson Dam Tainter Gates Unanalyzed Condition | During 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. | 05000261/LER-2016-003 | |
| ENS 51864 | 13 April 2016 18:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition | At 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. | Steam Generator Feedwater Main Steam Line | |
| ENS 50778 | 29 January 2015 00:57:00 | Other Unspec Reqmnt 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Safety System Will Not Function as Required | At 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. | Reactor Protection System | |
| ENS 50178 | 6 June 2014 20:22:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Notification of Unusual Event Due to a Fire Alarm Inside Containment | 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.
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 49708 | 10 January 2014 03:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | 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 | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Steam Safety Valve Control Rod | |
| ENS 49506 | 5 November 2013 23:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Momentary Loss of 4 Kv Buses and Safety System Actuation | 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. | Steam Generator Feedwater Service water Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Condenser | |
| ENS 49502 | 5 November 2013 05:41:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of the Auxiliary Feedwater System Due to Main Feed Pump Trip | At 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. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater | 05000261/LER-2013-002 |
| ENS 49469 | 23 October 2013 18:05:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification of a Fatality Onsite | This 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 49458 | 21 October 2013 22:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification Due to Release of Petroleum from an Unregulated Source | This 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. | Emergency Diesel Generator | |
| ENS 48566 | 7 December 2012 14:20:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Emergency Offsite Facility/Technical Support Center Ventilation Maintenance | 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 48558 | 5 December 2012 05:31:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Radiation Monitor Out of Service for Planned Maintenance | 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.
Radiation Monitor R-9 was returned to service at 2022 EST. The licensee informed the NRC Resident Inspector. Notified R2DO (Blamey). | ||
| ENS 48508 | 14 November 2012 14:27:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Radiation Monitor Out of Service | 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.
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 48354 | 28 September 2012 12:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Planned Emergency Offsite Facility/Technical Support Center Ventilation Maintenance | 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.
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 48288 | 8 September 2012 12:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Emergency Offsite Facility/Technical Support Center Ventilation Maintenance | 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.
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). | HVAC | |
| ENS 48199 | 17 August 2012 09:52:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to the Unplanned Loss of Annunciators > 15 Minutes | 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).
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 48180 | 11 August 2012 11:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Technical Support Center Ventilation Out of Service for Scheduled Maintenance | At 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. | HVAC | |
| ENS 47828 | 12 April 2012 13:30:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Tsc and Eof Out of Service Due to Maintenance | 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.
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 47781 | 28 March 2012 19:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Steam Generator High Water Level Condition | 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. | Steam Generator Feedwater Auxiliary Feedwater Main Steam Safety Valve Control Rod | |
| ENS 47748 | 16 March 2012 08:00:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Refueling Water Storage Tank Connected to an Unqualified System | The 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 47663 | 14 February 2012 17:40:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Response Facility Information System Due to Planned Modification | 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.
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). | Emergency Response Data System Safety Parameter Display System | |
| ENS 47618 | 23 January 2012 23:04:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unsual Event Declared Due to Fire Inside Containment Lasting > 15 Minutes | At 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 47600 | 17 January 2012 23:14:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Unit 2 Commenced a Tech Spec Required Shutdown After Declaring Station Battery "B" Inoperable | At 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 47551 | 23 December 2011 21:20:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Alert Declared Due to Halon Discharge During Testing | 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 47501 | 6 December 2011 14:00:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Emergency Assessment Capability | 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.
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. | Steam Generator | |
| ENS 47302 | 29 September 2011 10:02:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to Fire Alarm in the Containment | 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.
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 47293 | 26 September 2011 15:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to One Loop Low Flow Signal | 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. | Steam Generator Feedwater Auxiliary Feedwater Main Steam Safety Valve Control Rod | |
| ENS 46828 | 7 May 2011 12:27:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Emergency Data Systems Out of Service Due to Planned Modification | 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.
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) | Emergency Response Data System Safety Parameter Display System | |
| ENS 48011 | 4 May 2011 17:25:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Refueling Water Storage Tank Connected to Non-Seismically Qualified System | On 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. | Emergency Core Cooling System | |
| ENS 46317 | 7 October 2010 17:15:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Feedwater Isolation Inadvertently Disabled | At 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. | Steam Generator Feedwater | |
| ENS 46313 | 7 October 2010 04:13:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to a One Loop Reactor Coolant Low Flow | 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. | Steam Generator Feedwater Auxiliary Feedwater Main Turbine Main Steam Safety Valve Control Rod | 05000261/LER-2010-009 |
| ENS 46238 | 9 September 2010 18:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | 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. | Steam Generator Feedwater Auxiliary Feedwater Main Steam Safety Valve Control Rod | |
| ENS 46204 | 25 August 2010 12:45:00 | 10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness | Loss of Several Display Systems Due to Planned Computer Modifications | 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.
At 1430, the ERFIS system was returned to service. The NRC Resident Inspector has been notified. Notified the R2DO(Franke) | Emergency Response Data System Safety Parameter Display System | |
| ENS 46045 | 24 June 2010 18:34:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Instrument Buses 3 and 8 Failed Causing Closure of Rhr Valves | 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.
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). | Reactor Coolant System Service water Emergency Diesel Generator Shutdown Cooling Residual Heat Removal | 05000261/LER-2010-005 |
| ENS 46127 | 25 May 2010 17:14:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Report for Inadvertent Safety Injection System Actuation | At 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. | Emergency Diesel Generator | |
| ENS 45799 | 28 March 2010 22:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Alert Declared Due to Fire Affecting Safety Related Equipment | 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). | Steam Generator Reactor Coolant System Reactor Protection System Emergency Diesel Generator Main Steam Isolation Valve Auxiliary Feedwater Decay Heat Removal Control Rod | |
| ENS 45483 | 7 November 2009 03:02:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Feed Reg Valve Closure | 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. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 44660 | 17 November 2008 10:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to High Turbine Vibration | 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. | Steam Generator Feedwater Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Main Turbine Decay Heat Removal Main Steam | |
| ENS 43688 | 3 October 2007 22:30:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release | Offsite Notification to South Carolina Osha Due to Non-Work Related Fatality Onsite | On 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. |