Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56351 | 14 February 2023 17:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Two Reactor Coolant Pumps Tripping | The following information was provided by the licensee via email: On February 14, 2023 at 1103 CST, Arkansas Nuclear One, Unit 1, (ANO-1) automatically tripped on reactor protection system actuation due to two reactor coolant pumps tripping. ANO-1 is currently stable in MODE 3 (Hot Standby) maintaining reactor coolant system pressure and temperature with main feedwater and steaming to the condenser. No additional safety system actuations occurred. All immediate actions were completed satisfactorily. There are no indications of a radiological release on either unit as a result of this event. This report satisfies the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A) for the reactor protection system actuation. The NRC Senior Resident Inspector has been notified. Unit 2 was not affected. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee is investigating the cause of the two reactor coolant pump trips. | Reactor Coolant System Feedwater Reactor Protection System | |
| ENS 55614 | 29 November 2021 20:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | On November 29, 2021 at 1458 CST, Arkansas Nuclear One, Unit 1, (ANO-1) automatically tripped due to high Reactor Coolant System pressure after the 'A' Main Feedwater Pump was manually tripped due to lowering speed. ANO-1 is currently stable in MODE 3 (Hot Standby) maintaining pressure and temperature with the P-75 Auxiliary Feedwater pump and steaming to the Condenser. There are no indications of a radiological release on either unit as a result of this event. This report satisfies the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A) for the Reactor Protection System actuation. The NRC Senior Resident Inspector has been notified. Unit 2 was not affected. | Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 55138 | 14 March 2021 19:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Trip Due to Loss Bus Voltage | On March 14, 2021, at 1315 CDT, Arkansas Nuclear One, Unit 1(ANO-1) was manually tripped due to degraded voltage and momentary loss of the A-2, non-vital 4160 V Bus in accordance with Abnormal Operating Procedure. All control rods fully inserted. Degraded voltage of the A-2 non-vital 4160 V Bus resulted in de-energizing the A-4 vital 4160 V Bus. Emergency Diesel Generator No. 2, K-4B, started automatically and is currently powering the A-4 vital 4160 V Bus. All other Vital and Non-Vital Buses transferred power automatically to the Startup Transformer No. 1. Offsite power remains energized and available for ANO-1. All other systems responded as designed. The loss of the A-2 Non-Vital Bus is still under investigation. ANO-1 is currently stable in MODE 3 (Hot Standby), maintaining pressure and temperature with Main Feedwater pumps and steaming to the Condenser. There are no indications of a radiological release on either unit as a result of this event. This report satisfies the reporting criteria of both 10 CFR 50.72(b)(2)(iv)(B) for the Reactor Protection System actuation and 10 CFR 50.72(b)(3)(iv)(A) for the actuation of the Emergency Diesel Generator. The Licensee has notified the NRC Senior Resident Inspector. | Feedwater Reactor Protection System Emergency Diesel Generator Control Rod | |
| ENS 55028 | 10 December 2020 22:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Low Steam Generator Water Level | On December 10, 2020 at 1608 CST, Arkansas Nuclear One, Unit 2 (ANO-2) experienced an automatic reactor scram from 100 percent power due to Low Steam Generator Water Level in 2E-24A Steam Generator. Emergency Feedwater actuated automatically due to low water level in the A Steam Generator. Due to inadequate control of the B Main Feedwater Control System, water level in the B Steam generator rose to a level requiring manual trip of the B Main Feedwater pump. Emergency Feedwater responded as designed to feed both steam generators automatically. All other systems responded as designed. All electrical power is being supplied from offsite power and maintaining unit electrical loads as designed. Unit 2 is currently stable in Mode 3 (Hot Standby) maintaining pressure and temperature via Emergency Feedwater and secondary system steaming. There are no indications of a radiological release on either unit as a result of this event. This report satisfies the reporting criteria of both 10 CFR 50.72(b)(2)(iv)(6) for the Reactor Protection System actuation and 10 CFR 50.72(b)(3)(iv)(A) for the actuation of the Emergency Feedwater System. The Arkansas Nuclear One NRC Senior Resident Inspector has been notified.
The purpose of this (report) is to provide an update to NRC Event Number 55028. The cause of the inadequate control of the B Main Feedwater Control System to control B Steam Generator Level was verified to be associated with the failure that led to the A Steam Generator low level condition. After taking action to trip the B Main Feedwater Pump, Emergency Feedwater was manually actuated for the B Steam Generator and the Emergency Feedwater System was verified to maintain proper automatic control of both Steam Generator levels. At the time of the initial event notification, plant temperature and pressure control had been transferred from Emergency Feedwater to Auxiliary Feedwater along with secondary system steaming. The licensee notified the NRC Resident Inspector. Notified R4DO (Kellar). | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 54091 | 26 May 2019 10:12:00 | 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 Reactor Coolant Pump Trip on Ground Fault | This is a 4-hour Non-Emergency 10 CFR 50.72(b)(2)(iv)(B) notification due to a Plant Protection System (PPS) actuation. Arkansas Nuclear One, Unit 2, automatically tripped from 100 percent power at 0512 CDT. The reactor automatically tripped due to 2P-32B Reactor Coolant Pump tripping as a result of grounding. No additional equipment issues were noted. All control rods fully inserted. Emergency Feedwater (EFW) actuated and was utilized to maintain Steam Generator (SG) levels. The EFW actuation meets the 8-hour Non-Emergency Immediate Notification Criteria of 10 CFR 50.72(b)(3)(iv)(A). No Primary safety valves lifted. Main Steam Safety Valves (MSSVs) did lift initially after the trip. The NRC Resident Inspector has been notified. Decay heat is being removed via the steam dump valves to the main condenser. Unit 2 is in a normal shutdown electrical lineup. Unit 1 was not affected by the transient on Unit 2. The licensee notified the State of Arkansas. | Steam Generator Feedwater Main Steam Safety Valve Main Condenser Control Rod | |
| ENS 53793 | 18 December 2018 06:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip - Loss of Non-Vital Bus | On December 18, 2018 at 1126 CST, Arkansas Nuclear One, Unit 1 (ANO-1) reactor automatically tripped due to a loss of the A-1, non-vital 4160V bus. All control rods fully inserted. Loss of the A-1 bus resulted in de-energizing A-3 vital 4160V bus. Emergency Diesel Generator #1, K-4A, started automatically and is currently powering A-3 vital bus. Non-vital buses A-2, H-1, and H-2 and vital bus A-4 transferred power automatically to the Startup Transformer #1. Off-site power remains energized and available for ANO-1. The reason for loss of A-1 bus is unknown at this time. Currently, ANO-1 has stabilized in Mode 3, Hot Standby. Decay heat is being removed by the main condenser using the turbine bypass valves. The loss of the A-1, non-vital bus, is under investigation. The licensee has notified the NRC Resident Inspector and the state. | Emergency Diesel Generator Main Condenser Control Rod | |
| ENS 53459 | 16 June 2018 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip During Startup | At 1121 CDT on June 16, 2018, Arkansas Nuclear One, Unit 1 (ANO-1) performed a manual reactor trip due to a Turbine Bypass valve failing open on reactor startup. At the time, ANO-1 was in Mode 2 at approximately 2 percent power. The failed Turbine Bypass valve resulted in an overcooling event and the Overcooling Emergency Operating Procedure (EOP) was entered. Main Steam Line Isolation (MSLI) automatic actuation occurred on 2 of the 4 channels of Emergency Feedwater Initiation and Control during the overcooling event in the 'B' Steam Generator. The remaining channels of MSLI were manually actuated by the control room staff from the control room. Overcooling was terminated after the closure of the Main Steam Isolation Valve (MSIV) and reactor coolant parameters were stabilized as directed by the Overcooling EOP. Additionally, Gland Sealing Steam was lost to the main turbine due to the closure of the 'B' Steam Generator MSIV and Loss of Condenser Vacuum Abnormal Operating Procedure was entered. This is a 4-hour non-emergency 10 CFR 50.72 (b)(2)(iv)(B) notification due to a Reactor Protection System actuation (scram) and an 8-hour non-emergency 10 CFR 50.72 (b)(3)(iv)(A) notification for safety system actuation." All control rods fully inserted into the core during the trip. Heat removal is via the Atmospheric Dump Control valves to atmosphere. The NRC Resident Inspector has been notified. The licensee also notified the State of Arkansas. | Steam Generator Reactor Protection System Main Steam Isolation Valve Main Turbine Emergency Feedwater Initiation and Control Main Steam Line Control Rod | |
| ENS 53404 | 16 May 2018 22:50:00 | 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 Loss of Main Feedwater Pump | At 1750 CDT, the Arkansas Nuclear One, Unit 1 (ANO-1) reactor tripped due to the trip of the 'B' Main Feedwater Pump. Unit 1 was at 10 percent power with escalation of power in progress with one Main Feedwater Pump in service. Investigation is in progress as to the cause of the Main Feedwater Pump trip. The Main Feedwater Pump trip resulted in RPS (reactor protection system) actuation on loss of both Main Feedwater Pumps and resulted in Emergency Feedwater (EFW) actuation. All Control Rods inserted into the core properly and the reactor was verified shutdown. EFW experienced a half-trip on the 'A' train of Emergency Feedwater Initiation and Control (EFIC) at time of system actuation, but was successfully actuated manually immediately upon discovery. Train 'B' EFIC actuated in Automatic as designed. The half-trip of the 'A' train of EFIC is currently believed to be associated with EFIC Channel 'C'; however, investigation is underway to verify this. Currently, ANO-1 has been stabilized and is being maintained in Mode 3 with Auxiliary Feedwater in service. Heat removal is via Turbine Bypass valves to the Condenser. No radiological releases have occurred due to this event. There was no effect on Arkansas Nuclear One, Unit 2. The licensee notified the NRC Resident Inspector and the State of Arkansas. | Feedwater Auxiliary Feedwater Emergency Feedwater Initiation and Control Control Rod | |
| ENS 52710 | 26 April 2017 15:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Automatic Reactor Scram Due to Partial Loss of Offsite Power | At 1004 CDT, Arkansas Nuclear One, Unit 1 (ANO-1) reactor automatically tripped due to the partial loss of offsite power. At the time of the trip, the site was in a Tornado Warning and a Severe Thunderstorm Warning. The Emergency Feedwater (EFW) system auto-actuated due to the loss of main feedwater pumps and the loss of the Reactor Coolant pumps. Both Emergency Diesel Generators started as expected with only one loading as expected. All control rods fully inserted. Currently, ANO-1 has stabilized in Hot Standby via natural circulation. ANO-1 also lost Spent Fuel Pool cooling for approximately 69 minutes. The temperature of the spent fuel pool at the beginning of the event was approximately 102 (degrees) F. The spent fuel pool saw a heatup of 1 (degree) F during the loss of spent fuel pool cooling. The Spent Fuel Pool cooling has been restored. ANO-2 is currently in a refueling outage with all fuel in the spent fuel pool. ANO-2 completed a full core off load to the spent fuel pool and this was completed on April 12, 2017. Spent Fuel Pool cooling was lost for approximately 10 minutes. The Spent Fuel Pool temperature was 91 (degrees) F prior to the event. No heat up of the pool was identified during the event. Cooling has subsequently been restored. The #1 Emergency Diesel Generator auto-started as designed but did not supply the safety bus due to availability of offsite power. No radiological releases have occurred from either unit due to this event. The licensee has notified the NRC Resident Inspector. | Feedwater Emergency Diesel Generator Control Rod | 05000313/LER-2017-001 |
| ENS 51607 | 15 December 2015 11:44:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 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 Loss of Feed to One Steam Generator | This is a 4-hour Non-Emergency 10 CFR 50.72(b)(2)(iv)(B) notification due to a Reactor Protection System (RPS) actuation. Arkansas Nuclear One, Unit 1, was manually tripped from 43 percent power at 0544 CST. The reactor was manually tripped due to operator judgement during control issues with the Integrated Control System (ICS) during a planned downpower for Electro-Hydraulic Control (EHC) system maintenance. CV-2672 B, low load control valve, failed to close. Subsequently, CV-2674 B, low load block valve, began to close and caused a loss of feed to E-24B Steam Generator. No additional equipment issues were noted. All control rods fully inserted. Emergency Feedwater (EFW) actuated and was utilized to maintain Steam Generator (SG) levels. This (EFW actuation) meets the 8 hour Non-Emergency Immediate Notification Criteria ((10CFR50.72(b)(3)(iv)(A)). No Primary safety valves lifted. Main Steam Safety Valves (MSSVs) did lift initially after the trip. The NRC Resident (Inspector) has been notified. Decay heat is being removed via the steam dump valves to the main condenser. Unit 1 is in a normal shutdown electrical lineup. Unit 2 was not affected by the transient on Unit 1. The licensee notified the State of Arkansas. | Steam Generator Feedwater Reactor Protection System Main Steam Safety Valve Main Condenser Control Rod | |
| ENS 50063 | 28 April 2014 01:12:00 | 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 an Auxiliary Trip on Core Protection Calculator | At approximately 1932 (CDT) on 4/27/2014, the System Operations Center (SOC - Dispatcher) informed Unit 2 of a system wide grid emergency and ordered both Unit 1 and Unit 2 to come off line as soon as possible. At approximately 2012 (CDT), Unit 2 automatically tripped from 51% power due to an Auxiliary Trip on CPCs (Core Protection Calculator) due to Axial Shape Index (ASI) trip. All Control Element Assemblies inserted into the core. Both vital and non-vital 4160V and 6900V buses remain powered from Startup #3 Transformer. All Systems responded as designed. At 1932 (CDT), Unit 1 commenced a Rapid Plant Shutdown at a rate 5-7% per min with the intention to take the turbine offline and leave the reactor critical at 10-12% power on the Turbine Bypass Valves. When the Unit 2 reactor tripped, Unit 1 stopped the power reduction and stabilized the plant at approximately 19% Reactor Power and 125 Generated Megawatts. With SOC concurrence, Unit 1 stabilized power and was told to limit site output to <200 MWe. At 1932 CDT, Unit 1 began a down power from 100% power and Unit 2 began a down power from 95% power. On Unit 2, decay heat is being removed by the main condenser using the turbine bypass valves. Unit 2 is stable in Mode 3 with stable offsite power. The system wide grid emergency is believed to be caused by tornados in the region. The licensee has notified the NRC Resident Inspector and the State. | Main Condenser | |
| ENS 49995 | 3 April 2014 18:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At approximately 1301 (CDT) on 4/3/2014, Unit 2 tripped from 100% power for unknown reason(s). All Control Element Assemblies (CEA) inserted into the core. (Note that CEA 03 Plant Monitoring System indication still indicates fully withdrawn. However, CEA 03 in-limit light and control panel indication validate that CEA 03 is fully inserted.) 4160v bus 2A1 and 6900v bus 2H1 transferred to Startup Transformer #2. 4160v bus 2A2 appeared to de-energize and re-energize on Startup Transformer #3. 6900v bus 2H2 is de-energized. 2K4B emergency diesel generator started but did not tie onto the 2A4 4160v bus. Emergency Feedwater actuated on low steam generator level. Offsite power remains available and decay heat is being removed to the main condensers using the turbine bypass valves. Unit 1 was not affected. The licensee has notified the NRC Resident Inspector and will notify the State. | Steam Generator Feedwater Emergency Diesel Generator Main Condenser | |
| ENS 49618 | 9 December 2013 14:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 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 Unit 2 Auxiliary Transformer Explosion | At approximately 0748 (CST) on 12/9/2013, an electrical fault occurred resulting in a fire and explosion on the ANO (Arkansas Nuclear One) Unit 2 Unit Auxiliary Transformer. This caused a unit trip and a loss of power to Startup 3 Transformer, which is one of the two offsite power feeds to ANO Unit 2. ANO Unit 2 is currently in a stable shutdown condition. With Startup 3 and Unit Aux Transformer unavailable, power was lost to the Reactor Coolant Pumps and Circulating Water Pumps. RCS (Reactor Coolant System) natural circulation is in progress removing core decay heat. Emergency Feedwater actuated due to low steam generator levels and is supplying both steam generators. The unit is steaming to the atmosphere. 2A-1 and 2A-3 are powered from SU (Startup) 2 Transformer. 2A-4 is powered from 2K-4B Emergency Diesel Generator. ANO Unit 1 is currently operating at 98% power. The auto transformer tripped off line with the fault in ANO Unit 2 Unit Auxiliary Transformer. This has caused Startup Transformer 1 to be inoperable. This places ANO Unit 1 in a 72 hour Technical Specification action statement (T.S 3.6.1 for Loss of the SU-1 Transformer). No significant injuries were reported as a result of this condition and offsite agencies have been notified. At 0800 on 12/9/2013 Unit 2 declared a Notification of Unusual Event based on EAL HU4 Fire or Explosion inside Protected Area not extinguished in 15 minutes. At 0917 on 12/9/2013, the fire on Unit 2 Unit Auxiliary Transformer is OUT. The Licensee notified the NRC Resident Inspector. Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and Nuclear SSA vial email.
At 1215 CST, ANO terminated from the Unusual Event as per HU4 stable plant conditions. ANO Unit 2 is stable and will be cooled down to Mode 5 via natural circulation and decay heat is being removed via emergency feedwater and downstream dump valves to atmosphere. There is no radiological release and no personnel injuries. The licensee has notified the NRC Resident Inspector, Arkansas Department of Emergency Management, Arkansas Department of Health, and other local authorities. Notified R4DO (Vasquez), NRR EO (King) and IRC MOC (Morris). Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and Nuclear SSA vial email.
Outside agencies (National Response Center, Arkansas Department of Emergency Management, Corps of Engineers, and U.S. Coast Guard) were contacted due to a minor unknown amount of oil that entered the plant discharge to Lake Dardanelle. The oil was from the faulted Unit 2 Unit Auxiliary Transformer. The majority of oil was contained within the containment around the transformer or the oily water separator it drains to. Local inspection revealed only a small amount of what was released actually passed the containment booms that are continuously in place on the discharge canal. The oil boom was verified to be in good condition. The oil was verified to be mineral oil. The release was verified to be terminated. An additional oil boom was deployed. Reference (ANO) Condition Report CR-ANO-C-2013-03071. The licensee notified the NRC Resident Inspector. Notified the R4DO (Werner) | Steam Generator Feedwater Emergency Diesel Generator | |
| ENS 48869 | 31 March 2013 15:33:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 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 | Notification of Unusual Event Declared Due to a Breaker Explosion in the Protected Area | At 0750 (CDT) on 3/31/2013, during movement of the Unit 1 Main Turbine Generator Stator (~500 tons), the Unit 1 turbine temporary lift device failed. This caused a loss of all off-site power on Unit 1. The ANO Unit 1 #1 and #2 EDG (Emergency Diesel Generator) have started and are supplying A-3 4160V switchgear and A-4 4160V switchgear. P-4A Service Water pump and P-4C Service Water pump has been verified running. Unit 1 has entered (procedures) 1202.007 - Degraded Power, 1203.028 - Loss of Decay Heat, and 1203.050 - Spent Fuel Emergencies. Unit 1 is in MODE 6. ANO-1 entered TS 3.8.2 A, 'One Required Offsite Circuit Inoperable'. All required actions are complete. The event caused a loss of decay heat removal on ANO Unit 1 which was restored in 3 minutes and 50 seconds. Unit 2 tripped and is in MODE 3. Emergency Feed Water was initiated on Unit 2 and Unit 2 was in (Technical Specification) 3.0.3 from 0817 (CDT) to 0848 (CDT) due to Emergency Feedwater. Unit 2 is being powered by off-site. Unit 2 Startup 3 (transformer) lock out at 0921 (CDT). (Bus) 2A1 is on Start up 2 (transformer) and (bus) 2A3 is on #2 EDG. 10CFR50.72 (b)(3)(iv)(A) - 4-hr. notification due to the ES (Engineered Safeguard Feature) actuation on both Unit 1 and Unit 2. 10CFR50 72 (b)(2)(iv)(B) - 4-hr. notification due to RPS (Reactor Protection System) actuation on Unit 2. 10CFR50.72 (b)(2)(xi) - 4-hr. notification due to Government Notification. 29CFR1904.39a - (OSHA) 8-hr. notification due to death on site. At 1033 (CDT) on 3/31/2013, Unit 2 entered a Notification of Unusual Event based on EAL HU4 due to damage in 2A1 switchgear. Notification of the NUE will be made lAW Emergency Plan requirements. Follow-up notifications will be made as appropriate. At this time, the full extent of structural damage on Unit 1 is not known. There was one known fatality and 4 known serious injuries to workers. The local coroner is on site for the fatality and the injured personnel have been transported offsite to local hospitals. Investigation into the cause of the failure and extent of damage is ongoing. On Unit 2, all rods inserted during the trip. The core is being cooled via natural circulation. Decay heat is being removed via steam dumps to atmosphere. There is no known primary to secondary leakage. The licensee has notified the State of Arkansas, local authorities, OSHA and the NRC Resident Inspector. Notified DHS SWO, DHS NICC, FEMA and Nuclear NSSA (via email).
The licensee terminated the NOUE at 1821 CDT. The basis for termination was that the affected bus (2A2) is de-energized and no other equipment on Unit 2 was damaged. The licensee has notified the state and local authorities and will be notifying the NRC Resident Inspector. Notified R4DO (Pick), NRR EO (Howe), IRD (Gott), DHS SWO, DHS NICC, FEMA and Nuclear SSA (via email).
The licensee made the following edits to the third paragraph of their original report (edits in quotes): Unit 2 tripped and is in MODE 3. Emergency Feed Water initiated on Unit 2. Unit 2 was in (Technical Specification) 3.0.3 from 0817 (CDT) to 0848 (CDT) due to Emergency Feedwater "being procedurally overridden." Unit 2 "was initially" being powered by off-site. Unit 2 Startup 3 Lock out occurred at 0921. 2A1 is now on Startup 2, and "2A4" is on #2 EDG. Notified R4DO (Kellar) via email. | Feedwater Service water Main Turbine Decay Heat Removal | |
| ENS 48169 | 8 August 2012 13:23:00 | 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 High Reactor Coolant System Pressurizer Pressure | At 0823 hours on August 8, 2012, Arkansas Nuclear One, Unit 2 (ANO-2) experienced an automatic reactor trip. The reactor automatically tripped due to High Reactor Coolant System Pressurizer Pressure that was caused by a Main Turbine trip due to high condenser back pressure from a degraded vacuum condition. The Reactor Protection System (RPS) performed as designed in response to the High Reactor Coolant System Pressurizer Pressure condition resulting in automatic shutdown of the reactor from approximately 100 percent power. All Control Element Assemblies (CEAs) fully inserted on the trip. The Emergency Feedwater Actuation System (EFAS) actuated for the 'A' Steam Generator only due to level trending slightly below the setpoint. The plant has transitioned to supplying the steam generators using the Auxiliary Feedwater (AFW) system. The unit is currently in Mode 3 and implementing the transient response process. The investigation into the cause of the trip is ongoing and the local NRC Resident Inspectors have been notified. The unit is in a normal electrical lineup, and the decay heat is being removed by the main condenser via the turbine bypass valves. The State Department of Health was notified and ANO-2 will be issuing a press release. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 45872 | 26 April 2010 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit 1 Experienced an Automatic Reactor Trip During an Ni Calibration | This is a 4 hour Non-Emergency 10CFR 50.72(b)(2)(iv)(B) notification due to an Automatic Reactor Protection System (RPS) actuation (scram). At 2126 hours (CDT) on April 25, 2010, Unit 1 Reactor automatically tripped due to 2 of 4 Reactor Protection System (RPS) Channels tripped. At the time of the trip, reactor power, as indicated by heat balance, was ~20%, while excore Nuclear Instrumentation (NI) indicated ~30%. The RPS high reactor power trip setpoint was 50% power. An NI calibration initiated an automatic withdrawal command to the control rod drive system. The rod withdrawal, resulted in one RPS channel tripping on high reactor power and another RPS channel tripping on high reactor coolant system pressure. All control rods fully inserted into the core and no safety systems, other than RPS, actuated. Emergency feedwater did not actuate and was not needed. No primary safety valves lifted. Seven secondary safety valves lifted and subsequently reseated. The plant is currently stable in Mode 3. The NRC resident has been notified. The licensee also informed the State of Arkansas and does not plan a press release. The Unit 1 reactor trip was uncomplicated. Current means of decay heat removal is normal feedwater to the Steam Generators with steam discharge to the Main Condenser through Main Steam Bypass. The Main Generator was online at the time of the trip and the plant is currently in a normal post trip electrical line up. There is no indication of primary-secondary tube leakage. All systems functioned as required. | Steam Generator Reactor Coolant System Feedwater Reactor Protection System Decay Heat Removal Main Condenser Control Rod Main Steam | |
| ENS 45854 | 18 April 2010 18:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip During Reactor Startup | This is a 4-hour Non-Emergency, 10CFR 50.72(b)(2)(iv)(B) notification due to an RPS actuation (scram). Arkansas Nuclear One Unit 1 was manually tripped from 11% power at 1357 hrs. CDT. The cause of the trip was operator judgment due to a small fire reported in the high pressure turbine enclosure at governor valve number 3 by a fire watch stationed at that location and an unrelated failure of P-32C Reactor Coolant Pump 3rd stage seal (upper seal) occurring earlier that afternoon. No additional equipment issues were noted. An extinguishing agent (CO2) was applied within approximately 30 seconds. All control rods fully inserted. No primary safeties lifted. No secondary safeties lifted. Emergency feedwater did not actuate and was not needed. No safety systems actuated. The plant will be cooled down to repair the P-32C Reactor Coolant Pump Seal. The NRC Resident has been notified. There was no effect on Unit 2. The grid is stable with Unit 1 in a normal shutdown electrical lineup. Decay heat is being removed via the steam dumps to condenser using normal feed to the steam generators. The licensee will be notifying the Arkansas Department of Health. | Steam Generator Feedwater Control Rod | |
| ENS 45549 | 8 December 2009 14:42:00 | 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 Low Steam Generator Level Caused by Loss of Main Feed Water Pump | Arkansas Nuclear One - Unit 2 experienced a high temperature on the 'A' Main Feedwater Pump thrust bearing which required the pump to be manually tripped . Steam Generator levels lowered as a result of the Main Feedwater Pump trip to the point that operators initiated a manual reactor trip. The Emergency Feedwater System automatically actuated on low Steam Generator level as a result of the Steam Generator level transient. The manual reactor trip requires 4-Hr non-emergency notification IAW 10CFR 50.72(b)(2)(iv)(B). The automatic actuation of Emergency Feedwater requires 8-Hr non-emergency notification IAW 10CFR 50.72(b)(3)(iv)(A). All rods fully inserted. After the trip, decay heat was being removed using steam dumps to the condenser. Steam generator level was being maintained with the emergency feedwater pumps. The licensee will notify the NRC Resident Inspector. | Steam Generator Feedwater | |
| ENS 45367 | 20 September 2009 10:11:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Emergency Diesel Auto Start During Surveillance Testing | This report is being made due to an auto start of the Red Train Emergency Diesel Generator (2K-4A) which was the result of an unplanned loss of power to the Red Train 4160 Volt Electrical Bus (2A3) during surveillance testing. Concurrent with the loss of power, Shutdown Cooling Flow was temporarily lost. This condition occurred with the unit in Mode 5 during a refueling outage. The Reactor Coolant System (RCS) Pressurizer level was 85 percent and RCS temperature was 139 degrees Fahrenheit. During the momentary loss of power, 2A3 automatically shed its loads as designed. This caused the running Shutdown Cooling Pump (Low Pressure Safety Injection Pump 2P-60A) to secure which resulted in a loss of Shutdown Cooling Flow to the RCS for approximately three and a half minutes. Shutdown Cooling was restored using the applicable Abnormal Operating Procedure. RCS temperature rose approximately five degrees Fahrenheit. During the performance of planned surveillance testing, 2K-4A was unexpectedly auto started. An Offsite Power Transfer Test was being performed to test automatic transfer from Startup 3 Offsite Transformer to Startup 2 Offsite Transformer. The transfer was initiated by momentarily bypassing (jumpering) a relay. When the test was initiated, a slow transfer of the Red Train 4160 Volt Electrical Bus (2A1) occurred instead of the expected fast transfer. The slow transfer of 2A1 resulted in a momentary loss of power, for approximately two seconds, to the Red Train 4160 Volt Electrical Bus (2A3) which is supplied from 2A1. The under voltage condition on 2A3 caused 2K-4A to auto start, as designed. 2K-4A did not power 2A3, since 2A3 was powered from 2A1 after the slow transfer completed. The cause of the slow transfer versus a fast transfer is under investigation. The licensee notified the Arkansas Department of Health and the NRC Resident Inspector. | Reactor Coolant System Emergency Diesel Generator Shutdown Cooling | |
| ENS 44906 | 14 March 2009 02:51:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Feedwater Regulating Valve Failure Leads to Manual Reactor Trip | At 2151 CDT on 3/13/2009 with plant power at 84% following securing of the 2P-1A Main Feedwater Pump at 1934 due to bearing degradation, Unit 2 experienced a failure of the 'B' Main Feedwater Regulating Valve 2CV-0740 which caused 2E-24B Steam Generator level to lower. At 2151, with 2E-24B Steam Generator level at approximately 25% control room staff manually tripped the reactor anticipating an automatic trip at 22% steam generator level. Emergency Feedwater Actuation System automatically actuated at 22% steam generator level and responded to restore level to the normal post-trip levels. All Control Element Assemblies (rods) Inserted on the reactor trip. No safeties or relief valves opened after the trip. The plant is in its normal shutdown electrical line-up. Steam generator level is being maintained with emergency feedwater and decay heat is being dumped to the main condenser. The licensee notified the State of Arkansas and the NRC Resident Inspector. | Steam Generator Feedwater Main Condenser | |
| ENS 44837 | 7 February 2009 16:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration | Unusual Event Declared Due to a Fire Onsite | Arkansas Nuclear One Unit 1 declared an Unusual Event per EAL 7.5 at 1059 CST based on a fire onsite. The fire occurred on the 354 foot elevation iso-phase bus deck of the turbine building. The fire was attributed to a failed bonnet on a manual valve which had unthreaded during operation releasing hydrogen in the vicinity of the hydrogen add station. The licensee extinguished the fire using water and posted a reflash watch. There were no reported personnel injuries. Offsite assistance was requested from the London Fire Department which arrived onsite after the fire had been extinguished. The reactor was manually tripped with all control rods fully inserting. The plant is currently stable with decay heat removal via the steam dumps to the main condenser. The licensee informed state/local agencies and the NRC Resident Inspector.
The Unusual Event was terminated at 1338 CST after confirmation that the fire was extinguished and a review of the exit criteria as defined in plant procedures. The licensee is performing a damage assessment at this time. The licensee was in the process of informing state/local agencies and will inform the NRC Resident Inspector. Notified R4DO (Clark), NRR(Hiland), IRD (Grant), DHS (Banner) and FEMA (Canupp).
The subject event notification erroneously reported that all control rods fully inserted following the manual reactor trip initiated in response to a fire on site. In actuality, the rod bottom light for Rod 6 in Rod Group 7 on the CRD Position Indication Panel did not illuminate following the reactor trip and it was verified by plant computer to have inserted to 2.3 percent withdrawn. The control board operator identified this condition quickly and reported it to the Control Room Supervisor in his immediate action response. This condition did not necessitate any additional operator actions and shutdown margin requirements were maintained. The licensee has informed the NRC Resident Inspector. R4DO (Miller) notified. | Decay Heat Removal Main Condenser Control Rod | |
| ENS 44831 | 5 February 2009 21:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit 1 Manually Tripped Due to Control Rod Drive Motor High Temperatures | Arkansas Nuclear One Unit 1 was manually tripped from 61% power due to Control Rod Drive motor (CRD) high temperatures due to a loss of cooling water flow from Non-Nuclear Intermediate Cooling Water (ICW). No significant equipment issues were observed. No primary safeties lifted. Secondary safeties lifted for ~18 seconds. Emergency feedwater did not actuate and was not needed. Post trip response was normal and the plant is stable in mode 3. Post transient review is in progress. All control rods fully inserted into the core. There is no known primary-secondary steam generator tube leakage. The offsite electrical grid is stable. The loss of ICW flow is attributed to the mechanical failure of service air compressor C-3A. Decay heat removal is via main feedwater to the steam generator and steam to the main condenser. The licensee informed state, other government agencies and the NRC Resident Inspector and will inform local government agencies. No press release is planned. | Steam Generator Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 44736 | 20 December 2008 18:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip When Control Rod Bank Inserted Without Operator Action | Arkansas Nuclear one Unit 1 manually tripped from 100% power due to control rod mechanism group 7 unexpectedly dropping from 90% to 70%. No Primary safeties lifted. Secondary safeties lifted for approximately 1 minute. Emergency feedwater did not actuate and was not needed. Post trip response was normal and the plant is stable in mode 3. Post transient review is in progress. All control rods fully inserted on the trip. The plant is in its normal shutdown electrical lineup. Steam generator level is being maintained using aux feed and steam is being dumped to the main condenser through the turbine bypass valves. There is no primary to secondary leakage. The licensee notified the NRC Resident Inspector, the State of Arkansas, and the Arkansas Department of Health. See similar event number 44716. | Steam Generator Feedwater Main Condenser Control Rod | |
| ENS 44716 | 12 December 2008 14:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit 1 Was Manually Tripped from 32% Power Due to an Abnormal Rod Pattern | Following startup from refueling outage 1R21, Arkansas Nuclear One Unit 1 was holding stable at 32% reactor power in order to perform Nuclear Instrumentation (NI) calibration. Control Rod Drive Mechanism (CRDM) control was in automatic with the controlling group (Group 7) at approximately 75% withdrawn. Control room operators received an asymmetric rod alarm and noted abnormal rod pattern on Group 7 with reactor power lowering. At this point, operations manually scrammed the reactor. Post trip response was normal and the plant is stable in Mode 3. Reactor Coolant System pressure is approximately 2155 psig and temperature is approximately 550 degree F. Post Transient Review is in progress. Unit 1 is in a normal post trip electrical lineup. No primary or secondary reliefs lifted during the transient. Auxiliary feedwater was placed in service to supply the Steam Generators for decay heat removal via the Main Condenser. The licensee informed both the State and NRC Resident Inspector. | Steam Generator Reactor Coolant System Auxiliary Feedwater Decay Heat Removal Main Condenser Control Rod | |
| ENS 44125 | 7 April 2008 18:45:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Core Element Assembly Calculators Declared Inoperable | On 4/7/2008, at approximately 1345 CDT, with ANO-2 in Mode 3 (Hot Shutdown) at normal operating temperature and pressure, and with all control element assemblies (CEAs) withdrawn approximately 2 stops, the Control Room was notified by the Computer Support Group (CSG) that the software verification for CEAC1 and CEAC2 (Core Element Assembly Calculators) had not been verified. Both CEAC's were declared inoperable. The Reed Switch Position Indicators (CEA positions) are considered inoperable if the CEAC's are inoperable. Tech Spec 3.1.3.3 requires at least one of these position indicators be operable in Modes 3 through 5 for each CEA not fully inserted. All CEAs were approximately 2 steps withdrawn from fully inserted during plant heatup to clear the rod bottom indications in preparations for CEA Drop Time Testing. At 1350 the reactor trip breakers were opened and CEA's were verified to be fully inserted to meet the Action Statement for Tech Spec 3.1.3.3. Note that no actual position indication was lost, nor did the CEAC show any signs of malfunction. The software verifications have been completed by the CSG. The licensee notified the NRC Resident Inspector. The licensee also notified the state. | ||
| ENS 42229 | 26 December 2005 16:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Following Turbine Trip | A reactor trip occurred due to a Reactor Protective System (RPS) actuation from probable Main Turbine problems. The Emergency Feedwater Initiation and Control System (EFIC) actuated on the 'A' Steam Generator only due to a momentary level spike post trip. The RPS and EFIC performed as designed. Unit 1 will remain in mode 3 for troubleshooting efforts. No significant equipment issues observed. This report is being made in compliance with the 4 hour and 8 hour non-emergency notifications required by 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A), respectively. The licensee notified the NRC Resident Inspector and State authorities. All rods fully inserted. The trip was characterized as uncomplicated and all systems functioned as required. Electrical conditions are stable. The reactor is currently at no load temperature and pressure with decay heat being removed via the turbine bypass valves and cooling to the steam generators from the non-safety related auxiliary feedwater system. The trip had no impact on Unit 2. The licensee was in no significant LCO at the time of the trip. The licensee noted that one of the main steam code safeties did stick open for a short period of time but closed after secondary pressure was reduced without any overcooling. The cause of the turbine trip is under investigation. | Steam Generator Auxiliary Feedwater Main Turbine Emergency Feedwater Initiation and Control Main Steam | |
| ENS 40119 | 29 August 2003 19:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip on High Rcs Pressure | Unit 1 experienced an automatic reactor trip at approximately 14:28 CDT as a result of a turbine trip. Initial indications are that the trip was associated with a lightning strike on site. Plant is stable in Mode 3. All control rods fully inserted and there are no offsite radiation releases. Anomalies noted are 'C' Main Phase Transformer Sudden Pressure Trip Alarm, Main Generator Output Breakers failed to automatically open and were subsequently manually opened, and 'B' Once Through Steam Generator (OTSG) main feedwater block valve failed to indicate closed which resulted in a low level condition on 'B' OTSG and subsequent automatic actuation of Emergency Feedwater (EFW). EFW is controlling both OTSG levels at low level limits. A normal fast transfer to the Startup Transformer occurred following the trip and offsite power remains available. All EDGs are in standby, if needed. There is no safety related equipment out of service at this time. Decay heat is currently being removed via the steam dumps to the main condenser. The licensee will conduct an investigation into the equipment failures and determine if any plant equipment was damaged by the lightning strike. The licensee informed the NRC Resident Inspector.
This message is intended to provide clarifying information for Event Notification #40119. Unit 1 experienced an automatic reactor trip at approximately 14:28 CDT as a result of a turbine trip. Initial indications are that the trip was associated with a lightning strike on site. Plant is stable in Mode 3. All control rods fully inserted and there are no offsite radiation releases. The following anomalies were noted in conjunction with plant trip. A 'C' Main Phase Transformer Sudden Pressure Trip Alarm occurred as a result of an indication problem only. The Main Generator Output Breakers failed to automatically open and were subsequently manually opened. The 'B' Once Through Steam Generator (OTSG) main feedwater (MFW) block valve failed to fully close. Complications associated with manual control of feedwater flow when the MFW block valve failed to fully close resulted in a momentary low level in the 'B' OTSG and Emergency Feedwater (EFW) automatic actuation. EFW flow quickly restored the 'B' OTSG level to normal. Both MFW pumps remained operating throughout the transient, and MFW flow was never lost. The EFW System was allowed to recirc for approximately 2 hours. After the MFW block valve was manually closed, the auxiliary feedwater pump was placed in service and EFW and the MFW pumps were secured. The licensee will be notifying the NRC Resident Inspector.
The following information was provided as an update: This message is intended to update the information provided in previous notifications concerning Event Notification #40119. Based on subsequent investigation the automatic reactor trip which occurred at approximately 14:28 CDT on August 29, 2003, was a result of high reactor coolant system (RCS) pressure. The high RCS pressure resulted from the inadvertent closure of the main turbine governor valves. Although not conclusively determined, the closure of these valves is most likely due to a lightning strike on site. This automatic reactor trip also resulted in a main turbine trip. Anomalies associated with this event remain as previously reported. The primary PORV (2450 psia) and Safeties (2500 psia) did not lift during the pressure excursion. The High RCS Pressure setpoint for RPS actuation is 2355 psia. The licensee informed the NRC Resident Inspector. Notified R4DO(Runyan). | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Main Turbine Main Condenser Control Rod |