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The query [[Category:ENS Notification]] [[System::Reactor Protection System]] was answered by the SMWSQLStore3 in 0.6477 seconds.


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 Discovered dateReporting criterionTitleEvent description
ENS 5392310 March 2019 04:59:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Scram Resulting in Rps and Eccs ActuationAt 2259 CST on 3/9/2019, Browns Ferry Unit-3 received an automatic SCRAM on Main Generator Breaker Failure and Turbine Load Reject. Unit-3 declared a Notification of Unusual Event SU1 for loss of offsite AC power to Unit-3 specific 4kV Shutdown Boards for greater than 15 minutes. Primary Containment Isolation Systems (PCIS) Groups 1, 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required. Main steam relief valves lifted on the initial transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated on low reactor water level. HPCI remains in service for reactor level and pressure control. RCIC is not in service at this time, the station is investigating low flow from the pump. All four Unit-3 Diesel Generators started and loaded as expected. Residual Heat Removal System is in service for suppression pool cooling. 4kV Station Unit Boards have been restored from the 161kV system. Actions are in progress to restore 4kV Shutdown Boards to offsite power. This event is reportable within 1 hour in accordance with 10 CFR 50.72(a)(1)(i) for declaration of the Licensees Emergency Plan. Complete as documented on EN 53922. This event requires a 4 hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A). 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs), (4) ECCS (Emergency Core Cooling System) for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system, (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system, and (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. As of the event report, the MSIVs were opened and decay heat was being removed via the bypass valves to the condenser.
ENS 539083 March 2019 03:37:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Trip Due to Degrading Main Condenser VacuumOn March 2, 2019 at 2237 EST, North Anna Unit 2 reactor was manually tripped, while operating at approximately 12 percent power, due to degrading vacuum in the main condenser. The unit was in the process of a planned shutdown for refueling when condenser vacuum degraded to greater than 3.5 inches of mercury absolute. The operations crew entered the reactor trip procedure and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the reactor trip. The reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). There were no ESF system actuations. Decay heat is being removed by the Steam Generator Pressure Operated Relief valves. Unit 2 is in a normal shutdown electrical lineup. The NRC Resident Inspectors have been notified. The Louisa County Administrator will be notified."
ENS 539031 March 2019 04:17:0010 CFR 50.72(b)(3)(iv)(A), System ActuationSpecified System Actuation - Trip of Safety Related BusOn February 28, 2019, at 2217 CST, LaSalle Unit 2 experienced a trip of the 241Y Safety Related Bus during surveillance testing resulting in a valid undervoltage actuation signal to the Common Emergency Diesel Generator ('O' EDG), causing it to start and load to Bus 241Y. The purpose of the surveillance testing was to demonstrate the operability of the breakers necessary to provide the second off site source to Unit 2. This event is reportable 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. In addition to the 241Y bus trip and 'O' EDG actuation signal, the following plant responses occurred as designed due to the momentary loss of this AC Bus: "A" RPS de-energized due to the loss of the 2A Reactor Protection System Motor-Generator Set, and the running Unit 2 Fuel Pool Cooling pump tripped. The Non-Safety Related Bus 241X de-energized resulting in a trip of the Unit 2 Station Air Compressor. All systems have been restored and troubleshooting is currently in progress. Unit 1 remained in MODE 1 during this event. The NRC Senior Resident Inspector has been notified."
ENS 5389625 February 2019 05:24:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram Due to Generator TripAt 0024 EST on 2/25/19, with Unit 1 in Mode 1 at 74 percent power, the reactor automatically tripped due to a generator trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via the feed system. 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The generator trip is under investigation, but is believed to be due to grid perturbations.
ENS 5385231 January 2019 08:01:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Trip - Circulating Water Icing ConditionsAt 0301 (EST) on 1/31/19, with Unit 2 in Mode 1 at 100% power, the reactor was manually tripped due to icing conditions requiring the removal of 4 Circulating Water Pumps from service. The trip was not complex, with all systems responding normally post-trip. 21 CFCU (Containment Fan Cooler Unit) was inoperable prior to the event for a planned maintenance window and did not contribute to the cause of the event and did not adversely impact the plant response to the trip. An actuation of the Auxiliary Feedwater System occurred following the manual reactor trip. The reason for the Auxiliary Feed Water System auto-start was due to low level in a steam generator. Operations responded and stabilized the plant. Decay heat is being removed by the Main Steam Dumps and Auxiliary Feedwater System. 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). This event is also being reported as an eight hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Auxiliary Feed Water System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The icing condition was described as frazil ice. Unit-1 reduced power to 88% because one circulating water pump was shutdown.
ENS 538199 January 2019 05:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Reactor Trip from Full Power Due to Rps TestingAt 1034 EST on January 9, 2019, with the reactor at 100% power, an automatic reactor trip was initiated. The trip occurred while Reactor Protection System testing was in progress. The trip was uncomplicated with all systems responding normally following the rip. Troubleshooting and investigation of the cause is ongoing. All full-length control rods inserted fully. Auxiliary Feedwater System actuated as designed in response to low steam generator water levels. Operations stabilized the plant in Mode 3 (hot standby). Decay heat is being removed by the turbine bypass valve. This condition has no impact to the health and safety of the public. The licensee notified the NRC Resident Inspector.
ENS 537641 December 2018 08:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Trip of Unit 2 Due to a Load RejectionAt 1006 (PST), on December 1, 2018, with Unit 2 at 100 percent power, the reactor automatically tripped due to a load rejection from the 500 kV offsite electrical system. Operations responded and stabilized the plant. Decay heat is being removed by the Main Steam system to the main condenser using the steam dump valves. The cause of the load rejection is currently under investigation. 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). Additionally, due to the actuation of the Auxiliary Feedwater System, as expected, this event is being reported 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 Senior Resident Inspector was notified. A press release is planned for this event. All control rods fully inserted and the trip was uncomplicated. There was no effect on Unit 1.
ENS 537601 October 2018 05:00:0010 CFR 50.73(a)(1), Submit an LERInvalid Specified System ActuationThis 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of the Reactor Protection System (RPS). At 0147 (EST) on October 1, 2018, Seabrook Unit 1 was in Mode 3 shutdown, when an invalid Reactor Protection System actuation occurred due to a high Source Range detector. All equipment responded in accordance with the plant design. Specifically, all actuations were complete and successful. There were no safety consequences or impacts on the health and safety of the public. The event was entered into Seabrook's corrective action program for resolution. The NRC Resident Inspector has been notified. The Source Range detector which gave the invalid input has been replaced.
ENS 5369828 October 2018 05:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram During a Reactor ShutdownAt 0445 (CDT), with reactor power less than 1% rated thermal power on Instrument Range Monitor (IRM) ranges 6 and 7, Clinton Power Station received an automatic Reactor Protection System (RPS) actuation. The Reactor Scram Off Normal procedure was entered and all control rods were verified to be fully inserted. The apparent cause of the scram is cold water injection causing an upscale trip of the IRMs due to Motor Driven Reactor Feedwater Pump (MDRFP) Feedwater Regulating valve 1FW004 valve coming off the full shut seat momentarily. All systems responded appropriately following the scram and the plant is currently stable. Clinton Power Station will be proceeding to Mode 4 to support the planned Maintenance Outage. The NRC Senior Resident Inspector has been notified."
ENS 5366116 August 2018 05:00:0010 CFR 50.73(a)(1), Submit an LERInvalid Actuation of a General Containment Isolation Signal Affecting More than One SystemThis 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On August 16, 2018, at approximately 1736 CDT, Browns Ferry Nuclear Plant (BFN), Unit 2 experienced an unexpected loss of the 2B Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected with the exception of the Unit 1 Refuel Zone Supply Fan Outboard Isolation Damper, 1-FCO-64-5, that failed to indicate closed position. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the RPS MG (Motor Generator) Set trip was a failed (shorted) operating coil associated with the 480 VAC motor starter inside the control box. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Reports 1440047 and 1440050. The NRC Resident Inspector has been notified of this event."
ENS 536485 October 2018 04:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Scram Due to Low Reactor Water LevelOn Friday, October 5, 2018 at 1209 hours, with the reactor at 100 percent core thermal power, Pilgrim Nuclear Power Station (PNPS) automatically tripped due to reactor water level perturbation and receipt of a low reactor water level Reactor Protection System (RPS) signal. The cause of the low reactor water level is under investigation. The plant is in hot shutdown. All other plant systems responded as designed. Pressure is being controlled using the Mechanical Hydraulic Control System and Main Condenser. Reactor water level is being maintained with the feedwater and condensate system. During the automatic reactor scram the plant experienced the following isolation signals as designed: Group 2 Isolation: Miscellaneous containment isolation valves Group 6 Isolation: Reactor Water Clean-up Reactor Building Isolation System Actuation Due to the RPS actuation while critical, this event is being reported as a four-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B), 'any event that results in actuation of the reactor protection system (RPS) when the reactor is critical.' This notification is also being made in accordance with 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section ... ' (B)(2) 'General containment isolation signals affecting containment isolation valves in more than one system.' This event has no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee will notify the Massachusetts Emergency Management Agency.
ENS 536434 October 2018 04:00:0010 CFR 50.72(b)(3)(iv)(A), System ActuationEn Revision Imported Date 10/22/2018

EN Revision Text: MANUAL REACTOR TRIP DURING LOW POWER PHYSICS TESTING At 0544 EDT on October 4, 2018, with Unit 1 in Mode 2 with reactor power in the intermediate range performing low power physics testing, the reactor was manually tripped due to a rod control urgent failure alarm. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam system. Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods inserted as expected. The cause of the rod control urgent failure is being investigated.

  • * * UPDATE FROM KEVIN LOWE TO DONALD NORWOOD AT 1408 EDT ON 10/19/2018 * * *

This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A). During Dynamic Rod Worth Measurement testing, Control Bank Charlie was inserted approximately 153 steps when the urgent failure occurred (CBC positioned at 75 steps out). Following the scram, additional analysis concluded that the reactor was subcritical when the Reactor Protection System was actuated." The licensee notified the NRC Resident Inspector. Notified the R2DO (McCoy).

ENS 5360814 September 2018 05:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor ScramAt 1644 (CDT) a manual reactor scram was inserted by placing the Reactor Mode Switch to Shutdown. At 1643 (CDT) the Condensate Booster Pump A tripped on low suction pressure. At 1644 (CDT) the Reactor Feed Pump A tripped on low suction pressure. A Recirculation Flow Control Valve runback occurred as designed. Reactor Water level was approaching the Automatic Low Water Level 3 (11.4 inches) scram set point and manual actions were taken by placing the Mode Switch to Shutdown before the low level set point was reached. All systems responded as expected following the manual scram. The plant is stable in mode 3. This event is being reported under 10CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. The NRC Senior Resident Inspector has been notified. All control rods fully inserted, and decay heat is being removed through the turbine bypass valves to the main condenser. The licensee is investigating the cause of the event.
ENS 5360614 September 2018 04:00:0010 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 Failure of the Steam Generator Feed Regulating ValveAt 1323 (EDT) on 9/14/18, with Unit 2 in Mode 1 at 90% power, the reactor automatically tripped due to a failure of 23BF19, 23 Steam Generator (SG) Feed Regulating Valve. The trip was not complex, with all systems responding normally post-trip. No equipment was inoperable prior to the event. An actuation of the auxiliary feedwater system occurred following the automatic reactor trip. The reason for the auxiliary feed water system auto-start was due to low level in the steam generator. Operations responded and stabilized the plant. Decay heat is being removed by the main steam dumps and auxiliary feedwater system. Unit 1 is not affected. 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). 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 auxiliary feed water system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The Lower Alloways Creek Township will be notified."
ENS 5355722 August 2018 04:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor TripAt 0943 EDT on August 22, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The reactor automatically tripped due to a main turbine trip signal. An investigation is in progress. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event."
ENS 5346722 June 2018 04:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor TripAt 0841 EDT on June 22, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 95% power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The reactor automatically tripped due to a main turbine trip. The turbine trip was caused by main generator electrical trip. An investigation is in progress. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip.
ENS 5345916 June 2018 05:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Trip During StartupAt 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.
ENS 5342423 May 2018 18:28:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Trip on Low Departure from Nucleate Boiling Ratio

The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. On May 23, 2018, at approximately 1128 Mountain Standard Time (MST), the Palo Verde Nuclear Generating Station (PVNGS) Unit 2 control room received reactor protection system alarms for low departure from nucleate boiling ratio and an automatic reactor trip occurred. Prior to the reactor trip, Unit 2 was operating normally at 100 percent power. Plant operators entered the reactor trip procedures and diagnosed an uncomplicated reactor trip. All CEAs (control element assemblies) fully inserted into the core. No emergency classification was required per the PVNGS Emergency Plan. The Unit 2 safety-related electrical buses remained energized from normal offsite power during the event. There was no impact to the required circuits between the offsite transmission network and the onsite Class 1E Electrical Power Distribution System; the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 2 is currently stable in Mode 3 with the reactor coolant system at normal operating temperature and pressure. The cause of the reactor trip is under investigation. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.

The NRC Resident Inspector has been informed of the Unit 2 reactor trip. Decay is being removed via steam dumps to condenser. Units 1 and 3 at Palo Verde were unaffected by the transient and continue to operate at 100 percent power.

ENS 5341018 May 2018 13:51:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram Caused by Main Transformer TripAt 0651 (PDT) on May 18th, 2018, Columbia Generating station experienced a Main Transformer trip, that caused a Reactor Scram. Reactor Power, Pressure and Level were maintained as expected for this condition. MS-RV-1A (Safety Relief Valve) and MS-RV-1B (Safety Relief Valve) opened on reactor high pressure during the initial transient. MS-RV-1B appeared to remain open after pressure lowered below the reset point. The operating crew removed power supply fuses for MS-RV-1B and it currently indicates intermediate position. SRV (Safety Relief Valve) tail pipe temperatures indicate all valves are closed. Suppression pool level and temperature have remained steady within normal operating levels. All control rods inserted and reactor power is being maintained subcritical. RPV (Reactor Pressure Vessel) water level is being maintained with condensate and feed system with startup flow control valves in automatic. Reactor Pressure is being maintained with the Turbine Bypass valves controlling in automatic. The main condenser is the heat sink. No ECCS (Emergency Core Cooling Systems) systems actuated or injected; the EOC-RPT (End of Cycle-Recirculation Pump Trip) and RPS (Reactor Protection System) systems actuated causing a trip of the RRC pumps and a reactor scram. Core recirculation is being maintained with RRC-P-1A (Reactor Recirculation Pump) running. No release has occurred. At this time there will be no notifications to state, local or other public agencies. The NRC Senior Resident has been notified. The cause of the event is currently under investigation. Plant conditions are stable. The plant is in its normal electrical alignment and offsite power is available to the site.
ENS 533877 May 2018 07:36:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Rx Trip Due to High-High Level in Moisture Separator Drain TankOn May 7, 2018 at 0336 (EDT), DC Cook Unit 2 Reactor was manually tripped due to a high-high level experienced in the East Moisture Separator Drain Tank (MSDT) of the Moisture Separator Reheater (MSR). This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The NRC Resident Inspector has been notified. Unit 2 is being supplied by offsite power. All control rods fully inserted. All Aux Feedwater Pumps started properly. Decay heat is being removed via the Steam Generator Power Operated Relief Valves following Main Steam Stop Valve closure at 0431 due to a slow RCS (Reactor Coolant System) cooldown. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the Post Trip Review. No radioactive release is in progress as a result of this event.
ENS 5333614 April 2018 14:40:0010 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge
10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Automatic Reactor Scram and Emergency Core Cooling System Injection

At 1040 EDT, Fermi 2 automatically scrammed on RPV (Reactor Pressure Vessel) Level 3 following a loss of the Division 1 Station System Transformer (SST) #64. All control rods fully inserted. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically started as designed on Reactor Water Level (RWL) 2 and restored RWL. The lowest RWL reached was 101.8 inches (above Top of Active Fuel). HPCI injected for approximately 35 seconds. RWL is currently being maintained in the normal level band with RCIC. No Safety Relief Valves (SRVs) actuated. All isolations and actuations for RWL 3 and 2 occurred as expected. Investigation into loss of SST #64 continues. At the time of the scram, all Emergency Core Cooling Systems (ECCS) and Emergency Diesel Generators (EDGs) were operable, and no safety related equipment was out of service. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(A), any event that results in ECCS discharge into the reactor coolant system as a result of a valid signal and 10CFR50.72(b)(2)(iv)(B), any event that results in the actuation of the Reactor Protection System (RPS) when the reactor is critical. Following the loss of power and reactor scram, the Division 2 EECW (Emergency Equipment Cooling Water) Temperature Control Valve (TCV) controller was in Emergency Manual and maintaining max cooling. Operators placed the controller in Auto and the TCV is controlling normally. The NRC Senior Resident has been notified. Decay heat is being removed via Division 2 steam dumps to the condenser. The plant is in a modified shutdown electric lineup with offsite power available and stable. Emergency diesel generators did automatically start and load.

  • * * UPDATE ON 4/14/2018 AT 1838 EDT FROM JEFF MYERS TO HOWIE CROUCH * * *

This update provides additional clarification of the applicable reporting criteria for this event associated with Primary Containment Isolation Actuations. All isolations and actuations for RWL (Groups 4, 13, and 15) and RWL 2 (Groups 2, 10, 11, 12, 14, 16, 17, and 18) occurred as expected. This report is also being made in accordance with 10CFR50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B): RPS, HPCI, and RCIC. RPV pressure is being maintained by the bypass valves to the main condenser. All actuations that occurred were fully completed and restored. The licensee notified the NRC Resident Inspector. Notified R3DO (Stone).

  • * * UPDATE ON 4/15/2018 AT 1950 EDT FROM KELLEY BELENKY TO DAVID AIRD * * *

This update provides additional information regarding the specified system actuations and an additional applicable reporting criteria. The loss of Division 1 Station System Transformer (SST) #64 at 1040 EDT on 4/14/2018 resulted in the automatic initiation of Emergency Diesel Generators (EDG) 11 and 12. The EDGs started as expected and continue to supply their associated busses. This is reportable pursuant to 10CFR50.72(b)(3)(iv)(A), as an event or condition that resulted in a valid actuation of any system listed in paragraph (b)(3)(iv)(B), including EDGs. In addition, the loss of the Division 1 SST #64 resulted in the expected transfer from the normal to alternate power source for the Low Pressure Coolant Injection (LPCI) swing bus, rendering LPCI loop select inoperable. The alternate power source continued to energize the LPCI swing bus throughout the event until the system was realigned to the normal power source at 1239 EDT on 4/14/2018. This condition is reportable pursuant to 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified R3DO (Stone).

ENS 5332712 April 2018 13:20:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor TripAt 0920 EDT on April 12, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The cause of the automatic reactor trip is being investigated. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip.
ENS 533197 April 2018 12:36:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Trip and Pcis Actuation During Stator Cooling System Testing

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

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

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

ENS 5326918 March 2018 16:58:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Scram Due to Turbine Control Valve ClosureAt 1158 CDT on March 18, 2018, the Unit 1 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from High Reactor Steam Dome Pressure in response to Turbine Control Valve Closure. The reactor had been operating at 100 percent power. Investigation is in progress. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Steam Relief Valves (MSRVs) operating on the initial transient as expected. Main Turbine Bypass Valves are currently controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. All safety system operated as expected. At no time was public health and safety at risk. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation' and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 5325913 March 2018 14:00:0010 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual HeatOff-Site Power Unavailable Due to Winter StormOn March 13, 2018 at 1000 hours (EDT), with the reactor in Cold Shutdown condition, both 345kV incoming power lines and 23 kV Shutdown Transformer became unavailable during the Northeast winter storm. Per procedures, the emergency on-site emergency power supplies (Emergency Diesel Generators) were running and providing power to essential systems. In addition, the back-up Diesel Air Compressor was in service and one Reactor Protection System bus was on the back-up power supply prior to the loss. With both 345kV incoming power lines and 23 kV Shutdown Transformer unavailable, Pilgrim Nuclear Power Station procedures direct a report be made to the NRC per the requirements of Title 10 Code of Federal Regulations 50.72(b)(3)(v), any event that could have prevented the fulfillment of the safety function. No actual loss of safety function has occurred since the on-site emergency power supplies are maintaining the reactor in a safe shutdown condition and removing residual heat. The loss of incoming power is under investigation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified.
ENS 5321716 February 2018 15:14:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Trip During Solid State Protection System TestingAt 1014 (EST) hours on 2/16/18, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped when the Reactor Trip Breakers opened during Train B Solid State Protection System (SSPS) testing. The trip was uncomplicated with all systems responding normally post-trip. Operations manually started the motor driven auxiliary feedwater pumps. The turbine driven auxiliary feedwater pump (TDCAP) auto-started on low steam generator level. A Feedwater Isolation occurred as designed due to the Reactor Trip and Lo Tave condition. Operations stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected. Due to the Reactor Protection System actuation while critical, actuation of the TDCAP and motor driven auxiliary feedwater pumps along with the Feedwater Isolation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5321516 February 2018 04:53:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness
Automatic Reactor Trip Due to Low Departure from Nucleate Boiling Signal

The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. On February 15, 2018, at approximately 2153 Mountain Standard Time (MST), the Palo Verde Nuclear Generating Station (PVNGS) Unit 1 Control Room received Reactor Protection System alarms for Low Departure from Nucleate Boiling Ratio and an automatic reactor trip occurred. Prior to the reactor trip, Unit 1 was operating normally at 100 percent power. Plant operators entered the emergency operations procedures and diagnosed an uncomplicated reactor trip but noted that Reactor Coolant Pumps 1B and 2B were not running due to a loss of power. All CEAs (Control Element Assemblies) fully inserted into the core. Following the reactor trip, all nuclear instruments responded normally. No emergency classification was required per the PVGS Emergency Plan. The PVGS Unit 1 safety related electrical busses remained energized from normal offsite power during the event. The Unit 1 'B' Diesel Generator is currently removed from service for maintenance. Due to ongoing planned maintenance on NAN-X02, Startup Transformer 2, fast bus transfer for NAN-S02 (from NAN-S04) was blocked. This resulted in a loss of offsite power to NAN-S02 and NBN-S02. The offsite power grid is stable. Unit 1 is currently stable in Mode 3 with the reactor coolant system at normal operating temperature and pressure. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public. The NRC Resident Inspector has been informed of the Unit 1 reactor trip.

  • * * UPDATE ON 2/16/18 AT 1640 EST FROM DAVID HECKMAN TO DONG PARK * * *

Unit 1 is stable in Mode 3 following an uncomplicated trip. Offsite power has been restored to non-safety related electrical busses. Troubleshooting continues to determine the cause of the event. During performance of the alarm response procedure, it was identified that the seismic monitoring (SM) system had been in alarm since the reactor trip and was incapable of performing its emergency plan function. Pursuant to 10 CFR 50.72(b)(3)(xiii), this condition constitutes a major loss of emergency assessment capability. Compensatory measures have been implemented in accordance with PVNGS procedures to provide alternative methods for HU2.1 event classification with the SM system out of service. Maintenance is currently in progress to restore SM system functionality. The licensee notified the NRC Resident Inspector. Notified R4DO (Werner).

  • * * UPDATE AT 1537 EDT ON 03/30/18 FROM LORRAINE WEAVER TO JEFF HERRERA * * *

Station staff completed an evaluation of event EN #53215 reported on February 15, 2018, and determined that the seismic monitoring system remained capable of assessing a seismic event following the reactor trip. Therefore, a major loss of emergency assessment capability pursuant to 10 CFR 50.72(b)(3)(xiii) did not occur as reported in the update on February 16, 2018. The NRC Resident Inspectors have been notified. Notified the R4DO (Gaddy).

ENS 531921 February 2018 16:57:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Scram

At 1057 CST on February 1, 2018 with the unit in Mode 1 at approximately 27% power, a manual actuation of the Reactor Protection System (RPS) was initiated due to an unexpected trip of the B Recirc Pump with A Recirc Pump in fast speed. B Recirc Pump tripped during transfer from slow to fast speed resulting in single loop operation. Operators were unable to reconcile differing indications of core flow. This resulted in a conservative decision to initiate a manual scram. The cause of the B Recirc Pump trip and the apparent issues with core flow indication are under investigation. The plant is currently stable in Mode 3. The plant response to the scram was as expected. All control rods (fully) inserted as expected; the feedwater system is maintaining reactor vessel water level in the normal control band and reactor pressure is being maintained with steam line drains and main turbine bypass valves. The NRC Senior Resident (Inspector) has been notified.

  • * * RETRACTION AT 1015 EDT ON 03/22/2018 FROM DAVID DABADIE TO OSSY FONT * * *

This event was initially reported under 10 CFR 72(b)(2)(iv)(B) as a manual actuation of the RPS due to an unexpected trip of the B Reactor Recirculation Pump with the A Reactor Recirculation Pump running in fast speed (Single Loop Operations). Operations was unable to reconcile differing indications of core flow and made the conservative decision to perform a planned shutdown in accordance with normal operating procedures. Therefore, this event 'resulted from and was part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.73(a)(2)(iv)(A) and NUREG-1022 Section 3.2.6. Consequently, this event is not reportable as an actuation of RPS. The NRC Resident Inspector has been notified. R4DO (Groom) has been notified.

ENS 5318831 January 2018 00:22:0010 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 Main Turbine Load OscillationsOn 1/30/2018 at 1750 (CST), the Reactor Pressure Control Malfunctions ONEP (Off Normal Event Procedure) was entered due to main turbine load oscillations of approximately 30 MWe peak to peak. At 1822 (CST), a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown due to continued main turbine load oscillations. Reactor SCRAM ONEP, Turbine Trip ONEP, and EP-2 were entered. Reactor water level was stabilized at 36 inches narrow range on startup level and reactor pressure stabilized at 933 psig using main turbine bypass valves. Reactor Water Level 3 (11.4 inches) was reached which is the setpoint for Group 2 (RHR to Radwaste Isolation) and Group 3 (Shutdown Cooling Isolation). No valve isolated in these systems due to all isolation valves in these groups being in their normally closed position. The lowest Reactor Water level reached was -36 inches wide range. No other safety system actuations occurred and all systems performed as designed. That event is being reported under 10CFR 50.72(b)(2)(iv)(B) as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical and also reported under 10CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of RPS. The MSIVs are open with decay heat being removed via steam to the main condenser using the bypass valves. Off site power is stable, and the plant is in a normal shutdown electrical lineup. RCIC (Reactor Core Isolation Cooling) was out of service for maintenance, and the reactor water level did not reach the system activation level. The cause of the main turbine load oscillations being investigated. The licensee notified the NRC Resident Inspector.
ENS 5316210 January 2018 15:28:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Scram Due to Turbine Control Valve Fast Closure Scram SignalAt 0928 CST on January 10, 2018, the Unit 3 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from Turbine Control Valve Emergency Trip System pressure low. The reactor had been operating near 73 percent power for an emergent issue for Turbine Control Valve (TCV) No. 3. With TCV No. 3 out of service and closed, the unit was operating with RPS in a half scram condition. A subsequent failure of the TCV No. 2 sensing line resulted in RPS coincidence logic being met for TCV fast closure SCRAM. The investigation of the TCV No. 2 sensing line failure continues. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Turbine Bypass Valves controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a preplanned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident inspector has been notified.
ENS 531474 January 2018 19:10:0010 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 Partial Loss of Offsite Power During Winter StormOn January 4, 2018, at 1410 hours EST, with the reactor at approximately 100 percent power and steady state conditions, the winter storm across the Northeast caused the loss of offsite 345 kV Line 342. Reactor power was reduced to approximately 81 percent and a procedurally required manual reactor scram was initiated. All control rods fully inserted. As a result of the reactor scram, indicated reactor water level decreased, as expected, to less than +12 inches resulting in automatic actuation of the Primary Containment Isolation Systems for Group II - Primary Containment Isolation and Reactor Building Isolation System, and Group VI - Reactor Water Cleanup System. Reactor Water Level was restored to the normal operating band. The Primary Containment Isolation Systems have been reset. The Reactor Protection System signal has been reset. Following the reactor scram, the non-safety related Control Rod Drive Pump "B" tripped on low suction pressure. Control Rod Drive Pump "A" was placed in service. All other systems operated as expected, in accordance with design. This event is reportable per the requirements of Title 10, Code of Federal Regulations (CFR) 50.72 (b)(2)(iv)(B) - "RPS Actuation" and 10 CFR 50.72 (b)(3)(iv)(A) - "Specified System Actuation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The main steam isolation valves are open with decay heat being removed via steam to the main condenser. Offsite power is still available from 345kV line 355. As a contingency, emergency diesel generators are running and powering safety busses per licensee procedure. The licensee notified the Commonwealth of Massachusetts. The licensee will be notifying the town of Plymouth as part of their local notifications. The licensee will be issuing a press release.
ENS 5311211 December 2017 13:57:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Trip in Response to Indication of Multiple Dropped Control RodsWhile operating at 97% power, the Watts Bar Unit 2 reactor was manually tripped at 0857 EST on December 11, 2017 due to multiple dropped control rods. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and the Steam Dump System. The cause of the dropped rods is being investigated. The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. No safety or relief valves lifted during this event.
ENS 531109 December 2017 19:48:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material
10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Manual Reactor Scram Due to Loss of Division 1 Ac Power to Numerous Components

At approximately 1347 (CST) on 12/09/17, the Main Control Room received annunciators that indicated a trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1 breaker. Numerous Division 1 components lost power (powered from unit subs 1A and A1). The Division 1 containment Instrument Air isolation valves had failed closed by design due to the loss of power. Due to the loss of containment instrument air, several control rods began to drift into the core as expected and, by procedure, the reactor mode switch was placed in the shutdown position at 1353 (CST). All control rods fully inserted. Also due to the loss of power, the Fuel Building ventilation dampers failed closed by design. With the normal ventilation system secured, secondary containment differential pressure rose to slightly greater than 0 inches water gauge which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge at 1348 (CST). The Control Room entered EOP-8, Secondary Containment Control. Secondary Containment differential pressure was restored within Technical Specification requirements at 1351 (CST) by starting the Division 2 Standby Gas Treatment system. This event is being reported as a manual actuation of the Reactor Protection System (RPS) and as a Condition that Could Have Prevented Fulfillment of a Safety Function.

The cause is currently under investigation. The NRC Resident has been notified. The licensee informed the NRC Resident Inspector.

  • * * UPDATE FROM DALE SHELTON TO VINCE KLCO AT 1658 EST ON 12/10/2017 * * *

During a review of plant logs it was identified that the primary to secondary containment differential pressure was identified to be outside of Technical Specification 3.6.1.4 limits of 0 plus or minus 0.25 psid at 2009 on 12/9/17 due to the primary containment ventilation system dampers closing as a result of the loss of power. This parameter is an initial safety analysis assumption to ensure that primary containment pressures remain within the design values during a Loss of Coolant Accident (LOCA). As a result, this condition is reportable as an unanalyzed condition that significantly degrades plant safety. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).

  • * * UPDATE FROM MICHAEL ANTONELLI TO VINCE KLCO ON 12/11/17 AT 1805 EST * * *

During the post transient review of the trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1, it was identified that the unplanned INOPERABILITY of the Low Pressure Core Spray (LPCS) system due to the loss of power to the injection valve constitutes an event or condition that could have prevented fulfillment of a safety function and is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. The High Pressure Core Spray (HPCS) remained available to perform the core spray function, if necessary, during a design basis Loss of Coolant Accident (LOCA), however HPCS and LPCS are each considered single train safety systems. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).

ENS 5309025 November 2017 08:38:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Scram During Startup

At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. At 0149 (CST), with reactor power just above the point of adding heat, IRM (Intermediate Range Monitor) channels A, C, and D received a spurious upscale trip signal which immediately cleared. Upon investigation, operability of RPS (Reactor Protection System) scram function for Intermediate Range Detectors was placed in question. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. The licensee notified the NRC Resident Inspector.

  • * * UPDATE ON NOVEMBER 26, 2017, AT 1850 FROM GRAND GULF TO MICHAEL BLOODGOOD * * *

At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. At 0149 (CST), with reactor power just above the point of adding heat, Intermediate Range Monitor neutron flux detector (IRM) channels A, C, and D received a spurious Upscale Trip signal which immediately cleared. Upon investigation, IRM channels A, C, and D were declared Inoperable. IRM G was already Inoperable for another reason. RPS scram function from IRM channels B, E, F, and H was always Operable and available. That event is being reported under 10CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. This Revised Statement to Event Notification # 53090 is being made to make it clear that only four IRM channels (A, C, D, G) were Inoperable and that the IRM RPS SCRAM function was still available from the four remaining Operable IRM channels (B, E, F, and H). The licensee notified the NRC Resident Inspector. Notified R4DO (O'Keefe)

  • * * RETRACTION ON 01/16/2018 AT 1629 EST FROM JASON COMFORT TO DAVID AIRD * * *

On 11/25/17, at 0149 (CST), with reactor power just above the point of adding heat, Intermediate Range Monitor neutron flux detector (IRM) channels A, C, and D received a spurious Upscale Trip signal which immediately cleared. Upon investigation, IRM channels A, C, and D were declared Inoperable. IRM G was already Inoperable for another reason. At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. RPS scram function from IRM channels B, E, F, and H was always Operable and available. That event was initially being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS), when the reactor is critical. After the trip alarms were received, the Operators spent approximately twenty minutes investigating possible causes and implications, and consulted with Reactor Engineering and the Shift Technical Advisor. The investigation showed that the plant was stable and the upscale IRM alarms were spurious. A review of plant technical specifications by the operators determined that a plant shutdown was not required. After further discussions, Operations concluded that a shutdown to allow further investigation of the issue was the prudent course of action. Prior to shutting down, Operations spent approximately twenty minutes reviewing procedures, notifying personnel to exit containment, and conducting a brief. The shutdown was then conducted by inserting a manual reactor scram by placing the reactor mode switch in SHUTDOWN. This was initially reported under 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the RPS. Based on the sequence of events, and Operator actions in conducting the shutdown, the event is considered 'part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.73(a)(2)(iv)(A). In accordance with NUREG-1022, Section 3.2.6, the event is not reportable as an actuation of RPS. The licensee notified the NRC Resident Inspector. Notified R4DO (Taylor).

ENS 530836 October 2017 14:10:0010 CFR 50.73(a)(1), Submit an LERInvalid Specified System ActuationOn October 6, 2017 at 0910 CDT hours, with Unit 1 in Mode 1 (Power Operation), the 1A Diesel Generator Cooling Water Pump (DGCWP) automatically started. The cause was the misoperation of the 1B/C RHR (Residual Heat Removal) Room Cooler Fan (1VY03C) control switch, which was placed in the start position instead of the intended pull-to-lock position. The start of the 1VY03C fan resulted in the automatic actuation of the 1A DGCWP. This system actuation is reportable in accordance with 10CFR50.73(a)(2)(iv)(A). The invalid actuation was not part of a pre-planned sequence during testing or reactor operation. The 1A DGCWP, an emergency service water system that does not normally run and that serves as an ultimate heat sink, responded satisfactorily. This call is being made in accordance with 10CFR50.73(a)(1), which states that in the case of an invalid actuation reported under 10CFR50.73(a)(2)(iv), other than an actuation of the reactor protection system when the reactor is critical, the licensee may provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written Licensee Event Report. The licensee notified the NRC Resident Inspector.
ENS 530567 November 2017 10:04:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Unit 1 Automatic Reactor Trip Due to Main Unit Generator Over CurrentOn November 7, at 0504 (EST), BVPS (Beaver Valley Power Station) Unit 1 experienced an automatic reactor trip due to Main Unit Generator over current. The Auxiliary Feedwater system activated and remains in service. Offsite power supply is available. Normal and Emergency busses are being supplied by Offsite power. One Source Range channel failed to energize due to its corresponding Intermediate Range instrument being under compensated. It was manually energized and is not indicating as expected. The second Source Range instrument energized but is reading erratically. Both Source Range instruments have been declared inoperable and the appropriate Technical Specification has been complied with by making the Control Rods not capable of withdrawal and isolating all dilution flow paths. Plant trip response was as expected without complications, and all control rods fully inserted in the core. The plant is currently stable in Mode 3. This event is being reported as an actuation of the Reactor Protection system 10 CFR 50.72(b)(2)(iv)(B) and a Specified System Actuation (Auxiliary Feedwater System) 10 CFR 50.72(b)(3)(iv)(A). BVPS Unit 2 is unaffected by this event and remains at 100% power in Mode 1. The NRC Resident Inspector has been notified.
ENS 530524 November 2017 00:22:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Indian Point Unit 3 Reactor Trip on Low Steam Generator Level

On November 3rd, 2017 at 2022 EDT, the Indian Point Unit 3 Reactor Protection system automatically actuated at 100 percent power. Annunciator first out indication was from 33 SG (Steam Generator) Low Level. This automatic reactor trip is reportable to the NRC under 10 CFR 50.72(b)(2)(iv)(B). All control rods fully inserted on the reactor trip. All safety systems responded as expected. The Auxiliary Feedwater System actuated as expected. Offsite power and plant electrical lineups are normal. All plant equipment responded normally to the unit trip. No primary or secondary code safeties lifted during the trip. The Auxiliary Feedwater System actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). The Emergency Diesel Generators did not start as offsite power remained available and stable. The Unit remains on offsite power and all electrical loads are stable. Unit 3 is in Hot Standby at normal operating temperature and pressure with decay heat removal using auxiliary feedwater to the steam generators and normal heat removal through the condenser via the high pressure steam dumps. Unit 2 was unaffected and remains at 100 percent power. A post trip investigation is in progress. The licensee indicated that Radiation Monitor number 14 spiked twice during the transient, however, is currently not indicating any signs of radiation. The licensee will notify the NRC Resident Inspector and the NY Public Service Commission.

  • * * UPDATE AT 1523 EST ON 11/06/17 FROM RAMIREZ OVIDIO TO JEFF HERRERA * * *

The initial notification stated that Indian Point Unit 3 Reactor Tripped on 33 SG (Steam Generator) Low Level, this is incorrect. Indian Point Unit 3 Reactor Tripped on a Turbine Trip. The Turbine Trip was caused by a Generator Back-up Lockout Relay. The Turbine Trip was the 'first' annunciator first-out but was acknowledged instead of silenced during initial operator actions. The Turbine Trip first-out being acknowledged allowed a Low Steam Generator first-out to later annunciate. A Low Steam Generator Level is an expected condition post trip. This update does not change any actions taken by the operating team or required notifications under 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). A post trip investigation remains in progress. The licensee will notify the NRC Resident Inspector and the NY Public Commission. Notified the R1DO(Cook)

ENS 5291820 August 2017 23:05:0010 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 a Rise in Main Condenser Back Pressure

On August 20, 2017 at 1605 PDT, Columbia Generating Station was manually scrammed from 100 percent power due to a rise of Main Condenser back pressure. Manual scram of the unit is procedurally required upon a loss of Main Condenser back pressure. Preliminary investigations indicate that the Main Condenser air removal suction valve (AR-V-1) closed, resulting in the Condenser back pressure rising to within 1.0 inch Hg of the setpoint with reactor power greater than 25 percent. Further investigations continue. All control rods fully inserted. In addition to the closure of the air removal suction valve, one of two Reactor Feedwater startup flow control valves did not adequately operate to control Reactor vessel level and resulted in a high-level (Level-8) actuation tripping the Reactor Feedwater System. All other systems operated as expected. Reactor water level is currently being controlled manually with the start-up level control isolation valve. AR-V-1 has been manually opened with a jumper and temporary air supply. Reactor decay heat is being removed via bypass valves to the Main Condenser. This event is being reported under the following: 10 CFR 50.72(b)(2)(iv)(B), which requires a four-hour notification for any event or condition that results in actuation of the Reactor Protection System when the reactor is critical. The licensee notified the NRC Resident Inspector. The licensee plans to issue a press release.

  • * * UPDATE ON 8/24/17 AT 1937 EDT FROM MATT HUMMER TO DONG PARK * * *

The licensee is updating the notification to include an 8 hour notification under 10 CFR 50.72(b)(3)(iv)(A) for a specified system actuation due to a Level 3 isolation signal which occurred approximately 20 minutes after the scram. The licensee is currently in cold shutdown to repair the Reactor Feedwater startup flow control valve. The licensee has notified the NRC Resident Inspector. Notified R4DO (Farnholtz).

ENS 5291519 August 2017 01:55:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Scram While at 100 Percent PowerAt 2055 CDT on August 18, 2017, an automatic actuation of the reactor protection system occurred while the plant was operating at 100 percent power. No plant parameters requiring the actuation of the emergency diesel generators or the emergency core cooling system were exceeded. The main feedwater system remained in service following the scram to maintain reactor water level, and the main condenser remained available as the normal heat sink. The scram occurred after a planned swap of the main feedwater master controller channels in preparation for scheduled surveillance testing. When the channel swap was actuated, the feedwater regulating valves moved to the fully open position. The scram signal originated in the high-flux detection function of the average power range monitors, apparently from the rapid increase in feedwater flow. The cause of the apparent feedwater controller malfunction is under investigation. The NRC Resident Inspector has been notified. No safety relief valves opened. Decay heat is being removed via steam to the main condenser using the bypass valves and steam drains. The licensee intends to go to Cold Shutdown to investigate the malfunction.
ENS 5287225 July 2017 08:28:0010 CFR 50.72(b)(3)(iv)(A), System ActuationManual Reactor Trip Due to Rod Position Indication System Being InoperableOn July 25, 2017, at 0428 Eastern Daylight Time (EDT) Watts Bar Nuclear Plant (WBN) Unit 2 was in Mode 3, beginning a Reactor Startup. While in the initial phase of withdrawing the first of four Control Rod banks, the two associated group demand position indicators deviated greater than 2 steps from each other. In accordance with Technical Requirement 3.1.7, Position Indication System, Shutdown, with one or more group demand position indicators inoperable, the reactor trip breakers are to be opened immediately. Operations personnel opened the reactor trip breakers immediately by initiating a manual trip of the Reactor Protection System (RPS). The Auxiliary Feedwater system was in service and controlling Steam Generator water levels at the time of the event and did not receive any valid actuation signals. No other system actuations occurred as a result of this reactor trip and all systems operated as designed. The cause of the position indication system inoperability is currently under investigation. NRC Resident Inspector has been notified.
ENS 5285924 March 2017 19:48:0010 CFR 50.73(a)(1), Submit an LERInvalid Specified System Actuation During TestingThis telephone notification, as allowed by 10 CFR 50.73(a)(1), is being made pursuant to 10 CFR 50.73(a)(2)(iv)(A) to describe an unplanned, invalid actuation of containment isolation valves in more than one system which occurred during the most recent refueling outage at Fermi 2. On 3/24/2017, at approximately 1548 EDT, when synchronizing an emergency diesel generator (EDG) to the grid during testing, an electrical perturbation occurred. Further investigation found that the EDG was slightly out of phase when it was attempted to be synchronized to the grid. The electrical perturbation resulted in an unexpected half-scram of Reactor Protection System (RPS) A and actuation (closure) of some containment isolation valves. The actuations were invalid as they were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation. Fermi 2 was shut down for a refueling outage at the time, and therefore, the half-scram of RPS A occurred after the safety function had already been completed. Containment isolation valves actuated (closed) in Division 1 of the Torus Water Management, Drywell Pneumatics, and Drywell Floor and Equipment Drain Sumps systems. All valves operated as expected. Since containment isolation valves in more than one system were actuated by this perturbation, this event constitutes an event or condition that resulted in manual or automatic actuation of the system listed in paragraph 10 CFR 50.73 (a)(2)(iv)(B)(2) and is reportable under 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been informed of this notification.
ENS 528393 July 2017 14:15:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Scram on Low Condenser VacuumAt 1015 (EDT) a manual reactor scram was inserted due to degrading main condenser vacuum. All rods inserted into the core as expected and all systems functioned as expected during the scram. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' At 1033 (EDT) an automatic reactor scram occurred on low reactor water level. Due to the previous manual reactor scram, all rods were (already) inserted. This event is reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section, except when the actuation results from and is part of a preplanned sequence during testing or reactor operation. (1) Reactor protection system (RPS) including : Reactor scram and reactor trip.' Decay heat is being removed using main feedwater and the turbine bypass valves. The licensee notified the NRC Resident Inspector. This event was characterized as a "configuration control event" where a valve misposition allowed the offgas line to flood.
ENS 5280317 April 2017 15:20:0010 CFR 50.73(a)(1), Submit an LER60-Day Optional Telephonic Notification of Invalid Specified System ActuationOn April 17, 2017 at 1120 EDT, following scheduled maintenance, the Reactor Protection System (RPS) 'A' bus was returned to its normal supply, the RPS 'A' motor generator (MG) set. The RPS MG set had been running loaded for 1 hour when the RPS 'A' bus tripped. Maintenance personnel had connected probes of a grounded oscilloscope to check for proper operation of the MG set, resulting in the RPS 'A' bus trip. The controlling procedure did not contain a caution about using only an ungrounded oscilloscope. The trip of the RPS 'A' MG set resulted in a half scram and an invalid isolation signal causing primary containment isolation valves in multiple systems to isolate. This event is reportable per 10 CFR 50.73 (a)(2)(iv)(A) since the containment isolation was not part of a pre-planned sequence and the event resulted in the invalid actuation of general containment isolation valves in more than one system. Corrective actions include revising the governing procedure with the proper precaution and limitation to require the use of an ungrounded oscilloscope. A training needs analysis will also be performed to cover the lessons learned from this event. The licensee notified the NRC Resident Inspector.
ENS 5276419 May 2017 06:06:0010 CFR 50.72(b)(3)(iv)(A), System ActuationValid Reactor Protection System Actuation While ShutdownPursuant to 10CFR50.72(b)(3)(iv)(A), notifications are being performed for a valid actuation of the reactor protection system resulting in a full scram. The actuation was a result of pre-startup testing. The generator coastdown protective relay was left in service which needed to be bypassed to facilitate the testing. This resulted in a reactor scram occurring. The reactor was subcritical with all rods inserted at the time of the actuation. All systems functioned as designed. The licensee notified the NRC Resident Inspector.
ENS 5272730 April 2017 22:18:0010 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Open Bypass Valve Causes Loss of Safety FunctionOn April 30, 2017, at 1818 (EDT), the main turbine steam bypass valve #1 partially opened. Power was incrementally lowered. While lowering power the bypass valve would shut and then reopen and power would again be lowered. When power was lowered to approximately 74 percent the bypass valve remained closed. During the transient the reactor protection system (RPS) Turbine Stop Valve Closure and Control Valve Fast Closure trip functions were declared inoperable due to the opening of the bypass valve which affects the bypass setpoint for those RPS trip functions. With the loss of these RPS trip functions a loss of safety function existed intermittently for approximately 37 minutes. The manual reactor trip function and other RPS functions remained operable. Both channels of the rod withdrawal limiter (RWL) and the end of cycle reactor recirculation pump trip (EOC-RPT) function were also declared inoperable. These functions are credited in accident analysis, this also resulted in a loss of safety function. Currently the bypass valve is closed and the RWL, EOC-RPT and RPS function are operable. Troubleshooting continues to determine the issue with the main turbine that caused the bypass valve to open. NRC Resident Inspector has been notified.
ENS 5270021 April 2017 03:45:0010 CFR 50.72(b)(3)(iv)(A), System ActuationValid Reactor Protection System Actuation While ShutdownAt 2345 (EDT) on 04/20/2017, the Unit 1 Reactor Mode Switch was taken to the Shutdown position to comply with Technical Specification 3.10.4 due to having no operable IRM's (Intermediate Range Monitors) in one quadrant of the reactor vessel as a result of maintenance activities. Placing the mode switch to Shutdown inserts a valid scram signal into the Reactor Protection System (RPS). All control rods had been previously inserted and no rod movement occurred when the mode switch was positioned to Shutdown. Due to this valid RPS scram, and not being a part of a preplanned evolution, this condition is reportable under criteria 50.72(b)(3)(iv)(A) as an event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The licensee notified the NRC Resident Inspector.
ENS 5269620 April 2017 07:02:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor Scram During StartupOn 04/20/2017 at 0302 EST during a reactor startup, a reactor scram resulted from upscale spike on two Intermediate Range Monitors (IRMs), 1C51K601A and 1C51K601B. IRM A, 1C51K601A is in Reactor Protection System Channel A and IRM B, 1C51K601B is in Reactor Protection System Channel B. All control rods fully inserted. No PCIS (Primary Containment Isolation System) actuations occurred and none were expected to occur based upon plant condition following the reactor scram. Investigation is in progress. Condition was not due to a true flux event. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) as an event or condition that resulted in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. CR 10356172 The NRC Resident has been notified. The reactor was at 0.5% (percent) power at the time of the event and will remain in Hot Shutdown pending the results of the root cause investigation.
ENS 5268215 April 2017 09:41:0010 CFR 50.72(b)(3)(iv)(A), System ActuationValid Reactor Protection System and Partial Primary Containment Isolation System Actuations on Low Water LevelDuring shutdown activities with the reactor subcritical, actions were being taken to remove 11 Reactor Feed Pump from service in support of a scheduled refueling outage. Reactor Water Level on Safeguards level instrumentation dropped below +9 inches, which resulted in a valid Reactor Protection System (RPS) Scram signal and Partial Group 2 Primary Containment Isolation System (PCIS) signal. All systems functioned as required. Reactor Water Level on Safeguards instrumentation was restored to greater than +9 inches immediately. RPS and PCIS logic was reset. There was no impact to the health and safety of the public as a result of this event. This actuation of these systems is being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of any of the systems listed in 10 CFR 50.72(b)(3)(iv)(B). The NRC Resident Inspector has been notified.
ENS 5264829 March 2017 23:44:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Scram Initiated During StartupAt 1844 CDT on 3/29/2017, Unit 2 initiated a manual scram due to multiple rods inserting. At 1842 during Unit 2 start-up, Intermediate Range Monitor (IRM) 'G' drifted low. The operator adjusted the range down one position with no immediate reaction. At 1844, a spike on IRM 'G' caused a half scram on Reactor Protection System (RPS) 'A' trip system. The half scram was being reset after evaluating no trip condition was present. As the operator reset groups 2 and 3, a trip signal from IRM 'F' was received on the RPS 'B' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 2-AOI-100-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment Isolations Systems did not receive an actuation signal and performed as designed. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) Reactor Protection System(RPS) including reactor scram and reactor trip'. This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 5262520 March 2017 12:13:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Trip as a Result of Secondary Plant TransientOn March 20, 2017 at 0813 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 operations personnel manually tripped the plant from approximately 91 percent power based on lowering steam generator levels. Prior to the plant trip, the 2A Hotwell pump tripped at 0758 EDT and the 2C Condensate Booster Pump subsequently tripped at 0802 EDT. Operations personnel commenced to lower plant power after the 2A Hotwell pump trip in an attempt to maintain steam generator levels, but were unable to recover level and manually tripped the unit. All control rods fully inserted and all automatically actuated safety related equipment operated as designed. At 0905 EDT, operations personnel exited the emergency operating instructions after the plant was stabilized. The cause of the event is under investigation. This event is reportable to the NRC within four hours under 10 CFR 50.72(b)(2)(iv)(B) as a result of the actuation of the Reactor Protection System and in eight hours under 10 CFR 50.72(b)(3)(iv)(A) as a result of actuation of the Auxiliary Feedwater system. The licensee notified the NRC Resident Inspector.