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 Discovered dateReporting criterionTitleEvent description
ENS 5393715 March 2019 17:00:0010 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 Turbine Generator TripOn March 15, 2019 at 1300 EDT, Indian Point Unit 2 automatically tripped offline from mode 1 - 100% power operations. Reactor Operators verified the reactor trip and the plant is currently stable in mode 3. All automatic systems functioned as required. The auxiliary feedwater system actuated following the trip, as expected. All control rods fully inserted upon the trip, as expected. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). The unit remains on offsite power in hot standby at normal operating temperature and pressure. Decay heat is being removed from the steam generators via the auxiliary feedwater system and the condensate steam dump valves. Unit 3 remains in mode 6 for a scheduled refueling outage. The licensee notified the NRC Resident Inspector, the local transmission company, and New York State Independent System Operator. The Indian Point Unit 2 automatic trip was caused by the trip of the main generator. The cause of the generator trip is unknown at this time.
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 539062 March 2019 09:17:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Reactor Trip Due to Feedwater Isolation Valve ClosureAt 0317 CST, the Unit 2 Reactor tripped due to Feedwater Isolation Valve (FWIV) 2-04 going closed. All Auxiliary Feedwater Pumps started due to steam generator Lo-Lo levels. Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007B. The Emergency Response Guideline Procedure Network has been exited. Decay heat is being rejected to the Main Condenser via the Steam Dump Valves. The cause of the FWIV going closed is currently under investigation. All control rods fully inserted and the reactor trip was uncomplicated. Unit 2 is in a normal post-trip electrical line-up. There was no impact on Unit 1 due to the Unit 2 reactor trip. The licensee notified the NRC Resident Inspector.
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 538133 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 Due to Cycling of Turbine Governor ValveAt 2028 (EST) on January 3, 2019, with the reactor at 85% power, the reactor was manually tripped due to cycling of Turbine Governor Valve #4. The trip was uncomplicated with all systems responding normally following the trip. Investigation of the cause of the valve cycling 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 atmospheric dump valves. This condition has no impact to the health and safety of the public. The licensee notified the NRC Senior Resident Inspector. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A)."
ENS 5370329 October 2018 04:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Trip Due to Inadequate Feedwater FlowOn October 29, 2018 at 1317 EDT, with St. Lucie Unit 1 in Mode 1 at 100% power, the reactor was manually tripped due to inadequate feedwater flow to both 1A and 1B Steam Generators (S/Gs). The trip was uncomplicated with all systems responding normally post-trip. (All control rods fully inserted and there were no specified system actuations.) Operators responded and stabilized the plant in Mode 3. The cause of the inadequate feed flow to the 1A and 1B Steam Generators is currently under investigation. Decay Heat removal is being accomplished through forced circulation with stable conditions from Main Feedwater and the Steam Bypass Control System to the Main Condenser. Currently maintaining Pressurizer pressure at 2250 psia and Reactor Coolant System temperature at 532 degrees F. St. Lucie Unit 2 was unaffected and remains in Mode 1 at 100% power. This report is submitted in accordance with 10CFR50.72(b)(2)(iv)(B) for the reactor trip. The NRC Senior Resident Inspector has been notified."
ENS 5369727 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
Manual Reactor Trip Following StartupOn October 27, 2018, at 1533 EDT, Watts Bar Nuclear (WBN) Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. (Main Steam Isolation Valves) MSIVs were required to be isolated due to cooldown. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Generator Atmospheric Dump Valves. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident has been notified."
ENS 5366512 October 2018 04:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationAutomatic Reactor TripOn October 12, 2018 at 1353 EDT, St. Lucie Unit 2 experienced an automatic RPS actuation and Reactor Trip due to a fault on the 2A1 6.9kv bus during a transfer of the bus power supply from the 2A Auxiliary Transformer to the 2A Startup Transformer. The bus fault caused a fire in the 2A1 6.9kv switchgear that has been extinguished. Offsite support was not required to extinguish the fire. The specific cause of the fault is currently under investigation. Following the reactor trip, both Steam Generators are being supplied by main feedwater. All (Control Element Assemblies) (CEAs) fully inserted into the core. Decay Heat removal is being accomplished through forced circulation. Main Feedwater and Steam Bypass Control Systems are maintaining stable conditions in Mode 3. St. Lucie Unit 1 was unaffected and remains in Mode 1 at 100 percent power. This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the Reactor Trip. The fire was extinguished within 28 minutes. Plant loads are being supplied by the 2B Auxiliary Transformer. The licensee notified the NRC Resident Inspector.
ENS 536528 October 2018 05:00:0010 CFR 50.72(b)(3)(ii)(A), Seriously DegradedReactor Coolant System Pressure Boundary Leakage

On Monday, October 8, 2018 at 0111 CDT, during the initial containment entry for unit 2 refueling outage (A2R20), reactor coolant system pressure boundary leakage was discovered at the 2D Steam Generator bowl drain line. Unit cooldown to mode 5 is in progress.

This event is reportable under 10CFR50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.'

The licensee has notified the NRC resident inspector. Approximately 0.1 gpm was leaking from the drain line. LCO 3.4.13 was entered and the licensee anticipates being in mode 5 within a couple of hours. The leak will be repaired prior to exiting the refueling outage.

ENS 5361118 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
Manual Reactor Trip Due to a Steam Leak on a High Pressure Feedwater HeaterDue to a steam leak on the reheater line to 36C Feedwater Heater, operators initiated a manual trip of the reactor, verified the reactor trip, and closed all Main Steam Isolation Valves. The plant is currently stable in Mode 3 with the steam leak isolated. The Auxiliary Feedwater System actuated following the trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). The unit remains on offsite power in Hot Standby at normal operating temperature and pressure. Decay heat is being removed from the steam generators via the Auxiliary Feedwater System and atmospheric steam dumps. Unit 2 was unaffected and remains at 100 percent power. The licensee notified the NRC Resident Inspector, the State of New York, and the local transmission company.
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 5357531 August 2018 07:00:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessUnplanned Loss of Steam Line Monitor ChannelsThe 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 10 CFR 50.73. This event is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) as a Loss of Emergency Preparedness Capabilities at Palo Verde Nuclear Generating Station (PVNGS) Unit 2. On August 31, 2018, at approximately 0544 Mountain Standard Time (MST), the Unit 2 control room experienced an unplanned loss of Steam Generator #1 steam line monitor (RU-139), Channels A and B. This main steam line monitor is used in the PVNGS Emergency Plan to perform dose assessment in the event of a steam generator tube rupture. The NRC Resident Inspectors have been notified."
ENS 5355013 August 2018 05:00:0010 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 Generator LockoutAt 23:58 (Central Daylight Time) Unit 2 Reactor Tripped (automatic reactor trip) due to a Turbine Trip/ Generator Lock Out. All Auxiliary Feedwater Pumps started due to steam generator Lo Lo levels. Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IP0-007B. The Emergency Response Guideline Procedure Network has been exited. Decay heat is being rejected to the Main Condenser via Steam Dump Valves. The cause of the Generator Lockout is currently under investigation. All control rods fully inserted in response to the automatic reactor trip. The licensee notified the NRC resident.
ENS 5352223 July 2018 07:00:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessUnplanned Loss of Steam Line Monitor ChannelsThe 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 (Emergency Notification System) or under the reporting requirements of 10CFR50.73. This event is being reported pursuant to 10CFR50.72(b)(3)(xiii) as a Loss of Emergency Preparedness Capabilities at Palo Verde Nuclear Generating Station (PVNGS) Unit 2. On July 23, 2018, at approximately 1631 Mountain Standard Time (MST), the Unit 2 control room experienced an unplanned loss of Steam Generator #1 steam line monitor (RU-139), channels A and B. This monitor is used to assess dose projections for Main Steam line exhaust while in Modes 1-4 and is used in the PVNGS Emergency Plan to perform classification of Initiating Conditions 'RS1' and' RG1' and Emergency Action Levels (EALs) 'RS1.2' and 'RG1.2'. The PVNGS Emergency Plan does have two additional EALs that can be assessed for each Initiating Condition. The loss of this monitor constitutes a reportable loss of emergency assessment capability. The NRC Resident Inspector has been informed of this condition."
ENS 534843 July 2018 04:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Trip Due to High Steam Generator Water LevelAt 0954 (EDT) on July 3, 2018, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to high steam generator water level. 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 lines through the steam dumps and into the condenser. 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). 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 and Unit 1 is in an electrical shutdown lineup. The cause of the high steam generator water level transient is being investigated.
ENS 5347727 June 2018 07:00:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Trip on Low Steam Generator Water LevelThe 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 (Emergency Notification System) or under the reporting requirements of 10 CFR 50.73. On June 27th, 2018 at approximately 2310 Mountain Standard Time (MST), in Palo Verde Unit 3, the #1 Steam Generator Economizer valve started closing. This caused Steam Generator #1 water level to decrease. Both Feed water pumps speed increased to raise Steam Generator level. At approximately 2311 (MST), the B Main Feed water pump tripped resulting in a Reactor Power Cutback. Steam Generator #1 level continued to decrease resulting in an Automatic Reactor Trip on Low Steam Generator #1 water level. All control rods inserted to shut down the Reactor to Mode 3 using Main Feed water and Steam Bypass. Post trip Steam Generator #1 level then increased and at approximately 2316 (MST) a Main Steam Isolation Signal (MSIS) was received on high Steam Generator level. The 'B' Auxiliary Feed water pump was manually started to maintain Steam Generator water levels and Steam Generator pressure was controlled using the Atmospheric Dump Valves (ADVs). Following the reactor trip, all CEAs (Control Element Assemblies) inserted fully into the core. All systems operated as expected. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and 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 3 is stable and in Mode 3 feeding Steam Generators with Auxiliary Feed water Pump 'B'. 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 was informed of the Unit 3 reactor trip. Unit 1and Unit 2 were unaffected by the Unit 3 trip.
ENS 5347226 June 2018 05:00:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessMain Steam Line Radiation Monitor Determined to Be Non-FunctionalAt time 0003 (CDT), Main Steamline Radiation Monitor 2-RE-2326 (Main Steam line 2-02) reading spiked and (was) declared non-functional. With this radiation monitor non-functional, all of the emergency action levels for a steam generator tube rupture in steam generator 2-02 could neither be evaluated nor monitored. This unplanned condition is reportable as a loss of assessment capability per 10CFR50.72(b)(3)(xiii). Comanche Peak Nuclear Power Plant (CPNPP) has assurance of steam generator integrity and fuel cladding integrity and there is negligible safety significance to the current condition (with respect to the) public health and safety perspective. Additionally, compensatory measures are in place to assure adequate monitoring capability is available to implement the CPNPP emergency plan in the unlikely event of challenges to the steam generator or fuel cladding. The N16 radiation monitor serves as a backup with alarm function and Radiation Protection technicians have been briefed on taking local readings with a Geiger-Mueller tube on MSL 2-02. Corrective actions are being pursued to restore 2-RE-2326 to functional status. The NRC Resident Inspector has been notified."
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 534434 June 2018 14:20:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Trip on Lowering Steam Generator Water LevelAt 0920 CDT, Braidwood Unit 1 reactor was manually tripped due to lowering steam generator water levels following a trip of the 1C main feedwater pump. The cause of the 1C main feedwater pump trip is unknown at this time and is under investigation. Both trains of Braidwood Unit 1 auxiliary feedwater started automatically following the reactor trip to maintain steam generator water levels. All systems responded as expected. Steam generator power operated relief valves lifted momentarily and reseated as designed in response to the secondary transient due to the reactor trip. The main steam dump valves are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by offsite power with the diesel generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr. notification, and per 10CFR50.72(b)(3)(iv)(A) for an automatic actuation of the auxiliary feedwater system, 8-hr. notification. All rods inserted into the core during the trip. Concerning the relief valves momentarily lifting and reseating, there is no known primary-to-secondary leakage. The licensee has notified the NRC Resident Inspector.
ENS 5342323 May 2018 13:48:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessLoss of Emergency Assessment Capability

At time 0848 (CDT), Main Steamline Radiation Monitor 2-RE-2328 (Main Steamline 2-04) lost communications and was declared non-functional.

With this radiation monitor non-functional, all of the emergency action levels for a steam generator tube rupture in steam generator 2-04 could neither be evaluated nor monitored. This unplanned condition is reportable as a loss of assessment capability per 10CFR50.72(b)(3)(xiii). Comanche Peak Nuclear Power Plant (CPNPP) has assurance of steam generator integrity and fuel cladding integrity and there is a negligible safety significance to the current condition from a public health and safety perspective. Additionally, compensatory measures are in place to assure adequate monitoring capability is available to implement the CPNPP emergency plan in the unlikely event of challenges to the steam generator or fuel cladding. The N16 radiation monitor serves as a backup with alarm function and Radiation Protection technicians have been briefed on taking local readings with a Geiger-Mueller tube on MSL 2-04. Corrective actions are being pursued to restore 2-RE-2328 to a functional status. The NRC Resident Inspector has been notified.

ENS 533887 May 2018 18:35:0010 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor
10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat
10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident
Discovery of a Condition That Could Have Prevented Fulfillment of a Safety FunctionOn May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power. During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This becomes an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop will flash to steam and prevent RCS depressurization. In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaway's atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation. This issue has been analyzed by Westinghouse in WCAP-16632-P. This WCAP determined that to prevent loop stagnation, the RCS cooldown rate in these conditions should be limited to a rate dependent on the temperature differential present in the active loops. The WCAP analysis was used to support a revision to the generic Emergency Response Guideline (ERG) for ES-0.2 "Natural Circulation Cooldown." Figure 1 in ES-0.2 provides a curve of the maximum allowable cooldown rate as a function of active loop temperature differential which is directly proportional to the level of core decay heat. At the time of discovery of this condition, Callaway's EOP structure did not ensure that the ES-0.2 guidance would be implemented for a natural circulation cooldown with a stagnant loop. Callaway has issued interim guidance to the on-shift personnel regarding this concern and is in the process of revising the applicable EOPs. This condition is reportable per 10 CFR 50.72(b)(3)(v) for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to (A) Shutdown the reactor and maintain it in a safe shutdown condition, (B) Remove residual heat, or (D) mitigate the consequences of an accident." The licensee notified the NRC Resident Inspector of this condition.
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 5337130 April 2018 16:24:0010 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 TripAt 1124 CDT, Braidwood Unit 1 experienced an automatic Reactor Trip. The cause of the Reactor Trip was a Turbine Trip with reactor power greater than P-8. The turbine trip was actuated as a result of a Turbine Motoring Generator Trip. The cause of the generator trip is unknown at this time and is under investigation. After the Reactor Trip occurred, the 1A Auxiliary Feedwater pump was manually started to provide feedwater flow to all four steam generators. The 1A Auxiliary Feedwater pump was subsequently secured and placed in standby when the Startup Feedwater pump was placed in service. Train A Main Control Room Ventilation Filtration system shifted to Makeup Mode due to a spurious actuation signal. No secondary relief valves lifted and no secondary steam was released as a result of the Reactor Trip. The Main Steam dump valves are in service to the Main Condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by Offsite Power with the Diesel Generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr notification, and per 10 CFR 50.72(b)(3)(iv)(A) for a manual actuation of the Auxiliary Feedwater system, 8-hr notification. The licensee notified the NRC Resident Inspector and Illinois Emergency Management Agency.
ENS 5334820 April 2018 01:07:0010 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 Turbine IssueWhile performing a turbine startup, a turbine control anomaly caused a steam generator level transient. The rise in steam generator level above the setpoint caused the turbine to automatically trip. The high steam generator level of 73 percent caused a feedwater isolation signal at 2107 EDT, which also tripped both Main Boiler Feed Pumps. The tripping of the Main Boiler Feed Pumps auto started the motor driven Aux Boiler Feed Pumps 21 and 23. The reactor was manually tripped at 2108 EDT in accordance with AOP-FW-1 Loss of Main Feedwater. All control rods inserted. Electrical power is being provided from offsite via the Station Aux Transformer. Decay heat removal is being provided via the Atmospheric Dump Valves. An investigation into the cause of the turbine control anomaly is underway. The NRC Resident Inspector has been notified. The event did not have an affect on Unit 3 and there is no primary to secondary leakage.
ENS 5332913 April 2018 06:27:0010 CFR 50.72(b)(2)(iv)(B), RPS System ActuationManual Reactor Trip Following Main Feedwater Control ProblemOn 4/13/2018 at 0227 (EDT), the Oconee Unit 1 Reactor was manually tripped from 24 percent power due to the inability to control main feedwater flow through the Main Feedwater Control Valves using the Integrated Control System. Due to the RPS actuation while critical, this event is being reported as a 4-hour Non-Emergency per 10 CFR 50.72(b)(2)(iv)(B). Following the reactor trip, multiple Main Steam Relief Valves failed to reseat at the expected pressure. Using procedure guidance, Main Steam Pressure was lowered by 115 psig, resulting in the closing of all Main Steam Relief Valves. All other post-trip conditions are normal and all other systems performed as expected. Unit 1 is currently in Mode 3 and stable. Decay heat is being removed by the steam generators discharging steam to the main condenser using the turbine bypass valves. Units 2 and 3 are not affected by the Unit 1 reactor trip. The licensee notified the NRC Resident Inspector.
ENS 5329026 March 2018 18:38:00Other Unspec ReqmntTechnical Specifications Required Shutdown Due to Inoperable Main Steam Isolation Valve

On March 25, 2018 at 1833 CDT, while at 100 percent power, Farley Unit 1 (FNP-1) conservatively declared a single Main Steam Isolation Valve (MSIV) inoperable on the 1C Steam Generator line due to indication of Steam Generator pressure rise with a corresponding reduction in flow of that loop. FNP-1 began a reactor shutdown at 0400 CDT on March 26, 2018 to establish plant conditions to support testing the affected main steam line MSIVs while in the required action time of Technical Specification 3.7.2. At 1338 CDT on March 26, 2018, testing confirmed that the single MSIV was inoperable and that valve disassembly will be required. The duration of the valve repair would exceed the required action time of Technical Specification 3.7.2. This report is being made in accordance with 10 CFR 50.72(b)(2)(i), as a plant shutdown required by technical specifications. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM DOUGLAS HOBSON TO KEN MOTT AT 0202 EDT ON 5/16/18 * * *

This EN (event notification) is being updated to clarify the reporting criteria as 'Voluntary'. Farley Technical Specification 3.7.2 allows continuous operation in MODE 2 with an INOPERABLE MSIV as long as the other MSIV in the affected Main Steam Line is closed. The initiation of the shutdown was performed as a prudent action to repair and restore OPERABILITY of the affected MSIV and was not a requirement of the Farley Technical Specifications. The licensee notified the NRC Resident Inspector. The R2DO (Masters) was 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 5321616 February 2018 07:01:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Trip Due to Loss of Excitation on Main GeneratorOn February 16, 2018 at 02:01 EST Indian Point Unit 3 automatically tripped on a turbine trip due to a loss of main generator excitation. All control rods fully inserted and all plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv)(B). The auxiliary feedwater system actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. The plant is stable, in Mode 3, at no load operating temperature and pressure. Decay heat removal is via the steam generators to the main condenser via the condenser steam dumps. No radiation was released. Indian Point Unit 2 was unaffected by this event and remains at 100 percent power. A post trip investigation is in progress. The licensee has notified the NRC Resident Inspector.
ENS 5321215 February 2018 09:06:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessMain Steam Line Radiation Monitor Determined to Be Non FunctionalAt time 0306 (CST), Main Steamline Radiation Monitor 2-RE-2326 (Main Steamline 2-02) reading spiked and declared non-functional. With this radiation monitor non-functional, all of the emergency action levels for a steam generator tube rupture in steam generator 2-02 could neither be evaluated nor monitored. This unplanned condition is reportable as a loss of assessment capability per 10CFR50.72(b)(3)(xiii). Comanche Peak Nuclear Power Plant (CPNPP) has assurance of steam generator integrity and fuel cladding integrity and there is a negligible safety significance to the current condition from a public health and safety perspective. Additionally, compensatory measures are in place to assure adequate monitoring capability is available to implement the CPNPP emergency plan in the unlikely event of challenges to the steam generator or fuel cladding. The N16 radiation monitor serves as a backup with alarm function and Radiation Protection technicians have been briefed on taking local readings with a Geiger-Mueller tube on MSL 2-02. Corrective actions are being pursued to restore 2-RE-2326 to functional status. The NRC Resident Inspector has been notified.
ENS 531546 January 2018 17:26:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessMain Steam Line Radiation Monitor Determined to Be Non-FunctionalAt 1126 (CST), main steamline radiation monitor 2-RE-2326 (Main Steamline 2-02) reading was determined to be erratic and was declared non-functional. With this radiation monitor non-functional, all of the emergency action levels for a steam generator tube rupture in steam generator 2-02 could neither be evaluated nor monitored. This unplanned condition is reportable as a loss of assessment capability per 10 CFR 50.72(b)(3)(xiii). Comanche Peak Nuclear Power Plant (CPNPP) has assurance of steam generator integrity and fuel cladding integrity and there is a negligible safety significance to the current condition from a public health and safety perspective. Additionally, compensatory measures are in place to assure adequate monitoring capability is available to implement the CPNPP emergency plan in the unlikely event of challenges to the steam generator or fuel cladding. The N16 (Nitrogen-16) radiation monitor serves as a backup with alarm function and Radiation Protection technicians have been briefed on taking local readings with a Geiger-Mueller tube on MSL (Main Steam Line) 2-02. Corrective actions are being pursued to restore 2-RE-2326 to functional status. The NRC Resident Inspector has been notified.
ENS 531056 December 2017 02:00:0010 CFR 50.72(b)(3)(xiii), Loss of Emergency PreparednessUnplanned Loss of Multiple Radiation Monitors During MaintenanceAt 2000 (CST), Comanche Peak experienced a failure of SCADA B of the PC11 Radiation Monitor System. This failure caused a loss of Unit 1 Main Steam Line 1-01 and 1-03 Radiation Monitors (1-RE-2325 and 1-RE-2327) and Train A and Train B Station Service Water Radiation Monitors (1-RE-4269 and 1-RE-4270). With the Main Steam Line Radiation Monitors nonfunctional, all of the emergency action levels for a steam generator tube rupture in steam generators 1-01 and 1-03 could neither be evaluated nor monitored. With the Station Service Water Radiation Monitors non-functional, all of the emergency action levels for a radioactive release through station service water could neither be evaluated nor monitored. This unplanned condition is reportable as a loss of assessment capability per 10 CFR 50.72(b)(3)(xiii). Comanche Peak Nuclear Power Plant (CPNPP) has assurance of steam generator integrity, reactor coolant system integrity, and fuel cladding integrity and there is a negligible safety significance to condition from a public health and safety perspective. Additionally, compensatory measures are in place to assure adequate monitoring capability is available to implement the CPNPP emergency plan in the unlikely event of challenges to the steam generator, reactor coolant system, or the fuel cladding. Until these radiation monitors were restored, Operations implemented compensatory measures to monitor the Condenser Off Gas Radiation Monitor for early signs of a steam generator tube leak/rupture and Radiation Technicians were briefed on taking local readings with a Geiger-Mueller tube on the Main Steam Lines. Chemistry Technicians were performing hourly samples of Station Service Water and reporting results to the Control Room. Corrective actions were pursued to restore the non-functional radiation monitors back to service. Those actions are complete and all radiation monitors have been restored to service. The NRC Resident Inspector has been notified. PC11 is a computer common to both Units. The failure happened during radiation monitor maintenance to a single monitor, which unexpectedly affected multiple monitors.
ENS 5309126 November 2017 02:25:0010 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 Main Feed WaterAt time 2025 (CST) on 11/25/17, Unit 2 reactor was manually tripped due to a loss of all Main Feedwater. Operators observed both Main Feed Pumps tripped and SG (Steam Generator) levels decreasing, resulting in the direction for a manual reactor trip. The reactor trip actuated a turbine trip, both Motor Driven Auxiliary Feedwater Pumps started on the loss of both Main Feed Pumps, and Steam Generator Lo Lo levels started the Turbine Driven Auxiliary Feedwater Pump. All systems responded as expected. There was no work in progress at the time of the incident. Currently Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IP0-0078 and the Emergency Response Guideline Procedure Network has been exited. Decay Heat is being rejected to the Main Condenser via Steam Dump Valves. The licensee has notified the NRC Resident Inspector.
ENS 530608 November 2017 00:57:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Trip on Turbine Trip

On 11/7/2017 at 1957 (EST), VC Summer Nuclear Station automatically tripped due to a turbine trip. The cause of the turbine trip is under investigation at this time. All systems responded as expected. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The unit is currently stable in Mode 3 with decay heat removal via the Main Steam to the Main Condenser. The NRC Resident Inspector has been notified. The licensee will notify the South Carolina State Emergency Management Division, the Fairfield, Richland, Lexington and Newberry Counties.

  • * * UPDATE FROM BETH DALICK TO VINCE KLCO ON 11/8/17 AT 1409 EST * * *

All systems responded as expected, with the exception of 'B' Steam Generator Feedwater Isolation Valve XVG1611 B-FW. This valve did not appear to automatically close and was slow to indicate closed from the Main Control Board. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. Notified the R2DO (Musser).

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 5303626 October 2017 06:12:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Reactor Trip Following a Loss of LoadOn October 26, 2017 at 0212 EDT St. Lucie Unit 2 experienced a reactor trip due to a loss of load event resulting in an RPS (Reactor Protection System) actuation. The cause of the loss of load is currently under investigation. Following the reactor trip, an Auxiliary Feedwater Actuation Signal occurred due to low level in the 2A Steam Generator. One of the two Main Feed Isolation Valves to the 2A Steam Generator did not close on the Auxiliary Feedwater Actuation Signal. 2A Steam Generator level was restored by Auxiliary Feedwater. The 2B Steam Generator level is being maintained by Main Feedwater. All CEAs (Control Element Assemblies) fully inserted into the core. Decay heat removal is being accomplished through forced circulation with stable conditions from Auxiliary Feedwater/Main Feedwater and Steam Bypass Control System. Currently maintaining pressurizer pressure at 2250 psia and Reactor Coolant System temperature at 532 degrees F. St. Lucie Unit 1 was unaffected and remains in Mode 1 at 100 percent power. This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72(b)(3)(iv)(A) for the Specified System Actuation. The licensee notified the NRC Resident Inspector.
ENS 5296010 September 2017 22:55:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Manual Reactor Trip on Lowering Steam Generator Water LevelOn 09/10/17 at 1855 (EDT), (Turkey Point) Unit 4 reactor was manually tripped from 88% RTP (Rated Thermal Power) due to a failure of 4C Steam Generator main feed regulating valve causing lowering S/G (Steam Generator) level. All other systems operated normally. Auxiliary Feed Water initiated as designed to provide S/G water level control. EOP's (Emergency Operating Procedures) have been exited and General Operating procedures (GOP'S) were entered. Unit 4 is stable in Mode 3 at NOT/NOP (Normal Operating Temperature/Normal Operating Pressure). The licensee is investigating the failure of the feed regulating valve. Offsite power is available. Decay heat is being removed via main feedwater with steam discharged to atmosphere using the ADVs (Atmospheric Dump Valves). There is no known primary-secondary steam generator tube leakage. The licensee informed the NRC Resident Inspector.
ENS 5289811 August 2017 16:24:0010 CFR 50.72(b)(3)(iv)(A), System ActuationComanche Peak Automatic Turbine Trip from 10 Percent PowerAt 1124 CDT on 11 August 2017, CPNPP (Comanche Peak Nuclear Power Plant) Unit 2 experienced an automatic turbine trip and trip of both main feedwater pumps on high steam generator water level (P-14, 81.5 percent level) in steam generator 2-02. Following the turbine trip, the auxiliary feedwater system actuated as required. The plant was stabilized at 2-3 percent reactor power with auxiliary feedwater feeding all steam generators with all levels within their normal bands. The cause of the high steam generator level appears to be a mechanical malfunction of steam generator 2-02 flow control valve bypass valve 2-LV-2163 (SG 2-02 FW BYP CTRL VLV) to close when demanded. Troubleshooting and repair of 2-LV-2163 is in progress. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) for an actuation of auxiliary feedwater. The NRC Resident Inspector has been notified.
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 5286317 July 2017 21:17:0010 CFR 50.72(a)(1)(i), Emergency Class Declaration
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)(A), Loss of Safety Function - Shutdown the Reactor
Unusual Event Declared Due to Loss of Offsite Power

During a rain and lightning storm, plant operators observed arcing from the main transformer bus duct and notified the control room. The decision was made to trip the main generator which resulted in an automatic reactor trip. The plant entered EAL SU.1 as a result of the loss of offsite power for greater than fifteen minutes. Plant safety busses are being supplied by both emergency diesel generators while the licensee inspects the electrical system to determine any damage prior to bringing offsite power back into the facility. Offsite power is available to the facility. No offsite assistance was requested by the licensee. During the trip, all rods inserted into the core. Decay heat is being removed via the atmospheric dump valves with emergency feedwater supplying the steam generators. The main steam isolation valves were manually closed to protect the main condenser. There were no safeties or relief valves that actuated during the plant transient. There is no known primary-to-secondary leakage. Reactor cooling is via natural circulation. All safety equipment is available for the safe shutdown of the plant. The licensee has notified the NRC Resident Inspector, Louisiana Department of Environmental Quality and the local Parish emergency management agencies. Notified DHS SWO, FEMA, DHS NICC, FEMA National Watch Center (email) and Nuclear SSA (email).

  • * * UPDATE ON 7/17/17 AT 2007 EDT FROM MARIA ZAMBER TO DONG PARK * * *

This notification is also made under 10 CFR 50.72(b)(3)(v)(D). This is a non-emergency notification from Waterford 3. On July 17, 2017 at 1606 CDT, the reactor automatically tripped due to a loss of Forced Circulation, which was the result of Loss of Offsite Power (LOOP) to the electrical (safety and non-safety) buses. Both 'A' and 'B' trains of Emergency Diesel Generators (EDGs) started as designed to reenergize the 'A' and 'B' safety buses. The LOOP caused a loss of feedwater pumps, resulting in an automatic actuation of the Emergency Feedwater (EFW) system. Prior to the reactor trip, at 1600 CDT, personnel noticed the isophase bus duct to main transformer 'B' glowing orange due to an unknown reason. Due to this, the main turbine was manually tripped at 1606 CDT. Following the turbine trip, the electrical (safety and non-safety) buses did not transfer to the startup transformers as expected due to an unknown reason. The plant entered the Emergency Operating Procedure for LOOP/Loss of Forced Circulation Recovery. At 1617 CDT, an Unusual Event was declared due to Initiating Condition (IC) SU1 - Loss of all offsite AC power to safety buses (greater than) 15 minutes. All safety systems responded as expected. The plant is currently in mode 3 and stable with the EDGs supplying both safety buses and with EFW feeding and maintaining both steam generators. Offsite power is in the process of being restored. The licensee has notified the NRC Resident Inspector, Louisiana Department of Environmental Quality and the local Parish emergency management agencies.

  • * * UPDATE FROM ADAM TAMPLAIN TO HOWIE CROUCH AT 2203 EDT ON 7/17/17 * * *

The licensee terminated the Notification of Unusual Event at 2056 CDT. The basis for terminating was that offsite power was restored to the safety busses. The licensee has notified Louisiana Department of Environmental Quality, St. John and St. Charles Parishes, Louisiana Homeland Security Emergency Preparedness, and will be notifying the NRC Resident Inspector. Notified IRD (Stapleton), NRR (King), R4DO (Hipschman), DHS SWO, FEMA, DHS NICC, FEMA National Watch Center (email) and Nuclear SSA (email).

  • * * UPDATE FROM SCOTT MEIKLEJOHN TO HOWIE CROUCH AT 1724 EDT ON 7/19/17 * * *

This update is being reported under 10 CFR 50.72(b)(3)(v)(B). During the event discussed in EN# 52863, at 1642 CDT (on July 17, 2017), Condensate Storage Pool (CSP) level lowered to less than 92% resulting in entry to Technical Specification (TS) 3.7.1.3. Level in the CSP was lowered due to feeding from both Steam Generators with EFW. Normal makeup to the CSP was temporarily unavailable due to the LOOP. Filling the CSP commenced at 1815 CDT (on July 17, 2017), and TS 3.7.1.3 was exited on July 18, 2017 at 0039 CDT. The licensee notified the NRC Resident Inspector. Notified R4DO (Hipschman).

  • * * UPDATE FROM SCOTT MEIKLEJOHN TO HOWIE CROUCH AT 1233 EDT ON 9/14/17 * * *

Waterford 3 is retracting a follow up notification made on July 19, 2017 for EN# 52863, concerning the loss of safety function associated with the Condensate Storage Pool (CSP) per 10 CFR 50.72(b)(3)(v)(B). The Condensate Storage Pool was performing its required safety function by providing inventory to the Emergency Feed Water pumps when the required Tech Spec level (T.S. 3.7.1.3) dropped below 92%. The Technical Specification was entered at 1624 (CDT) on July 17, 2017 and exited after filling at 0039 on July 18, 2017. The total allowed outage time allowed by Tech Spec 3.7.1.3 is 10 hours to be in Hot Shutdown if not restored. The Condensate Storage Pool level was restored to greater than 92% prior to exceeding the allowed outage time. Based on level being restored and the Condensate Storage Pool performing its required safety function, 10 CFR 50.72(b)(3)(v)(B) does not apply. Prior to the automatic reactor trip, Condensate Storage Pool level was greater than 92%. The NRC Resident Inspector has been notified of the retraction. Notified R4DO (Groom).

ENS 5283329 June 2017 12:57:0010 CFR 50.72(b)(2)(iv)(B), RPS System Actuation
10 CFR 50.72(b)(3)(iv)(A), System Actuation
Automatic Trip Due to a Loss of Normal Feedwater Flow to B Steam Generator

On 6/29/2017 at 0857 (EDT), VC Summer Nuclear Station automatically tripped due to a loss of normal feed water flow to the B Steam Generator.

There were no complications with the trip. All control rods fully inserted. All emergency feedwater pumps automatically started and recovered steam generator levels. The plant is stable in Mode 3. Station personnel are investigating the cause of the loss of normal feedwater to the B Steam Generator. This is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. The licensee notified the State of South Carolina as well as Fairfield, Lexington, Richland and Newberry Counties regarding the event.

ENS 5282926 June 2017 04:00:0010 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
Manual Reactor Trip Due to Loss of Main FeedpumpOn June 26, 2017, at 1531 (EDT), Indian Point Unit 2 inserted a manual reactor trip prior to Steam Generator levels reaching the automatic reactor trip setpoint. Steam Generator water level perturbation resulted from a loss of 22 Main Boiler Feed Pump. All Control Rods verified inserted. The Auxiliary Feedwater System started as designed and supplied feedwater to the Steam Generators. Heat removal is via the Main Condenser through the High Pressure Steam Dumps. Offsite power is being supplied through the normal 138kV feeder 95332. The cause of the 22 Main Boiler Feed Pump loss is currently under investigation. Entergy is issuing a press release/news release on this issue. Unit 2 is stable and in Mode 3. There was no impact on Unit 3. The licensee notified the State of New York and the NRC Resident Inspector.
ENS 527916 June 2017 21:51:00Other Unspec ReqmntBoth Standby Steam Generator Feed Pumps Out of Service for Pre-Planned MaintenanceThis notification is in accordance with Turkey Point Technical Specification (TS) 3.7.1.6, Action b.1 to report the inoperability of both Standby Steam Generator Feedwater Pumps (SSGFPs) for greater than 24 hours. On 6/6/17 at 1751 hours (EDT) both SSGFPs will be inoperable for greater than 24 hours in support of planned valve repairs that require isolation of the common suction and discharge piping. TS 3.7.1.6, Action b.1 requires a report within four hours if both SSGFPs have been inoperable for 24 hours providing the cause of the inoperability and restoration plans. The valve repairs are currently planned for completion on or about 2300 hours on 6/9/17 which will restore the 'A' pump to service and allow TS 3.7.1.6, Action b.1 to be exited. The diesel-driven 'B' pump will remain out of service for radiator repair which is currently planned for completion on or about 0800 hours on 6/14/17. The function of the Standby Steam Generator Feedwater System is as a backup to the Auxiliary Feedwater System and is not credited in the safety analysis. The NRC Resident Inspector has been notified. Both pumps were taken out of service at 1751 EDT on 6/5/17.
ENS 5279028 April 2017 09:00:0010 CFR 50.73(a)(1), Submit an LER60-Day Optional Telephonic Notification for an Invalid Auxiliary Feedwater ActuationThis 60-day telephone notification is being submitted in accordance with paragraphs 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A) to report an invalid Auxiliary Feedwater (AFW) actuation at Watts Bar Nuclear Plant (WBN) Unit 1. On April 28, 2017 at approximately 0500 Eastern Daylight Time (EDT), Unit 1 maintenance personnel were performing 1-IMI-3.005, '18 Month Calibration of Anticipated Transient Without Scram System Actuation Circuitry (AMSAC),' when an AMSAC actuation signal was received. Both Motor Driven Auxiliary Feedwater Pumps (MDAFWPs) were already in service when this actuation occurred. The Turbine Driven Auxiliary Feedwater Pump (TDAFWP) did not start as it had been removed from service. Additionally, steam generator blowdown isolated as required. During this event, AMSAC actuation was complete and equipment functioned as expected by its operating state. Upon identification of the AMSAC actuation, maintenance activities were halted and a prompt investigation was initiated. WBN found that the procedure in use for the 18 month calibration had been recently revised. The procedure called for maintenance to connect an analog multi-meter set on resistance to incorrect points during performance of the procedure. When the analog multi-meter was connected to the incorrect points, a relay was energized resulting in an AMSAC actuation. The procedure was revised and the test completed. The licensee notified the NRC Resident Inspector.
ENS 527324 May 2017 21:09:0010 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 Failed Reactor Coolant Pump Power TransferOn May 4th, 2017, at 1709 EDT, Watts Bar Nuclear Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Generator Atmospheric Dump Valves. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72(b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident (Inspector) has been notified.
ENS 5271829 April 2017 22:44:0010 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 Steam Generator Hi-Hi Level Signal and Feedwater IsolationAt 1844 (EDT) on 04/29/2017, while the unit was in a low power condition exiting from a refueling outage, the reactor was manually tripped following a P-14 signal (Steam Generator Hi-Hi Level) and a resulting feedwater isolation signal. All control rods were verified to be fully inserted. The cause of the ('B') steam generator high level is currently being investigated. Emergency feedwater actuated at 1845 due to a low-low water level in steam generator 'D'. Plant equipment response is being evaluated and the plant is stabilized in Mode 3 with decay heat removal through the steam dump system to the condensers. There was no release and the emergency feedwater system has been restored to standby. The 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 10 CFR 50.73. The licensee notified the NRC Resident Inspector.
ENS 5269921 April 2017 01:10:0010 CFR 50.72(a)(1)(i), Emergency Class DeclarationUnusual Event Declared Due to Hydrazine in Containment

At 2110 EDT, Salem control room received data that supported unacceptable levels of hydrazine concentration in the U2 Containment atmosphere based on Site Protection atmospheric sampling. The high hydrazine levels were caused due to Steam Generator (S/G) venting into the Containment atmosphere in support of maintenance for the current Salem Unit 2 Refueling Outage (2R22). The NIOSH habitability limit for hydrazine is 0.03 ppm (2 hour limit). Area samples indicated concentrations as high as 0.25 ppm. Salem Unit 2 Containment has been evacuated while a mitigation plan is being developed. There were no personnel injuries as a result of this occurrence. Salem Unit 2 defueling activities were in progress during this event. All fuel assemblies have been placed in a safe condition. All Salem Unit 2 Containment activities are currently on hold. There has been no impact to the equipment in the Unit 2 Containment, no adverse impact to any equipment located in the vicinity of the high hydrazine concentration, and no operational impact to the plant including Shutdown Cooling which is currently on RHR. The Unusual Event was declared under EAL HU3.1, Toxic/Flammable Gas Release Affecting Plant Operations. The licensee plans to issue a press release. The licensee notified the NRC Resident Inspector, Lower Alloways Creek Township, State of New Jersey and State of Delaware. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATED FROM JOHN COOK TO DONALD NORWOOD AT 1305 EDT ON 4/21/2017 * * *

Salem Unit 2 terminated the Unusual Event at 1258 EDT on 4/21/17. The basis for termination was no longer restricting access to the containment after getting two sets of satisfactory air sample results. With the access restored, normal plant operations can resume and EAL HU3.1 is no longer applicable. The details of the sample results are: Fire Protection performed satisfactory results of no detectable Hydrazine (0.01 ppm with a NIOSH limit of 0.03 ppm) completed both at 1001 EDT and 1247 EDT at the following locations: - (3) at 130 ft. elevation - at 78 ft. in the bioshield - at 78 ft. outside the bioshield. Additional mitigating actions taken following U2 Containment evacuation were as follows: - FME screen installed on open manways for 21/23 S/G with additional plastic covering and tape to prevent further gas release into containment. - Modified Containment Purge in service to maximize ventilation in Containment. - 21/24 S/G draining to support filling and draining evolutions to reduce Hydrazine concentrations in the S/G's. - Releasing tags on the AFWST to commence filling and further support filling and draining evolutions on the U2 S/G's. The licensee notified the NRC Resident Inspector. Notified R1DO (Arner), NRR EO (King), and IRD (Stapleton). Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

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.
ENS 5262118 March 2017 15:19:0010 CFR 50.72(a)(1)(i), Emergency Class DeclarationAlert Declared for Indications of Fire in Safety Related Switchgear

Alert declared at 1119 EDT 3/18/17 based on EAL H.A.2 - Fire or Explosion affecting plant safety systems. Fire alarms in the Unit 3 4kV switchgear rooms resulting in a loss of the 3A 4kV bus and trip of all three Reactor Coolant Pumps. The reactor tripped and was stabilized in Mode 3. No actual fire was observed. The 3A 4kV is deenergized. The 3B Reactor Coolant Pump was restarted for forced circulation. All other safety systems functioned as required. A refueling outage was scheduled to begin on 3/20/17. All control rods fully inserted on the reactor trip. Decay heat is being removed using feedwater and steam generator atmospheric steam dumps. One person was injured with a minor burn and possible sprained ankle and was taken to a local hospital. The licensee notified the NRC Resident Inspector. Notified DHS SWO, DOE, FEMA, HHS, NICC, USDA, EPA, FDA (e-mail), NWC (e-mail), NNSA (e-mail), and NRCC SASC (e-mail).

  • * * UPDATE AT 1426 EDT ON 3/18/2017 FROM DAN HAGARDY TO BETHANY CECERE * * *

Unit 3 was determined by the Emergency Coordinator to be in a safe and stable condition, the Emergency Plan personnel at the Technical Support Center and Emergency Operations Facility were no longer required for support, the Operations Support Center was staffed for recovery efforts, and plant personnel were sufficient and capable for continuing mitigation efforts. Investigation of the fault on the 3A 4kV bus is ongoing. Based on the above conditions, the Alert was exited at 1420 hours (on 3/18/2017). The injured electrician was taken to an offsite hospital to treat minor burns and possible sprained ankle. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ehrhardt), NRR EO (Miller and King), IRD MOC (Stapleton), DHS SWO, DOE, FEMA, HHS, NICC, USDA, EPA, FDA (e-mail), NWC (e-mail), NNSA (e-mail), and NRCC SASC (e-mail).

ENS 5246930 December 2016 18:05:0010 CFR 50.72(b)(3)(iv)(A), System ActuationManual Reactor Trip Due to Control Rod Not Withdrawing as ExpectedOn 12/30/16 at 1302 EST, Unit 1 began withdrawing control bank rods for an approach to criticality following a refueling outage. At 1305 EST, operators observed that control rod H-6 did not withdraw. Operators entered the applicable Abnormal Operating Procedure. Operators tripped the reactor as required by plant procedures and entered applicable Emergency Operating Procedures. The act of manually tripping the reactor generated a valid trip signal in the plant Reactor Protection System. Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater was already supplying the Steam Generators; a Feedwater Isolation occurred due to plant conditions. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators, and decay heat removal via the steam dumps. Method of decay heat removal is Steam Generators via the steam dumps. Current reactor coolant system conditions: Temperature at 548 degrees F and stable, pressure 2235 psig and stable. All control and shutdown banks are inserted. Electrical alignment is normal, supplied by offsite power. No impact to Unit 2, Unit 2 is in Mode 1 at 100 percent power. The licensee notified the NRC Resident Inspector.