Emergency Notification System
|Region 1||1||[Region 1 Events]||[RSS]|
|Region 2||2||[Region 2 Events]||[RSS]|
|Region 3||1||[Region 3 Events]||[RSS]|
|Region 4||3||[Region 4 Events]||[RSS]|
by Reactor type
|Entered date||Site||Region||Reactor type||System||Scram||Event description|
|ENS 54116||13 June 2019 03:59:00||Brunswick||NRC Region 2||High Pressure Coolant Injection|
Reactor Core Isolation Cooling
|At 2127 EDT on June 12, 2019, during routine testing, the HPCI turbine experienced an overspeed trip and then subsequently restarted and ramped to the required speed. As a result, the response time of the system exceeded the 60-second acceptance criteria, thereby rendering the system inoperable. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The Reactor Core Isolation Cooling (RCIC) System and Automatic Depressurization System (ADS) are operable.
The safety significance of this event is minimal. Troubleshooting activities are in progress.There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
|ENS 54113||12 June 2019 12:17:00||Columbia||NRC Region 4||A contract employee supervisor had a confirmed positive test for alcohol during a random fitness-for-duty test. The employee's access to the plant had been terminated. The NRC Resident Inspector has been notified.|
|ENS 54112||12 June 2019 11:55:00||Watts Bar||NRC Region 2||At 0849 (EDT), a significant air leak on an inline air filter was identified. At 0908, the leak on the filter was isolated. A subsequent review of this situation determined that this air leak impacted operation of the A Train of the Control Room Emergency Air Temperature Control System (CREATCS) which is required to be operable in accordance with Technical Specification 3.7.10. At the time of this event, the B Train of CREATCS was out of service for planned maintenance.
With both trains of CREATCS out of service, both Watts Bar Units entered a condition that could have prevented fulfillment of a safety function. This condition was terminated when the leaking air filter was isolated. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D).NRC Resident Inspector has been notified."
|ENS 54111||11 June 2019 17:55:00||Monticello||NRC Region 3||Primary containment||At 1132 CDT on 6/11/2019, both manual primary containment isolation valves in a one-inch service air line were found open. This resulted in an open primary containment penetration. Both valves are required to be closed for Primary Containment Isolation Valve Operability. Both valves were closed and independently verified closed at 1149 CDT on 6/11/2019.
This is being reported under 10 CFR 50.72(b)(3)(v)(C) and (D), and 10 CFR 50.72(b)(3)(ii)(B). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.The licensee also notified the State of Minnesota State Duty Officer.
|ENS 54109||11 June 2019 17:03:00||Wolf Creek||NRC Region 4||At 1324 CDT, on 6/11/19, Coffey County Emergency Management issued the following alert:
The Civil Authorities have issued a Nuclear Power Plant Warning for Coffey, KS beginning at 1323 CDT and ending at 1423 CDT (WIBW radio AM/FM). Coffey County Emergency Management Required Weekly Test. A press release is planned to notify residents that the warning was only a test. There was no impact to the health and safety of the public as a result of this event as the offsite response capabilities remain functional. The site is operating with no emergency conditions present. This event is being reported under 10 CFR 50.72(b)(2)(xi), as an inadvertent notification of the IPAWS (Integrated Public Alert Warning System) system. A press release is planned.The NRC Resident Inspector has been notified."
|ENS 54108||11 June 2019 16:57:00||Arkansas Nuclear||NRC Region 4||A non-licensed contract employee supervisor had a confirmed positive for a controlled substance during a pre-access fitness for duty test. The individual's unescorted access to the plant has been terminated and the badge removed.|
|ENS 54103||6 June 2019 12:02:00||Susquehanna||NRC Region 1||Main Steam Isolation Valve||This 60-day telephone notification is being made in accordance with the reporting requirements specified by 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of a general containment isolation signal affecting multiple Main Steam Isolation Valves (MSIVs).
On April 18, 2019 at approximately 0110 EDT, during performance of an ASME reactor vessel leak check, all four inboard MSIVs closed as a result of actuation of the Main Steam Line (MSL) high flow instrumentation. The high flow signal was spurious on the 'D' channel with no flow in the MSLs. Since an actual high flow condition did not exist at the time of the actuation, the actuation was considered invalid. The MSIVs functioned as designed on actuation of the MSL high flow instrumentation. All outboard MSIVs were closed at the time of the actuation in accordance with the vessel leak check procedure.The NRC Resident Inspector has been notified."
|ENS 54099||4 June 2019 11:39:00||Cooper||NRC Region 4||On 06/04/2019, Nebraska Public Power District will issue a press release concerning the spurious actuation of emergency sirens near Cooper Nuclear Station and Indian Cave State Park.
This is a four hour report per 10 CFR 50.72(b)(2)(xi) for any event or situation for which a news release is planned or notification to other government agencies has been or will be made which is related to heightened public or government concern. The cause of the siren actuation is still under investigation. The licensee notified the NRC Resident Inspector.
During this event, State & local government agencies (Nemaha County, Atchison County, Richardson County, and Indian Cave State Park) were contacted regarding the spurious actuation of emergency sirens. This is an update to the original Event Notification # 54099.Notified R4DO (Kellar).
|ENS 54098||4 June 2019 04:10:00||Limerick||NRC Region 1||Feedwater||Manual Scram||At 0145 EDT, on 6/4/19, Unit 2 was manually scrammed during a Rapid Plant Shutdown. At 64 percent reactor power, a Rapid Plant Shutdown was initiated due to lowering Main Condenser vacuum as a result of the loss of a plant electrical panel that powers Offgas System controls.
The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Turbine Bypass Valves to the Main Condenser and normal level control using Feedwater and Condensate. Main Condenser Vacuum has been restored. The licensee notified the NRC Resident Inspector. Additionally, State and local government agencies were notified.Prior to restarting Unit 2, an evaluation needs to be done due to the Unit 1 Diesel currently out of service for maintenance. The Unit 1 Diesel is a power supply for some of the common systems under the Unit 2 Technical Specifications and therefore required.
|ENS 54097||1 June 2019 17:58:00||South Texas||NRC Region 4||On June 1,2019, at 1618 (CDT), a notification under 10 CFR 50.72(b)(2) is being made due to notification to offsite agencies as a result of gasoline leakage to the site drainage system in the owner controlled area at South Texas Project.
During a routine tour, the facilities department notified the site environmental group about a gasoline leak on fuel tank sight glass at the fuel island on site. The site environmental (group) has determined the leak amount requires notification to the Texas Commission of Environmental Quality and the Environmental Protection Agency National Response Center. The Texas Commission of Environmental Quality was notified at 1618 on June 1, 2019, and the Environmental Protection Agency National Response Center at 1626 on June 1, 2019. The NRC Resident Inspector has been notified.The licensee stated that approximately 1,384 gallons of gasoline leaked over a period of time. The spill is located at an equipment warehouse area at a distance from the plant. The leak has been isolated and the cleanup is expected to be completed by tomorrow.
|ENS 54096||1 June 2019 03:15:00||River Bend||NRC Region 4||Reactor Core Isolation Cooling||Manual Scram||At 2345 CDT at River Bend Station (RBS) Unit 1, a manual Reactor scram was inserted in anticipation of receiving an automatic Reactor Water Level 3 (9.7") scram due to the isolation of the 'B' Heater String with the 'A' Heater String already isolated. The 'B' heater string isolation caused loss of suction and subsequent trip of the running Feed Water Pumps 'A' and 'C'. All control rods fully inserted with no issues. Subsequently Reactor level was controlled by the Reactor Core Isolation Cooling (RCIC) system. Feed Water Pump 'C' was restored 4 minutes after the initial trip and the RCIC system secured. Currently RBS-1 is stable and is being cooled down using Turbine Bypass Valves.
No radiological releases have occurred due to this event from the unit. The plant is currently under a normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector.
This amended event notification is being made to provide additional information that was not included in the original notification made on 6/1/19 at 0315 EDT. This event was reportable under 10 CFR 50.72(b)(3)(iv)(A) which was not annotated or described in the original report. Forty-two minutes after the Feed Water Pump 'C' was started, the pump tripped causing a Reactor Water Level 3 (9.7") RPS actuation. Feed Water was restored five minutes later using the Feed Water Pump 'A'. The NRC Resident Inspector has been notified.Notified the R4DO (Warnick).
|ENS 54094||30 May 2019 00:10:00||Fermi||NRC Region 3||Secondary containment||On May 29, 2019, at 2210 EDT, plant personnel notified the Main Control Room that both doors in the Secondary Containment Airlock on the Reactor Building First Floor were opened simultaneously for a period of approximately two seconds. This resulted in Technical Specification (TS) Surveillance Requirement (SR) 188.8.131.52.3 not being met. Secondary Containment pressure observed during that time remained unchanged and within TS limits. There were no radiological releases associated with this event.
Declaring Secondary Containment inoperable as a result of not meeting TS SR 184.108.40.206.3 is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material.The licensee has notified the NRC Resident Inspector."
|ENS 54092||27 May 2019 11:53:00||Fermi||NRC Region 3||On May 27, 2019 at 0940 EDT, a portable chemical toilet was found tipped over. Approximately one gallon of contents spilled to the gravel only and did not reach any waterways or storm drains. Cleanup efforts are in progress. A notification to the Michigan Department of Environmental Quality and local health department is required, as well as a press release.
This event is being reported pursuant to 10 CFR 50.72(b)(2)(xi).The licensee has notified the NRC Resident Inspector.
|ENS 54091||26 May 2019 09:25:00||Arkansas Nuclear||NRC Region 4||Feedwater|
Main Steam Safety Valve
|Automatic Scram||This is a 4-hour Non-Emergency 10 CFR 50.72(b)(2)(iv)(B) notification due to a Plant Protection System (PPS) actuation. Arkansas Nuclear One, Unit 2, automatically tripped from 100 percent power at 0512 CDT. The reactor automatically tripped due to 2P-32B Reactor Coolant Pump tripping as a result of grounding.
No additional equipment issues were noted. All control rods fully inserted. Emergency Feedwater (EFW) actuated and was utilized to maintain Steam Generator (SG) levels. The EFW actuation meets the 8-hour Non-Emergency Immediate Notification Criteria of 10 CFR 50.72(b)(3)(iv)(A). No Primary safety valves lifted. Main Steam Safety Valves (MSSVs) did lift initially after the trip. The NRC Resident Inspector has been notified.Decay heat is being removed via the steam dump valves to the main condenser. Unit 2 is in a normal shutdown electrical lineup. Unit 1 was not affected by the transient on Unit 2. The licensee notified the State of Arkansas."
|ENS 54090||26 May 2019 01:58:00||Braidwood||NRC Region 3||At 1930 (CDT) on 5/25/2019, communications were lost with the main control room area radiation monitors. These detectors are used to determine if an emergency action level (EAL) has been reached for initiating condition RA3 (Radiation levels that impede access to equipment necessary for normal plant operations, cooldown, or shutdown).
This unplanned loss of the ability to evaluate an EAL for initiating condition RA3 is considered a loss of emergency classification capability and is reportable as a Major Loss of Emergency Preparedness Capabilities per 10 CFR 50.72(b)(3)(xiii). This is an 8-hour reportable notification. Portable area radiation monitors have been established as a compensatory measure per station procedures.The NRC Resident Inspector has been notified."
|ENS 54089||25 May 2019 00:30:00||Nine Mile Point||NRC Region 1||A licensed employee was determined to be under the influence of alcohol during a random (fitness-for-duty) test. The employee's access to the plant has been canceled. The licensee notified the NRC Resident Inspector.|
|ENS 54087||24 May 2019 17:12:00||Wolf Creek||NRC Region 4||Emergency Diesel Generator|
|At 1310 CDT on 5/24/2019, Wolf Creek experienced a loss of offsite power to the safety-related NB02 bus, due to an external fire on a bushing on the startup transformer. The NB02 bus was reenergized when the 'B' Emergency Diesel Generator started and the output breaker automatically closed. The shutdown sequencer automatically started equipment as expected.
Due to the undervoltage condition on the NB02 bus, an AFAS-T (Auxiliary Feedwater Actuation Signal) signal was generated which started the turbine driven auxiliary feedwater pump. Turbine load was reduced to maintain reactor power less than 100% in response to the start of turbine driven and 'B' motor driven auxiliary feedwater pumps. The fire was extinguished using a fire extinguisher at 1320 CDT. The unit is stable at 97% power. The NB02 bus remains on the 'B' Emergency Diesel Generator (EDG). The other EDG is operable in standby.The NRC Resident Inspector was notified.
|ENS 54085||24 May 2019 13:09:00||Perry||NRC Region 3||Core Spray|
|At 0730 (EDT) on May 24, 2019, it was discovered that the Low-Pressure Core Spray System was inoperable. At Perry, the Low-Pressure Core Spray system is considered a single train system in Modes 1, 2, and 3; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). Inoperability of the Low-Pressure Core Spray system was caused by Emergency Service Water Pump A inoperability. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."|
|ENS 54083||24 May 2019 09:51:00||Limerick||NRC Region 1||Primary containment|
Residual Heat Removal
|This 60-Day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of Limerick Generating Station Unit 2 containment isolation logic.
On April 18, 2019, while performing a relay replacement on the Division 2/4 Main Steam Line logic, a partial containment isolation occurred due to a blown fuse. The following systems had components that actuated due to the partial isolation: Reactor Water Clean-Up System Primary Containment Instrument Gas System Drywell Chilled Water System Reactor Enclosure Cooling Water System Core Spray System The Residual Heat Removal System received an isolation signal; however, the system remained in service because the isolation was defeated in accordance with plant procedures. This event resulted in partial Group 2A, 3, 7A, 8A, and 8B isolations. The systems successfully functioned per the plant design and plant configuration.The licensee notified the NRC Resident Inspector.
|ENS 54078||22 May 2019 06:01:00||Susquehanna||NRC Region 1||HVAC||On 5/22/2019, the 'A' Control Structure Chiller (Div I) tripped due to a loss of (motor control center) MCC 0B136. The 'B' Control Structure Chiller was already inoperable due to Div II (Emergency Service Water) ESW being out of service for planned maintenance. With the loss of Control Structure HVAC System the ability to maintain temperatures in various spaces including relay rooms, Control Room Floor Cooling and Emergency Switchgear rooms was lost. The 'B' Control Structure Chiller was restarted at 0251 EDT and cooling was reestablished to the required areas, however the 'B' chiller is not considered operable at this time.
Units 1 and 2 entered (Technical Specification) TS 3.0.3 at 0256 EDT and a controlled shutdown of both units commenced, Unit 2 at 0340 EDT and Unit 1 0350 EDT. This constitutes a TS required shutdown and requires a 4 hour (Emergency Notification System) ENS notification in accordance with 10 CFR 50.72(b)(2)(i). The failure also requires an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(v) due to the loss of a safety function. The licensee needs to restore the 'B' loop of ESW to exit the Limiting Condition of Operation (LCO). The licensee is currently performing a flow surveillance, once complete and assuming the data is acceptable, the licensee will be able to exit the LCO. The units are in a normal electrical lineup. The licensee will be notifying the state of Pennsylvania FEMA Operations Center. The licensee has notified the NRC Resident Inspector.
On 5/22/2019 at 0601 EDT Susquehanna Steam Electric Station reported a shutdown had been commenced at 0340 EDT for Unit 2 and 0350 EDT for Unit 1 due to inoperability of both control structure chillers. Power has been restored to MCC 0B136, and at 0901 EDT the 'A' control structure chiller was declared operable and LCO 3.0.3 was exited. Power reduction for both units was halted at 0901 EDT and preparations for power restoration initiated. As of 1255 EDT on 5/22/2019, Unit 1 power is 94% and Unit 2 power is 92%.Notified the R1DO (Arner).
|ENS 54077||22 May 2019 05:45:00||Watts Bar||NRC Region 2||Feedwater|
|Manual Scram||On May 22, 2019, at 0233 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 reactor was manually tripped due to a failure of the #2 Main Feedwater Regulating Valve during power ascension following a refueling outage. 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 Dumps. Unit 2 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). There was no impact to WBN Unit 1.The NRC Senior Resident has been notified."
|ENS 54075||21 May 2019 17:14:00||Summer||NRC Region 2||At approximately 1833 EDT on 5/20/2019, VC Summer Nuclear Station identified a potential missed surveillance of the Spent Fuel Storage Canister Heat Removal System Surveillance Requirement (SR) 3.1.2. This is a surveillance requirement in the Holtec CoC No. 1032 Amendment 0, Revision 1 for the HI-STORM FW. It was identified that the station suspended verifying all overpack inlets and outlets were free of blockage from solid debris or floodwater every 24 hours and began utilizing the alternate surveillance method of installed temperature monitoring equipment. However, this monitoring equipment had not been declared operable from the completion of the Dry Cask Storage Campaign on April 11, 2019. This surveillance was last performed satisfactorily on May 15, 2019.
The station has documented this missed surveillance in the Corrective Action Program under CR-19-01866. The NRC Senior Resident has been notified. The licensee stated that the temperature monitoring equipment while not operable, was functional. The licensee successfully performed the surveillance requirement on 5/21/2019.
ENS Event Number 54075, made on May 21, 2019, is being retracted. NRC Notification 54075 was made to ensure the 24-Hour Non-Emergency reporting requirement of 10 CFR 72.75(d)(1)(i) was met for a potential missed surveillance of the Spent Fuel Storage Canister Heat Removal System Surveillance Requirement (SR) 3.1.2. On June 4, 2019, a Past Operability Evaluation determined the temperature monitoring equipment, utilized to complete the Technical Specification Surveillance 3.1.2, was operable during the time period of 5/16/2019-5/21/2019. The station performed satisfactory calibration checks on the temperature monitoring equipment and had not experienced any significant weather events capable of producing vent blockage (i.e. flood, tornado, snow storm, etc.). Therefore, utilizing the installed temperature monitoring equipment is not considered a missed surveillance of SR 3.1.2 and renders this event not reportable under 10 CFR 72.75(d)(1)(i). VCSNS's evaluation of this event is documented in the Corrective Action Program in Condition Report (CR-19-01866). The licensee has notified the NRC Resident Inspector.Notified R2DO (Rose) and NMSS Events Notification (email).
|ENS 54073||20 May 2019 00:02:00||Arkansas Nuclear||NRC Region 4||Safety Parameter Display System||On May 19, 2019, at 1809 CDT, the Safety Parameter Display System (SPDS) was lost to both Arkansas Nuclear One Units 1 and 2 due to the SPDS Inverter (2Y-26) failure. The SPDS Inverter is the power supply to both units' SPDS. The Unit 2 Control Room dispatched operators in response to a smoke alarm received from the 2Y-26 Inverter room. Upon arrival, smoke was reported emanating from 2Y-26. There was no report of fire at any time. Field operators de-energized 2Y-26 and the smoke ceased. The loss of SPDS also caused the Power Operating Limits (POL) function of the Unit 2 Core Operating Limits Supervisory System (COLSS) to be lost, so Unit 2 reduced power to 91 (percent) in accordance with Technical Specifications. Both units are at power and stable.
The NRC Resident has been notified.This is reportable per 10 CFR 50.72(b)(3)(xiii)."
|ENS 54072||18 May 2019 14:00:00||Turkey Point||NRC Region 2||Reactor Protection System|
Reactor Coolant System
|Manual Scram||This is a non-emergency notification to the NRC Operations Center in accordance with 10 CFR 50.72(b)(2)(iv)(B) for a valid actuation of the Reactor Protection System (RPS) (four hour notification) and 10 CFR 50.72(b)(3)(iv)(A) for a valid Engineered Safeguards (ESF) actuation (eight hour notification) due to Auxiliary Feedwater (AFW) initiation.
Unit 3 manual reactor trip following grid disturbance. Following the grid disturbance, a manual reactor trip was initiated due to lowering steam generator water levels. All control rods fully inserted. AFW started as expected. All other systems responded as expected. Current reactor temperature is 547 degrees F. Current reactor pressure is 2235 psig. Decay heat is being removed through the Atmospheric Steam Dumps (no known primary to secondary Reactor Coolant System leakage exists). The unit is in a normal post-trip electrical lineup. There was no affect on Unit 4. The cause of the grid disturbance is under investigation.The licensee notified the NRC Resident Inspector.
|ENS 54071||18 May 2019 02:09:00||Pilgrim||NRC Region 1||Feedwater|
Reactor Protection System
Main Steam Isolation Valve
|Manual Scram||On Friday, May 17, 2019 at 2303 (EDT), with the reactor at 70 (percent) core thermal power, Pilgrim Nuclear Power Station initiated a manual reactor scram due to degrading condenser vacuum as a result of the trip of Seawater Pump B. All control rods inserted as designed. The plant is in hot shutdown.
Plant safety 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 manual 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 Actuation
Due to the Reactor Protection System 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 or multiple main steam isolation valves (MSIVs).' 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 54069||17 May 2019 03:35:00||Callaway||NRC Region 4||Auxiliary Feedwater|
Reactor Coolant System
|EN Revision Text: REACTOR TRIP DUE TO SOURCE RANGE HI FLUX SIGNAL
This is an 8-hour, non-emergency notification for a valid reactor trip signal with the reactor not critical, and a valid auxiliary feedwater system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A) - Valid System Actuation.
At 2303 (CDT) on May 16, 2019, the plant was administratively in mode 2 due to withdrawing control rods for startup following refuel. The reactor had not been declared critical. The P-6 permissive at 10E-10 Amps was met for one of two Intermediate Range detectors allowing for block of the Source Range high flux trip (1E5CPS). Prior to performing the block, the Source Range high flux trip setpoint was exceeded and a reactor trip received. All systems responded as expected. A feedwater isolation signal was received due to the reactor trip with feedwater temperature less than 564 degrees Fahrenheit. Auxiliary feedwater was started to maintain steam generator levels. The plant is being maintained stable in mode 3 with no complications. The NRC Resident Inspector was present during the startup and was notified of the reactor trip.
A correction is being made for the sixth sentence in the second paragraph above, which states, 'A Feedwater Isolation signal was received due to the reactor trip with feedwater temperature less than 564 degrees Fahrenheit.' Within this sentence, 'feedwater temperature' is to be replaced with 'reactor coolant system temperature.'The licensee has notified the NRC Senior Resident Inspector.
|ENS 54068||16 May 2019 18:07:00||Waterford||NRC Region 4||Steam Generator|
|Automatic Scram||This is a non-emergency notification from Waterford 3.
On May 16, 2019, at 1348 CDT, Waterford 3 experienced an automatic reactor trip due to Steam Generator number 1 high level, which was the result of a Main Turbine trip and subsequent reactor power cutback which had occurred at 1345 CDT. The cause of the Main Turbine trip is currently under investigation. Subsequent to the Reactor trip, Main Feedwater Isolation Valves number 1 and number 2 closed on high Steam Generator levels. Emergency Feedwater automatically actuated for Steam Generator number 2 at 1419 CDT and Steam Generator number 1 at 1425 CDT. Main Feedwater was restored to both Steam Generators by 1629 CDT. The plant entered the Emergency Operating Procedure for an uncomplicated reactor trip and is in the process of transitioning to the normal operating shutdown procedure. The plant is currently in Mode 3 and stable with Main Feedwater feeding and maintaining both Steam Generators. The NRC Senior Resident Inspector has been notified.All control rods fully inserted. Decay heat is being removed through the main condenser. The plant is in a normal shutdown electrical lineup.
|ENS 54064||15 May 2019 02:21:00||Comanche Peak||NRC Region 4||Auxiliary Feedwater|
Emergency Diesel Generator
|At 2151 CDT, on 14 May 2019, Comanche Peak Nuclear Power Plant (CPNPP) experienced a voltage transient within the onsite 138kV switchyard due to the loss of one of the offsite switchyards supplying power to the CPNPP 138kV switchyard. The reduction in safeguards bus voltage due to the transient caused the Unit 2 safeguard busses to load shed and perform a slow transfer to power supplied from 345kV transformer XST2A. Unit 2 was subjected to actuation of both blackout sequencers causing an automatic start of both motor driven Auxiliary Feedwater (AFW) pumps as well as the turbine-driven AFW pump.
No emergency diesel generators started by design. All AFW pumps have been returned to standby status. All other safety systems functioned as designed. Unit 1 is currently defueled, and was unaffected by this event.The licensee has notified the NRC resident inspector."
|ENS 54062||12 May 2019 15:28:00||Grand Gulf||NRC Region 4||Service water|
Reactor Protection System
|Manual Scram||At 1039 CDT the reactor was manually (scrammed) due to a partial loss of plant service water. The loss of plant service water was caused by a loss of (balance of plant) BOP transformer 23. Reactor power was reduced in an attempt to restore pressure to plant service water. Reactor level is being maintained with condensate and feedwater. Reactor pressure is being maintained with bypass control valves. Standby Service Water A and B were manually initiated to supply cooling to Control Room A/C and (Engineered Safety Feature) ESF switchgear room coolers. The cause is under investigation.
The NRC Resident Inspector has been notified. 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 and also reported under 10 CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of RPS and Standby Service Water. The plant is currently in a normal electrical lineup.
This is an update to the original notification. The Drywell and Containment exceeded the technical specification (TS) temperature limits of 135 degrees F (TS Limiting Condition of Operation (LCO) 220.127.116.11) and 95 degrees F (TS LCO 18.104.22.168), respectively. An 8-hour notification is being added 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 control the release of radioactive material per 10 CFR 50.72(b)(3)(v)(C).Notified R4DO (Alexander).
|ENS 54061||12 May 2019 04:51:00||Callaway||NRC Region 4||On 5/11/19, Callaway Energy Center entered Mode 4 at 1217 (CDT). At 2305, the door from the Auxiliary Building to the RAM Storage building was found blocked open. This door is an Auxiliary Building pressure boundary for the Emergency Exhaust system. The Emergency Exhaust system is required in Modes 1,2,3,4, and during movement of irradiated fuel assemblies in the Fuel Building. The door was being blocked open with a large ramp. This rendered the Emergency Exhaust system not capable of performing its design safety function. LCO (Limiting Conditions for Operation) 3.7.13.B was entered, and preparations to move the ramp commenced. LCO 3.7.13.B is for two Emergency Exhaust trains being inoperable due to an inoperable auxiliary building boundary. The allowed outage time is 24 hrs. to restore the boundary to Operable. The door was closed and LCO 3.7.13.B was exited at 0111 on 5/12/19.
This event 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 (C) control the release of radioactive material, or (D) mitigate the consequences of an accident.The NRC Senior Resident has been notified."
|ENS 54060||10 May 2019 15:30:00||Dresden||NRC Region 3||At 0720 CDT (on 5/10/19), security was notified of a prohibited item (un-opened alcohol container) reported in the protected area. Security assumed escort of the non-supervisory (contract) individual and took custody of the prohibited item.
The employee's access to the plant has been suspended.The NRC Resident Inspector has been notified.
|ENS 54055||9 May 2019 13:51:00||Susquehanna||NRC Region 1||A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the site has been terminated. The NRC Resident Inspector has been informed.|
|ENS 54054||9 May 2019 07:01:00||Dresden||NRC Region 3||Reactor Protection System||Automatic Scram||On May 9, 2019 at 0348 CDT, an automatic scram was received on Unit 2 following a turbine trip.
All rods inserted to their full-in positions. All Group 2 and Group 3 automatic isolations actuated as expected. Systems operated as expected. Reactor vessel inventory and pressure are being maintained in normal control bands. This event is being reported pursuant to 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. The NRC Resident Inspector has been notified.Decay heat is being removed using the steam bypass valves to the condenser and the safety relief valves did not lift as a result of the trip.
|ENS 54052||6 May 2019 22:49:00||Brunswick||NRC Region 2||At 2204 EDT on 5/6/19, a Notification of Unusual Event (NOUE) was declared due to a fire lasting greater than 15 minutes. The fire occurred in the '2B' Heater Drain Pump motor located in the turbine building. The fire was extinguished following initial Emergency Declaration.
There were no releases to the environment. Unit 1 was unaffected by the event and remains in Mode 1 at 100 percent power. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
The NOUE was terminated as of 2359 EDT on 5/6/19. No off-site resources were required to extinguish the fire. The turbine building is now free of smoke. The licensee will notify the NRC Resident Inspector, State of North Carolina, Brunswick County, New Hanover County, and the Coast Guard. Notified R2DO (Heisserer), NRR EO (Miller), and IRD (Gott).Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
|ENS 54049||5 May 2019 20:41:00||Cooper||NRC Region 4||Secondary containment||At 1405 CDT, Secondary Containment differential pressure exceeded the Technical Specification limit due to a potential equipment failure. This required entry into (Limiting Condition of Operation) LCO 22.214.171.124 Condition A for Secondary Containment inoperability. An event or condition that could have prevented the fulfillment of a safety function requires an 8 hour report per 10 CFR 50.72(b)(3)(v)(C) for Control of Rad Release. Secondary Containment differential pressure was restored to greater than or equal to 0.25 inches vacuum, water gauge in accordance with plant procedures. Secondary Containment was declared operable at 1600 CDT. The issue has been entered in the Corrective Action Program and investigation of the cause is in progress. The NRC Senior Resident Inspector has been informed of this condition."|
|ENS 54047||3 May 2019 19:00:00||McGuire||NRC Region 2||Reactor Coolant System|
Reactor Protection System
|Automatic Scram||At 1554 EDT on 5/3/19, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped on Over Temperature Delta Temperature following a pressure transient in the Reactor Coolant System. The trip was uncomplicated with all systems responding normally post trip. Operations manually started the motor driven auxiliary feedwater pumps and has stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected.
Due to Reactor Protection System actuation while critical and actuation of the motor driven auxiliary feedwater pumps, 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. Unit 1 is in a normal electrical lineup. Prior to the automatic trip, the backup pressurizer heaters were in service as is normal during power ascension. The pressure transient started when the backup heaters were in the process of being removed from service.The licensee notified the NRC Resident Inspector.
|ENS 54046||3 May 2019 16:43:00||Turkey Point||NRC Region 2||A non-licensed contract supervisor had a confirmed positive during a for-cause fitness-for-duty test. The individual's authorization for site access has been terminated. The NRC Resident has been notified.|
|ENS 54040||1 May 2019 21:03:00||Farley||NRC Region 2||Reactor Protection System||Manual Scram||EN Revision Text: MANUAL REACTOR TRIP DUE TO MISALIGNED CONTROL ROD
At 1643 (CDT), with Unit 2 in Mode 2 during low power physics testing, the reactor was manually tripped per procedure due to a misaligned control rod. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the atmosphere using the atmospheric relief valves. Unit 1 is not affected. Due to the Reactor Protection System actuation, 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.
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). The reactor was tripped during low power physics testing. The misaligned rod was encountered during rod group insertion and the affected bank had been inserted to the extent that the reactor was subcritical when the operators tripped the reactor. The licensee notified the NRC Resident Inspector.Notified R2DO (Lopez)
|ENS 54037||30 April 2019 07:37:00||Indian Point||NRC Region 1||A non-licensed employee supervisor had a confirmed positive test for a prohibited substance during a follow-up fitness-for-duty test. The individual's unescorted access to the plant has been terminated. The NRC Senior Resident Inspector was notified by the licensee."|
|ENS 54031||26 April 2019 20:19:00||River Bend||NRC Region 4||Standby Liquid Control||At 1147 (CDT) on 4/26/19, a through wall leak (reported as 1 drop every 1 to 2 minutes) was identified and confirmed by operation and NDE (Non-Destructive Examination) personnel on the Standby Liquid Control injection line during pressure testing activities. The line is 1.5 inch in diameter and classified as an ASME Section Ill, Class 1 line. The leak is currently isolated from the reactor vessel by a danger tagged manual valve. The licensee notified the NRC Resident Inspector.|
|ENS 54027||25 April 2019 12:04:00||Saint Lucie||NRC Region 2||Automatic Scram||At 0918 (EDT) on 4/25/19, with (Saint Lucie) Unit 1 in Mode 1 at 100% power, the reactor automatically tripped due to a Turbine Trip. The reactor trip was uncomplicated with all systems responding normally.
Operations is maintaining the plant stable in Mode 3. Decay heat removal is being accomplished by main feed water and the main condenser using the turbine steam bypass valves. Unit 2 is not affected and remains at 100% power. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B).The NRC Resident Inspector has been notified."
|ENS 54024||24 April 2019 10:40:00||South Texas||NRC Region 4||On April 23, 2019, at 12:56 Central Time, South Texas Project Nuclear Operating Company (STPNOC) identified a programmatic failure, degradation, or discovered vulnerability of the fitness for duty (FFD) program that may permit undetected drug or alcohol use or abuse by individuals within a protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program. The NRC Resident Inspector has been notified.|
|ENS 54021||23 April 2019 09:44:00||Watts Bar||NRC Region 2||Control Room Emergency Ventilation||At 0232 EDT on April 23, 2019, a Main Control Room (MCR) alarm was received for low control room positive pressure.
At 0233 EDT, a Control Room Envelope (CRE) door was found ajar and immediately closed. Technical Specification 3.7.10 Control Room Emergency Ventilation System (CREVS) was declared not met for both trains. Watts Bar Unit 1 entered Condition B. Watts Bar Unit 2 was not performing movement of irradiated fuel assemblies and did not meet the APPLICABILITY for CREVS per LCO 3.7.10. At 0233 EDT on April 23, 2019, the alarm cleared, CREVS was declared operable and LCO 3.7.10 Condition B was exited. The safety function of the CRE boundary is to ensure the in-leakage of unfiltered air into the CRE will not exceed the in-leakage assumed in the licensing basis analysis of Design Basis Accident (DBA) consequences to CRE occupants. From 0232 EDT to 0233 EDT, (Watts Bar Nuclear) WBN was unable to validate that CREVS could fulfill its required Safety Function. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D).The NRC Resident Inspector has been notified."
|ENS 54020||22 April 2019 20:41:00||Byron||NRC Region 3||Auxiliary Feedwater|
|At 1324 CDT, on 4/22/2019, with unit 2 in Mode 3 at 0 percent power, an intentional manual initiation of the Auxiliary Feedwater System occurred in response to a loss of feedwater condition. The loss of feedwater condition occurred after the non-safety related Startup Feedwater Pump was secured due to high bearing temperatures. The A Train Auxiliary Feedwater Pump was started per procedure. The Auxiliary Feedwater System started and operated as designed following intentional manual initiation.
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 Feedwater System.There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
|ENS 54017||22 April 2019 16:16:00||Vogtle||NRC Region 2||A contract supervisor tested positive for drugs on a follow-up fitness-for-duty test. The contractor's access to the facility has been revoked and his badge was confiscated. Additionally, the supervisor failed a random test administered the next day (see EN #54018). The licensee notified the NRC Resident Inspector.|
|ENS 54018||22 April 2019 16:16:00||Vogtle||NRC Region 2||EN Revision Text: CONTRACT SUPERVISOR TESTED POSITIVE ON A RANDOM FITNESS-FOR-DUTY TEST
A contract supervisor tested positive for drugs on a random fitness-for-duty test. The contractor's access to the facility has been revoked and his badge was confiscated. Additionally, the supervisor failed a follow-up test administered the previous day (see EN #54017).
On April 16, 2019, an individual was selected for a follow-up drug test. The same individual was selected again on April 17, 2019 for a random drug test. The results for both tests were ruled by the Medical Review Officer (MRO) on the same day and ruled positive for the same drug on April 22, 2019. These FFD violations were reported to the NRC on April 22, 2019, as EN #54017 and EN #54018, respectively. As allowed by 10 CFR 26.185(o), the MRO further reviewed the quantitation of the drug in both tests and determined that no further drug use had occurred since the first positive test. Therefore, the MRO concluded that this should be considered one FFD violation, and EN #54018 is being retracted. No changes are needed to EN #54017. The NRC Resident Inspector has been notified of this retraction.Notified R2DO (Heisserer) and FFD Group (email).
|ENS 54016||22 April 2019 01:51:00||Brunswick||NRC Region 2||Shutdown Cooling|
Reactor Water Cleanup
Control Room Emergency Ventilation
Primary Containment Isolation System
Reactor Coolant System
|Automatic Scram||At 2307 EDT on April 21, 2019, in Mode 1 at approximately 100 percent reactor power, Unit 1 automatically tripped due to a Main Turbine Trip. The Main Turbine Trip was a result of two out of three level instruments sensing a false high reactor water level. All control rods inserted as expected during the scram. Safety Relief Valves G and K lifted per design. The same level instruments that failed also tripped both Reactor Feed Pumps. As a result, reactor water level dropped below the Low Level 1 and 2 actuation setpoints. Per design, the Low Level 1 signal resulted in 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 Low Level 2 signals resulted in Group 3 (i.e. Reactor Water Cleanup) isolation, a secondary containment isolation signal, and an auto start of Standby Gas Treatment and Control Room Emergency Ventilation. Also, the Low Level 2 resulted in (high pressure coolant injection) HPCI and (reactor core isolation cooling system) RCIC automatically starting and injecting into the vessel.
All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Decay heat is currently being removed via the turbine bypass valves. Condensate and feed water are maintaining water level. The reactor is still at saturation temperature and 475 psi, lowering slowly. The reactor is still in a normal electrical lineup. There was no impact to Unit 2 as a result of this event.
The licensee updated the event report to include a 4-Hr Non-Emergency Notification in accordance with 10 CFR 50.72(b)(2)(iv)(A) for Emergency Core Cooling System, HPCI, Discharge to the Reactor Coolant System.Notified R2DO (Dickson), NRR EO (Miller) and IR MOC (Gott).
|ENS 54015||21 April 2019 12:43:00||Limerick||NRC Region 1||Event of Public Interest performed to notify State and Local agencies for emergency vehicle response required due to an on-site non-work related illness. The individual was unresponsive and was unable to be resuscitated due to the medical issue. The individual was outside the Radiological Controlled Area (RCA) and no radioactive material or contamination was involved. The NRC Resident Inspector was notified.|
|ENS 54014||21 April 2019 08:46:00||Limerick||NRC Region 1||Primary containment||This 60-Day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of Limerick Generating Station Unit 1 containment isolation logic.
On February 22, 2019, while performing work on the 1C Main Seam Line Rad Monitor a partial containment isolation occurred due to a blown fuse. The blown fuse caused a single channel 'C' isolation signal for the Refueling Area Ventilation Exhaust High Radiation and the Reactor Enclosure Ventilation Exhaust-High Radiation logic. The following systems had components that actuated due to the partial isolation: - Plant Process Radiation Monitoring System - Nuclear Boiler System - Control Rod Drive Hydraulic System - Containment Atmospheric Control System - Primary Containment Instrument Gas System This event resulted in partial Group VIC and partial Group VIIIB isolations. All the components that would actuate on a single 'C' isolation signal responded as designed.The licensee notified the NRC Resident Inspector.
|ENS 54012||20 April 2019 09:54:00||Duane Arnold||NRC Region 3||Manual Scram||At 0507 (CDT on April 20, 2019), the DAEC (Duane Arnold Energy Center) experienced a trip of both reactor feed pumps. Operators inserted a manual scram. All control rods inserted, as required. As a result of the feed pump trips and scram, HPCI and RCIC automatically injected. Also, containment isolations occurred, as expected for this event. All systems responded as designed.
Operators are currently taking the unit to cold shutdown conditions. Vessel level is being controlled by RCIC with Condensate System available. Pressure is being controlled using Main Steam Line drains and the Main Condenser is available. Normal electrical lineup remains. The cause of the reactor feed pumps tripping is believed to be an instrument air leak to flow control valves, causing loss of suction to both feed pumps.The licensee notified the NRC Resident Inspector.
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