|Entered date||Site||Region||Reactor type||System||Scram||Event description|
|ENS 54163||15 July 2019 16:30:00||Fermi||NRC Region 3||Service water||At 1335 EDT on 7/15/2019, during dredging activities in Fermi 2's General Service Water (GSW) intake canal, a hydraulic line on the dredging machine became disconnected and approximately one quart of hydraulic oil spilled into Lake Erie. The oil leak to navigable waters has been stopped. The oil was contained within a boom, cleanup activities commenced immediately, and cleanup was completed at 1500 EDT. The cause of the oil leak is under investigation. Environmental spill reports were made to local, state, and federal government agencies. This is considered a news release or notification to other government agencies, therefore this event is reportable under 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. The State agencies notified were Michigan Department of Environmental Protection and the Michigan Pollution Emergency Alerting System. The licensee also notified the National Response Center.|
|ENS 54162||12 July 2019 22:50:00||Browns Ferry||NRC Region 2||High Pressure Coolant Injection|
|At 1640 CDT on 7/12/19, Unit 1 High Pressure Coolant Injection (HPCI) received an invalid auto isolation signal which closed the HPCI steam supply valves rendering HPCI inoperable. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D), as an 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 mitigate the consequences of an accident. The isolation occurred while performing a calibration and functional check of a level switch for the Unit 1 Core Spray system. Continuity was checked across the incorrect set of contacts which completed the circuit in logic bus 'A' for the auto isolation signal in the HPCI system. There was no impact to the safety of the public or plant personnel during the time HPCI system was isolated. HPCI was returned to operable at 2110 CDT on 7/12/19. CR 1532094 documents this condition in the Corrective Action Program. The licensee has notified the NRC Resident Inspector|
|ENS 54152||8 July 2019 18:40:00||Arkansas Nuclear||NRC Region 4||A non-licensed contract supervisor had a confirmed positive for a controlled substance during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.|
|ENS 54147||3 July 2019 18:32:00||Arkansas Nuclear||NRC Region 4||Feedwater||This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to provide information pertaining to an invalid Engineered Safety Feature actuation signal. On May 9, 2019, at Arkansas Nuclear One (ANO) Unit 1, while performing an Emergency Feedwater Initiation and Control (EFIC) Channel B monthly test, a test pushbutton was mispositioned, resulting in an inadvertent initiation of the Emergency Feedwater (EFW) System. In accordance with the Engineered Safeguards Actuation System (ESAS) Trip Test portion of the surveillance, the first technician placed EFIC Train B in the tripped condition. The second technician then went to the front of the control room to verify Remote Switch Matrix (RSM) indications. The first technician recalls thinking he was given the order to reset Train B EFW Bus 1 Trip. Therefore, the first technician performed the step using three-part communication, but there is uncertainty about what was said. Due to the amount of time the second technician spent in front of the control room, the first technician assumed Operations reset the RSM to complete the Train B reset. The second technician returned to the ESAS cabinet and directed the first technician to perform the reset of Train B EFW Bus 1 Trip. The first technician, expecting his next action to be the trip of Train B EFW Bus 2, placed Bus 2 in the tripped condition. This put both buses of Train B EFW in trip and caused the actuation of P-7A EFW Pump. This inadvertent actuation was caused by human error and was not a valid signal resulting from parameter inputs. The 1992 Statements of Consideration define an invalid signal to include human error. Therefore, this actuation is considered invalid. This event was entered into ANO's corrective action program for resolution. This event did not result in any adverse impact to the health and safety of the public. The plant responded as expected. In accordance with 10 CFR 50.73(a)(i) a telephone notification is being made in lieu of submitting a written Licensee Event Report. The licensee has notified the NRC Resident Inspector."|
|ENS 54145||3 July 2019 15:24:00||South Texas||NRC Region 4|
At 1026 CDT on July 3, 2019, the National Weather Service notified the South Texas Project (STP) that the Bay City NOAA radio transmitter was out of service. This affects the tone alert radios used to notify the public in the event of an emergency condition. This condition impacts the ability to readily notify a portion of the Emergency Planning Zone (EPZ) population for the STP Electric Generating Station. The loss of the alert radio system meets the reporting criterion for 10 CFR 50.72(b)(3)(xiii). The Matagorda County alert sirens, which are also part of the Public Prompt Notification System, remain operable. Compensatory measures have been verified to be available should the Prompt Notification System be needed. These measures consist of a reverse 911 system available from Matagorda County as well as local law enforcement personnel who would perform route alerting for the affected areas of the EPZ. The event has been entered into the Corrective Action Program and the NRC Resident Inspector has been notified. A return to service time for the radio transmitter is not currently available. Matagorda County was also notified by the National Weather Service and STP. This meets the reporting requirement for notification of an offsite agency, 10 CFR 50.72(b)(2)(xi).
At 1321 CDT, the National Weather Service communication tower was returned to service. South Texas Project verified that the automatic notification radios have been returned to service. The licensee has notified the NRC Resident Inspector. Notified R4DO (Haire).
|ENS 54144||3 July 2019 12:42:00||Brunswick||NRC Region 2||Emergency Diesel Generator||This 60-day optional telephone notification is being made in lieu of an LER submittal, as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 2000 EDT on May 9, 2019, an invalid actuation of emergency diesel generator (EDG) 1 occurred. At the time, EDG 1 was removed from service for planned maintenance. The invalid actuation occurred when the starting air clearance was being lifted while simultaneously performing a Post Maintenance Test (PMT) where an external DC power source was applied to a relay that provided continuity directly to the starting air solenoids. As a result, the air start solenoids were energized causing EDG 1 to start. EDG 1 started and functioned successfully. The actuation was complete; EDG 1 successfully started and ran unloaded. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. The licensee has notified the NRC Resident Inspector.|
|ENS 54137||27 June 2019 16:06:00||Susquehanna||NRC Region 1||Reactor Protection System|
Primary Containment Isolation System
Emergency Diesel Generator
|At 0937 EDT on 6/27/19, Susquehanna Unit 2 had a control power fuse fail that caused a loss of power to one of two power distribution systems of the Reactor Protection System. This loss of power resulted in a valid isolation signal and actuation of the Primary Containment Isolation System that affected multiple systems, and is reportable per 10 CFR 50.72(b)(3)(iv)(A). The situation has been resolved and the plant is currently using an alternate power source. Offsite power and the Emergency Diesel Generators are still available. The licensee notified the NRC Resident Inspector.|
|ENS 54131||25 June 2019 11:09:00||Waterford||NRC Region 4||On June 25, 2019, at 0428 CDT, the Waterford 3 shift operating crew declared the control room envelope inoperable in accordance with Technical Specification (TS) 184.108.40.206 due to both Broad Range Gas Monitors being inoperable. Operations entered TS 220.127.116.11 action b, which requires that with one or more control room emergency air filtration trains inoperable due to inoperable control room envelope boundary in Modes 1, 2, 3, or 4, then: 1. Immediately initiate action to implement mitigating actions; 2. Within 24 hours, verify mitigating actions ensure control room envelope occupant exposures to radiological, chemical, and smoke hazards will not exceed limits; and 3. Within 90 days, restore the control room envelope boundary to operable status. Action b.1 was completed by placing the control room in isolate mode at time 0441 CDT. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(A) and 10 CFR 50.72(b)(3)(v)(D), event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to (A) shutdown the reactor and maintain it in shutdown condition and (D) mitigate the consequences of an accident, due to the control room envelope being inoperable. The NRC Senior Resident Inspector has been notified."|
|ENS 54130||24 June 2019 21:18:00||FitzPatrick||NRC Region 1||During a review of industry Operating Experience it was identified that there were unprotected DC control circuits for non safety-related DC motors which are routed from the Battery Charger Rooms to other separate fire areas. Circuit Breakers used to protect the motor power conductors appear to be inadequate to protect the control conductors. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one area can cause short circuits potentially resulting in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable as an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Requirements of the Technical Requirements Manual (TRM) for the affected fire areas will be implemented." The licensee has notified the NRC resident inspector.|
|ENS 54127||20 June 2019 17:59:00||Watts Bar||NRC Region 2||Shield Building||At 1340 EDT on June 20, 2019, a breach in excess of allowable margin in the Unit 2 Shield Building annulus was identified. T.S. LCO 3.6.15, Condition A was entered. The breach is expected to be repaired within the 24 hours allowed LCO time. No other equipment issues were identified. The Shield Building ensures that the release of radioactive material from the containment atmosphere is restricted to those leakage paths and associated leakage rates assumed in the accident analysis during a Loss of Coolant Accident (LOCA). This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(C). NRC Resident Inspector has been notified. The breach consists of a tear in a flexible boot seal for a penetration associated with the suction path for gas treatment fans. There is no release of radioactive material associated with this event.|
|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) 18.104.22.168.3 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 22.214.171.124.3 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
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
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) 126.96.36.199) and 95 degrees F (TS LCO 188.8.131.52), 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|
EN Revision Text: SECONDARY CONTAINMENT DECLARED INOPERABLE DUE TO POTENTIAL EQUIPMENT FAILURE 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 184.108.40.206 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.
CNS (Cooper Nuclear Station) is retracting the 8-hour notification made for event 54049 which occurred on May 5, 2019 at 1405 CDT. Subsequent evaluation determined that no equipment failure occurred. In addition, there were no procedure inadequacies or human performance issues identified. The indications observed were expected and part of a pre-planned evolution which included entry into a planned LCO for the Secondary Containment. The NRC Resident Inspector has been notified. Notified the R4DO (Kozal).
|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.|