|Entered date||Site||Region||Reactor type||Event description|
|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 54098||4 June 2019 04:10:00||Limerick||NRC Region 1||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 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 54083||24 May 2019 09:51:00||Limerick||NRC Region 1||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 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 54054||9 May 2019 07:01:00||Dresden||NRC Region 3||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 54020||22 April 2019 20:41:00||Byron||NRC Region 3||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 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||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 53998||14 April 2019 03:21:00||Nine Mile Point||NRC Region 1||On April 14, 2019 at 0003 (EDT), Nine Mile Point Unit 1 experienced an automatic reactor scram during reactor startup. The cause of the automatic scram was due to high (Reactor Pressure Vessel) pressure following closure of the turbine stop valves. All control rods fully inserted and all plant systems responded per design following the scram. Following the automatic scram, the High Pressure Coolant Injection (HPCI) System automatically initiated as expected. At Nine Mile Point Unit 1, a HPCI System actuation signal on low Reactor Pressure Vessel (RPV) level is normally received following a reactor scram, due to level shrink. HPCI is a flow control mode of the normal feedwater systems, and is not an Emergency Core Cooling System. At 0004, RPV level was restored above the HPCI System low level actuation set point and the HPCI System initiation signal was reset. Pressure control was established on the Turbine Bypass Valves, the preferred system. No Electromatic Relief Valves actuated due to this scram. Nine Mile Point Unit 1 is currently in Hot Shutdown, with reactor water level and pressure maintained within normal bands. The offsite grid is stable with no grid restrictions or warnings in effect. The unit is currently implementing post scram recovery procedures. The NRC Resident Inspector has been notified. The Licensee will notify the State of New York.|
|ENS 53977||4 April 2019 15:16:00||Oyster Creek||NRC Region 1||Oyster Creek NGS (Nuclear Generating Station) Tech Support Center (TSC) ventilation is not functional due to a broken belt on exhaust fan FN-843-14. The TSC ventilation system will remain non-functional until fan belt replacement can be completed. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff to an alternate location in accordance with applicable site procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to partial loss of the TSC. An update will be provided once the TSC ventilation has been restored to normal operation. The NRC Resident Inspector has been notified."|
|ENS 53933||14 March 2019 17:08:00||Peach Bottom||NRC Region 1||A licensed employee was determined to be under the influence of alcohol during a random test. The employee's access to the plant has been suspended pending an investigation. The licensee notified the NRC Resident Inspector.|
|ENS 53921||9 March 2019 13:59:00||Calvert Cliffs||NRC Region 1||A sewage line on the south end of the plant backed up causing sanitary wastewater to flow into storm drains and out to the Chesapeake Bay. This is a required notification of the Maryland Department of the Environment under COMAR (Code of Maryland Regulations) 26.08 for discharge of a pollutant into navigable waters or the adjoining shoreline. The amount has been estimated at less than 1000 gallons and the source has been isolated and storm drains have been covered to stop any flow into them and subsequently to the Chesapeake Bay. This notification is made in accordance with 10CFR50.72(b)(2)(xi) due to notification of a state agency. The licensee notified the NRC Resident Inspector.|
|ENS 53903||1 March 2019 04:03:00||LaSalle||NRC Region 3||On February 28, 2019, at 2217 CST, LaSalle Unit 2 experienced a trip of the 241Y Safety Related Bus during surveillance testing resulting in a valid undervoltage actuation signal to the Common Emergency Diesel Generator ('O' EDG), causing it to start and load to Bus 241Y. The purpose of the surveillance testing was to demonstrate the operability of the breakers necessary to provide the second off site source to Unit 2. This event is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A), as an event that results in a valid actuation of the emergency AC electrical power system. In addition to the 241Y bus trip and 'O' EDG actuation signal, the following plant responses occurred as designed due to the momentary loss of this AC Bus: "A" RPS de-energized due to the loss of the 2A Reactor Protection System Motor-Generator Set, and the running Unit 2 Fuel Pool Cooling pump tripped. The Non-Safety Related Bus 241X de-energized resulting in a trip of the Unit 2 Station Air Compressor. All systems have been restored and troubleshooting is currently in progress. Unit 1 remained in MODE 1 during this event. The NRC Senior Resident Inspector has been notified."|
|ENS 53861||6 February 2019 02:22:00||Quad Cities||NRC Region 3||On February 5, 2019, at 1804 (CST), during a Unit 1 High Pressure Coolant Injection (HPCI) operability surveillance, a fuse blew in the logic for the motor speed changer for the turbine. The Unit 1 HPCI system was taken out of service for planned maintenance earlier in the day. The fuse issue was not related to any maintenance activities. Had HPCI been demanded, this fuse failure would not have allowed HPCI to reach its required speed. HPCI remains inoperable pending resolution of the issue. The Reactor Core Isolation Cooling (RCIC) system was confirmed operable. There were no other systems inoperable at the time of the event. HPCI had been last successfully tested on November 6, 2018. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10 CFR 50.72(b)(3)(v)(D). The HPCI system is a single train system and the loss of HPCI could impact the plant's ability to mitigate the consequences of an accident. The NRC Senior Resident Inspector has been notified. Inoperable HPCI places the unit in a 14 day Technical Specification Limiting Condition of Operability.|
|ENS 53851||30 January 2019 17:41:00||Dresden||NRC Region 3||At 0910 (CST) on January 30, 2019, the Dresden Station Heater Boiler 'B' tripped while placing the station Heater Boiler 'A' in service. With colder temperatures, the density of the supply air increased and contributed to a greater quantity of air entering the Reactor Building than what was previously being supplied with heating steam in service. The Reactor Building differential pressure (DP) degraded and dropped below 0.25 inches water column vacuum. This condition represents a failure to meet Technical Specification (TS) Surveillance Requirement 126.96.36.199.1. Entry into TS 188.8.131.52 Condition A was made due to Secondary Containment becoming inoperable. Standby Gas Treatment System was initiated to assist with Reactor Building DP control. Reactor Building DP was restored to greater than 0.25 inches water column vacuum. TS 184.108.40.206 Condition A was exited. This event is being reported under 10 CFR 50.72(b)(3)(v)(C), '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.' The NRC Resident Inspector has been notified."|
|ENS 53831||16 January 2019 23:51:00||Fort Calhoun||NRC Region 4||At 1908 CST, a fire was reported in an unoccupied exclusion area opening (EAO) enclosure outside of the Fort Calhoun Station protected area. Offsite fire departments responded at 1923 CST and the fire was extinguished by 1930 CST. There were no injuries reported. The State Fire Marshal of Nebraska was notified by the Blair (Nebraska) Fire Department at approximately 1913 CST. His investigation determined the cause to be a malfunctioning heating element in a climate control unit. There was no release of radioactivity or hazardous materials. The climate control unit was clarified to be a ceiling mounted heater in the enclosure. The licensee notified the NRC Decommissioning Inspector.|
|ENS 53828||16 January 2019 08:12:00||FitzPatrick||NRC Region 1||On January 16, 2019, with James A. Fitzpatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated that Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge while isolating Reactor Building Ventilation. The Secondary Containment differential pressure was less than 0.25 inches of vacuum water gauge for approximately ten (10) seconds, and then immediately returned to greater than or equal to 0.25 inches of vacuum water gauge. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 220.127.116.11.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector."|
|ENS 53824||13 January 2019 17:49:00||Clinton||NRC Region 3|
EN Revision Text: HIGH PRESSURE CORE SPRAY SELF TEST FAILURE On January 13, 2019, the Self Test System reported a fault associated with the logic system for the High Pressure Core Spray (HPCS) high reactor water level closure function that could prevent the system from performing its safety function. The HPCS system was subsequently declared inoperable with actions taken per LCO (Limiting Condition for Operation) 18.104.22.168 to close and deactivate the 1E12-F004 valve, a primary containment isolation valve. Since HPCS is an emergency core cooling system and is a single train safety system, this condition is reportable under 10 CFR 50.72(b)(3)(v)(D) 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.' The NRC Resident Inspector has been notified. HPCS is in a 14-day technical specification LCO action statement.
Testing of the logic system load driver card for the High Pressure Core Spray (HPCS) high reactor water level closure function was completed both on site and at General Electric Hitachi (GEH). This testing determined the cause of the self-test system fault report was limited to the self-test portion of the load driver card and did not impact the ability of HPCS system to perform its specified safety function. Based on the testing results, this event is not reportable under 10 CFR 50.72(b)(3)(v)(D), 'Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.' Therefore, EN 53824 is being retracted. The NRC Resident Inspector has been notified. Notified the R3DO (Hills).
|ENS 53784||8 December 2018 06:12:00||Braidwood||NRC Region 3|
EN Revision Text: INOPERABLE CONTROL ROOM ENVELOPE Braidwood Station was performing Control Room Envelope Testing. During testing the Station identified a failed acceptance criteria. The Control Room Envelope is a single train system and could constitute a Loss of Safety Function. If a single train system is inoperable per Technical Specifications (TS), it is Reportable as a Loss of Safety Function per 10 CFR 50.72(b)(3)(v) regardless of the system's continued ability to meet the accident analysis requirements.
Both Units remain Mode 1, 100% power. The licensee will be notifying the NRC Resident Inspector. The acceptance criteria that failed was to maintain the control room pressure above the miscellaneous electrical equipment room pressure. The station has realigned ventilation to normal, and has entered TS Limited Condition for Operation (LCO) 3.7.10 condition B, which requires the station to restore to operable the control room envelope within 90 days or shutdown the plant. The station has also initiated contingency actions to verify SCBA (self contained breathing apparatus) are available and control room personnel are qualified to use SCBA.
On Wednesday, December 19, 2018, Braidwood Station concluded that the ENS notification 53784 could be retracted. It has been determined that the issue was not with the Control Room Envelope structure. Troubleshooting identified that the Unit 1 Upper Cable Spreading Room Area Supply Flow Control (OVC035Y) damper which supplies the Train A control room ventilation equipment room with air flow was not opening enough to supply the required flow. The subject duct work is shared by both A-train and B-train, and the flow through OVC035Y is controlled by a two-position actuator. The damper is less open when A-train is in operation (actuator energized) and more open when B-train is in operation (actuator de-energized). The only adjustments performed were to the actuator energized stroke limits which only affect the A train and thus a single train failure which could affect the safety function of both trains did not exist. Further calculations of unfiltered air inleakage into the Control Room Envelope (CRE) under a slightly negative differential pressure condition resulted in a calculated in leakage to the CRE of less than the maximum allowable unfiltered air inleakage for a radiological event of 436 scfm. The unfiltered air inleakage into the CRE assumed in the licensing basis analyses of Design Basis Accident consequences was never exceeded. Thus, TS Surveillance Requirement 22.214.171.124 continued to be met and entry into TS 3.7.10 Condition B was not required. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Stone).
|ENS 53778||5 December 2018 17:06:00||FitzPatrick||NRC Region 1||At 1010 (EST) on December 5, 2018, Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge. This condition existed for approximately 3 minutes before the differential pressure was restored to normal when the Standby Gas Treatment system was manually initiated. This event was caused by a trip of the service air compressor 39AC-2A. The loss of instrument air pressure caused Reactor Building ventilation to isolate and raise Secondary Containment differential pressure. The instrument air pressure was restored when 39AC-2A was isolated and the two backup air compressors started. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 126.96.36.199.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee notified the NRC Resident Inspector.|
|ENS 53765||2 December 2018 06:17:00||Nine Mile Point||NRC Region 1|
During the post-maintenance testing run of the Division III Emergency Diesel Generator (EDG), (a field operator) reported smoke coming from the diesel and an emergency shutdown was required. After the EDG was shutdown, significant damage (thrown rod) to the EDG was observed. Emergency Action Level HA 2.1 (an Alert) was declared at 0530 (EST). Currently, the plant is stable and operating at 100 percent power. All safety systems are available. The damage occurred approximately 20 minutes into the required 1 hour run. The licensee's emergency response organization has been activated. No offsite assistance was required or requested. There is a 14-day shutdown limiting condition for operation (LCO) in effect under technical specification 3.5.1 for the high pressure core spray system. Notified DHS Senior Watch Officer, FEMA Operations Center, DHS NICC Watch Officer, HHS Operations Center, DOE Operations Center, EPA Emergency Operations Center, FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email). The licensee has notified state and local authorities and the NRC Resident Inspector.
The licensee terminated the Alert at 0731 EST on 12/2/18. The basis for termination was that the licensee has met all procedural requirements to terminate the emergency and on-shift personnel can operate the unit without further assistance. Notified R1DO (Burritt), NRR EO (Miller), IRD MOC (Gott), HQPAO (Couret), ERDS Activation Group, DHS Senior Watch Officer, FEMA Operations Center, DHS NICC Watch Officer, HHS Operations Center, DOE Operations Center, EPA Emergency Operations Center, FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email).
|ENS 53698||28 October 2018 08:57:00||Clinton||NRC Region 3||At 0445 (CDT), with reactor power less than 1% rated thermal power on Instrument Range Monitor (IRM) ranges 6 and 7, Clinton Power Station received an automatic Reactor Protection System (RPS) actuation. The Reactor Scram Off Normal procedure was entered and all control rods were verified to be fully inserted. The apparent cause of the scram is cold water injection causing an upscale trip of the IRMs due to Motor Driven Reactor Feedwater Pump (MDRFP) Feedwater Regulating valve 1FW004 valve coming off the full shut seat momentarily. All systems responded appropriately following the scram and the plant is currently stable. Clinton Power Station will be proceeding to Mode 4 to support the planned Maintenance Outage. The NRC Senior Resident Inspector has been notified."|
|ENS 53696||27 October 2018 01:50:00||Ginna||NRC Region 1||RCS (Reactor Coolant System) Pressure: vented to containment, refueling cavity greater than 23ft. (above reactor vessel). RCS temperature: 96 degrees Fahrenheit. The 12A bus de-energized, 'A' EDG (Emergency Diesel Generator) automatically started and loaded on (emergency) buses 14 and 18. The RCS configuration is refueling cavity level greater than 23ft. above the reactor flange with no impact to shutdown cooling. Radiation monitor R-1, Control Room radiation monitor, lost power for 2 hrs 10 min. This placed Ginna in a major loss of emergency preparedness capabilities. A temporary radiation monitor has been installed in the Control Room. Prior to the notification, the licensee had restored the 12A bus from offsite power and the R-1 monitor was re-energized. The licensee notified the NRC Resident Inspector.|
|ENS 53693||24 October 2018 17:45:00||Quad Cities||NRC Region 3||On October 24, 2018 at 0901 CDT, during performance of the 'Functional Test of Unit 1 Second Level Undervoltage,' a loss of Bus 13-1 and Bus 18 occurred. The 1/2 Emergency Diesel Generator (EDG) automatically started due to a valid actuation on loss of power to Bus 13-1, but did not load due to required testing alignment. The loss of Bus 13-1 caused the loss of the 1A loop of Core Spray, both loops of Low Pressure Coolant Injection (LPCI), and Bus 18. All equipment responded as expected. Bus 13-1 and Bus 18 were restored at 0911(CDT) on 10/24/18. Other affected systems are in the process of being restored. An investigation as to the cause of the event has been initiated. This notification is being made in accordance with 10 CFR 50.72(b)(3)(iv), 'Event or Condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B),' because the 1/2 EDG auto started due to the loss of power condition. This notification is also being made in accordance with 10 CFR 50.72(b)(3)(v)(B), 'Event or Condition that Could Have Prevented Fulfillment of a Safety Function,' because both loops of LPCI were inoperable for a short time period. During the ten minutes where LPCI was unavailable, Unit 1 was in Technical Specification LCO 3.0.3. Unit 1 is currently in LCO 3.8.1(b) until the EDG is restored. Unit 2 was not affected by this event. The licensee will notify the NRC Resident Inspector.|
|ENS 53666||14 October 2018 07:35:00||Ginna||NRC Region 1||Notified New York State Department of Environmental Conservation for draining of sodium hypochlorite (12-15% by weight) from the storage tank into it's engineered secondary containment of approximately 1300 gallons. Reportable per regulation 6 NYCRR Part 597. The NRC Resident Inspector will be notified by the licensee. Licensee investigation into the cause of the leak is ongoing.|
|ENS 53663||12 October 2018 03:58:00||Calvert Cliffs||NRC Region 1||During a post maintenance start of the 1B diesel generator, the air start solenoid valves did not close as expected. This resulted in lowering air pressure in the common air start headers causing inoperability of the 2A and 2B diesel generators at time 23:03. The 1B diesel generator was isolated from the common air start header, which restored the air start header pressure to the 2A and 2B diesel generators. The 2A and 2B diesel generators were declared operable at 23:34. The NRC Resident Inspector was notified.|
|ENS 53657||10 October 2018 03:26:00||Quad Cities||NRC Region 3||On October 9, 2018 at 2002 CDT the Control Room Emergency Ventilation Air Condition (CREV AC) system was in the process of being returned to service following maintenance. During the return to service, the end bell on the CREV AC Condenser developed a significant leak requiring isolation. No work was performed on the CREV AC Condenser during the work window. The CREV AC system maintains a habitable control room environment and ensures the operability of components in the control room emergency zone during accident conditions. This notification is being made in accordance with 10CFR50.72(b)(3)(v)(D), "Event or Condition That Could Have Prevented Fulfillment of a Safety Function " because the CREV system is a single train system required to mitigate the consequences of an accident. The NRC Resident Inspector has been notified.|
|ENS 53652||8 October 2018 04:13:00||Braidwood||NRC Region 3|
On Monday, October 8, 2018 at 0111 CDT, during the initial containment entry for unit 2 refueling outage (A2R20), reactor coolant system pressure boundary leakage was discovered at the 2D Steam Generator bowl drain line. Unit cooldown to mode 5 is in progress.
This event is reportable under 10CFR50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.'
The licensee has notified the NRC resident inspector. Approximately 0.1 gpm was leaking from the drain line. LCO 3.4.13 was entered and the licensee anticipates being in mode 5 within a couple of hours. The leak will be repaired prior to exiting the refueling outage.
|ENS 53630||30 September 2018 15:29:00||Peach Bottom||NRC Region 1||On Sunday, September 30, 2018, at 1130 EDT, an automatic scram was received on U3 following a loss of two condensate pumps. Following the reactor scram, water level lowered from normal level of 23" to below 1" which resulted in automatic Group II and Group III isolations. Reactor water level lowered to -48" which resulted in initiation of the High Pressure Coolant Injection and Reactor Core Isolation Cooling systems. Reactor water level and reactor pressure have been restored to their normal bands. All systems responded properly to the event. Unit 3 remains in Mode 3 with reactor pressure being controlled on the turbine bypass valves. The cause and details of the event are under investigation. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). All control rods inserted. Decay heat is being removed via the main condenser. The NRC Resident Inspector has been notified. A notification to the media and a press release were made. Unit 2 was unaffected and continues coastdown to refueling.|
|ENS 53626||26 September 2018 23:25:00||Quad Cities||NRC Region 3|
On September 26, 2018 at 1908 CDT. an automatic scram was received on U1 following main generator 345 kV output breaker 7-8 trip with 345 kV output breaker 6-7 already opened for maintenance on line 0401. Following the reactor scram, reactor water level decreased to approximately minus 15 inches, which resulted in automatic Group II and Group Ill isolations (expected response). Reactor pressure rose to approximately 1083 psig, and the 3B and 3C low set relief valves opened briefly to control reactor pressure. Reactor water level and reactor pressure have been restored to their normal bands. All systems responded properly to the event. Unit 1 remains in Mode 3, with reactor pressure being controlled on the turbine bypass valves. The cause and details of the event are under investigation.
Unit 2 was unaffected by the event and remains at 100% power. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A)." All control rods inserted. Decay heat is being removed via the main condenser. The licensee notified the NRC Resident Inspector.
|ENS 53623||26 September 2018 15:10:00||Clinton||NRC Region 3||At 0946 CDT on 9/26/2018, a disruption in power to the offsite 138 kV line and the subsequent trip of the Emergency Reserve Auxiliary Transformer (ERAT) Static VAR Compensator (SVC) resulted in a degraded voltage signal on the Division 1- 4.16 kV safety bus. The degraded voltage signal resulted in a trip of the ERAT feed to the bus, blocking closure of the 345 kV Reserve Auxiliary Transformer (RAT) feed to the bus and auto start of the Division 1 Emergency Diesel Generator (EDG). The Division 1 EDG successfully started and re-energized the Division 1- 4.16 kV bus as designed. The unit is stable with the Division 1 EDG carrying the Division 1- 4.16 kV bus. The Ameren Transmission System Operator in St. Louis, MO informed the station that they had received a report that a 138 kV to 13.8 kV transformer at Clinton Route 54 substation was on fire and the South feed to the Tabor substation cycled as a result of this fault. The NRC Resident Inspector and Illinois Emergency Management Agency Resident Inspector have been notified."|
|ENS 53594||11 September 2018 17:14:00||Peach Bottom||NRC Region 1||On September 11, 2018 the Technical Support Center (TSC) ventilation was discovered to be non-functional during system testing. At 1310 EDT, an air leak was identified that prevented the modulating dampers to operate as designed to maintain required pressure. The air leak was not able to be repaired within a 60 minute period. This failure affected the ability of the TSC ventilation system to maintain adequate radiological habitability in the event of an emergency with an airborne radiological release. All other capabilities of the TSC are unaffected by this emergent condition. The air leak has been repaired and post maintenance testing was completed satisfactory at 1654 (EDT), restoring the TSC ventilation to a functional condition. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the potential loss of an emergency response facility because of the unavailability of the ventilation system. The NRC Resident Inspector has been notified.|
|ENS 53584||6 September 2018 17:18:00||Nine Mile Point||NRC Region 1||Pursuant to 50.73(a)(1) the following information is provided as a sixty (60) day telephone notification to the NRC. This notification, reported under 50.73(a)(2)(iv), is being provided in lieu of the submittal of a written LER (Licensee Event Report) to report a condition that resulted in an invalid actuation of the high pressure coolant injection (HPCI). At Nine Mile Point Unit 1, HPCI is a flow control mode of the normal feedwater system and is not an emergency core cooling system. On March 19, 2018 Nine Mile Point Unit 1 (NMP1) was at 0 percent power and in cold shutdown in support of a planned maintenance outage. At approximately 0118 (EDT), a reactor water level transient initiated by the fill and vent of 12 Reactor Recirculation Pump (12 RRP) occurred. During the fill and vent, Reactor Pressure Vessel (RPV) level lowered quickly from the initial level of 68 inches and a low level alarm was received. Control Room Operators reduced Reactor Water Clean-Up (RWCU) reject flow to turn the level trend and clear the low level alarm generated off of the compensated, GEMAC, level instrumentation. RWCU reject flow was reduced by 50 percent which caused RPV level to start to rise. RPV level was raised to approximately 72 inches at which time the Reactor Operator began to raise reject flow to reestablish the normal level band. During the RPV level transient, with actual water level at 74 inches on the GEMAC, the Yarway level instrumentation, which is not density compensated and therefore invalid, reached 92 inches causing an invalid high RPV water level turbine trip signal and associated invalid HPCI initiation signal. At no point in time did actual RPV water level reach the high RPV water level turbine trip set point of 92 inches. The potential for a turbine trip signal to occur due to shutdown activities was understood and tags were hung to lockout the Feedwater Pumps to prevent the HPCI start signal. Therefore, no HPCI injection occurred. The Licensee has notified the NRC Resident Inspector."|
|ENS 53576||31 August 2018 23:26:00||LaSalle||NRC Region 3||This notification is being provided in accordance with 10 CFR 50.72(b)(2)(iv)(B). On August 31, 2018 at 2105 CDT, Unit 2 Reactor Manual Scram signal was inserted due to Main Condenser vacuum degrading. The turbine was tripped following the scram. Main Condenser vacuum is at 6 inches of backpressure slowly improving following the scram and turbine trip. During the scram, one Control Rod (30-31) did not fully insert. Control Rod 30-31 has been manually inserted to position 00 with the first position identified as position 24. Plant is in a stable condition with reactor pressure being maintained by the Turbine Bypass valves. Reactor water level is being controlled with feedwater. Investigation into the cause of the elevated condenser in leakage is in progress. The Senior NRC Resident has been notified."|
|ENS 53565||27 August 2018 03:12:00||Nine Mile Point||NRC Region 1|
EN Revision Text: AUTOMATIC SCRAM DUE TO A GENERATOR TRIP At 0033 EDT Nine Mile Point Unit 2 experienced an automatic scram on high reactor pressure due to a turbine trip. The cause of the turbine trip was due to a generator trip. All control rods inserted. There were no safety system actuations. The cause of the generator trip is being investigated. This is a 4-Hour report for 10CFR50.72(b)(2)(iv)(B) RPS Actuation. The NRC Resident Inspector has been notified. Decay heat is being removed via the main condenser. Reactor vessel water level is being maintained by the condensate and feedwater systems. The licensee will be notifying the state of New York.
After further review, the licensee has determined that the cause of automatic scram was due to turbine control valve fast closure as a result of the turbine trip, not high reactor pressure, as originally reported. The licensee has notified the NRC Resident Inspector. Notified R1DO (Lilliendahl).
|ENS 53556||22 August 2018 02:00:00||Limerick||NRC Region 1||At 2322 EDT, Limerick Generating Station notified the Pennsylvania DEP (Department of Environmental Protection) that our plant waste water pond (holding pond) overflowed due to heavy rainfall in the area. Plant alignment changes were made and the holding pond stopped overflowing at 0017 EDT. Limerick Generating Station has not determined this release to contain oil, grease, or pollutants hazardous to the public. The licensee notified the NRC Resident Inspector.|
|ENS 53552||17 August 2018 08:17:00||Three Mile Island||NRC Region 1|
EN Revision Text: FITNESS-FOR-DUTY TEST POSITIVE FOR NON-LICENSED EMPLOYEE At 1519 EDT on August 16, 2018, Exelon determined a non-licensed employee had a confirmed positive for a controlled substance during a random Fitness-for-Duty test. The employee's unescorted access to the plant has been terminated. The NRC Resident Inspector has been notified.
The following is a correction to the reason for the Fitness-for-Duty test: At 1519 EDT on August 16, 2018, Exelon determined a non-licensed employee had a confirmed positive for a controlled substance during a follow-up Fitness-for-Duty test. The employee's unescorted access to the plant has been terminated. The NRC Resident Inspector has been notified of this correction. Notified R1DO (Bower) and FFD E-mail group.
|ENS 53551||16 August 2018 12:08:00||Byron||NRC Region 3||In accordance with 10 CFR 26.719(b)(2)(ii), this notification reports a licensed Operations employee had a confirmed positive for alcohol during a random fitness for duty test. The individual was not in the protected area and not performing licensed duties at the time of discovery. The employee's access to the plant has been suspended. The NRC Resident Inspector has been notified."|
|ENS 53493||7 July 2018 08:29:00||Calvert Cliffs||NRC Region 1||At 0242 EDT, the CCNPP (dual unit, single control room) control room supply damper failed shut. This rendered the Unit 1 and Unit 2 control room ventilation inoperable and the appropriate LCOs were entered. At 0249, control room ventilation was restored to service and the appropriate LCOs were exited. This is being reported under 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector.|
|ENS 53491||6 July 2018 15:36:00||Byron||NRC Region 3||At 1201 (CDT), Station Auxiliary Transformer 242-2 experienced a bushing failure, resulting in a loss of offsite power to Unit 2. The 2A and 2B Diesel Generators started and sequenced loads onto the Unit 2 ESF buses appropriately. All other buses normally powered from the Station Auxiliary Transformers automatically transferred to the Unit Auxiliary Transformers. ESF Bus 241 and 242 Undervoltage Relays actuated to start the Diesel Generators and the 2A Auxiliary Feedwater Pump started on the 2A Diesel Generator sequencer. ESF Battery Charger 212 tripped at the same time, which was an unexpected condition. DC Bus 212 was cross-tied with DC Bus 112. This notification is being made under 10 CFR 50.72(b)3(iv)(A) due to the actuation of both Unit 2 Diesel Generators and the 2A Auxiliary Feedwater Pump. The NRC Senior Resident Inspector has been notified. Currently, offsite power was restored via the Unit 1 Unit Auxiliary Transformer. Both Unit 2 Emergency Diesel Generators have been secured. DC Busses are still cross-tied. The licensee is currently in a 72-hour shutdown action statement for the loss of offsite power and a 7-day action statement for having the Unit 2 DC Bus cross-tied to Unit 1.|
|ENS 53463||20 June 2018 17:51:00||Clinton||NRC Region 3||On June 20, 2018, at 1145 hours (CDT), during panel walkdown, it was identified that High-Pressure Core Spray (HPCS) injection valve 1E22F004 was in the open position. Valve 1E22F004 is normally closed for containment integrity purposes. Operations personnel verified that the valve was open locally and that the plant computer indicated the valve is in the 'not closed' position. No alarms or status lamps indicated why the valve would be open and there was no valid demand signal. Reactor power, pressure, level, and feedwater parameters remain steady and unchanged, with no indication of HPCS injection having occurred or in progress. A low-water level signal, or a high drywell pressure signal, or manual operation initiates HPCS. When a high-water level in the reactor vessel is detected, HPCS injection is automatically stopped by a signal to close injection valve 1E22F004. With valve 1E22F004 in the open position without a demand signal, closure on a high reactor water level condition was not assured. Therefore, HPCS was declared inoperable. The following Technical Specifications were entered: 3.5.1, Emergency Core Cooling Systems (ECCS) - Operating and 188.8.131.52, Primary Containment Isolation Valves (PCIVs). Subsequently, HPCS injection valve 1E22F004 was observed to be cycling without operator action. The valve was deactivated in the closed position to assure the containment isolation function. The cause of valve 1E22F004 cycling without operator action is under investigation. HPCS is a single train safety system that consists of a single motor-driven pump, a spray sparger in the reactor vessel, and associated piping, valves, controls and instrumentation. HPCS is part of the ECCS network, which also includes Low-Pressure Core Spray, Low-Pressure Coolant Injection, and the Automatic Depressurization system. This event is being reported as an 8-hour non-emergency notification per 10 CFR 50.72(b)(3)(v) as, 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident.' The licensee notified the NRC Resident Inspector.|
|ENS 53443||4 June 2018 14:00:00||Braidwood||NRC Region 3||At 0920 CDT, Braidwood Unit 1 reactor was manually tripped due to lowering steam generator water levels following a trip of the 1C main feedwater pump. The cause of the 1C main feedwater pump trip is unknown at this time and is under investigation. Both trains of Braidwood Unit 1 auxiliary feedwater started automatically following the reactor trip to maintain steam generator water levels. All systems responded as expected. Steam generator power operated relief valves lifted momentarily and reseated as designed in response to the secondary transient due to the reactor trip. The main steam dump valves are in service to the main condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by offsite power with the diesel generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr. notification, and per 10CFR50.72(b)(3)(iv)(A) for an automatic actuation of the auxiliary feedwater system, 8-hr. notification. All rods inserted into the core during the trip. Concerning the relief valves momentarily lifting and reseating, there is no known primary-to-secondary leakage. The licensee has notified the NRC Resident Inspector.|
|ENS 53401||14 May 2018 13:17:00||Calvert Cliffs||NRC Region 1||CE||On May 14, 2018, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of a tornado generated missile, Calvert Cliffs identified a non-conforming condition in the plant design such that specific TS equipment is considered to not be adequately protected from a tornado generated missile. A tornado could generate a missile that could strike the Unit 1 Saltwater system header and associated piping. This could result in damage to the unit 1 Saltwater system header which could affect the ability of the Unit 1 Saltwater subsystems to perform their design function if such a tornado would occur. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(D) 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 mitigate the consequences of an accident. This condition is being addressed in accordance with NRC enforcement guidance provided in EGM 15-002 and DSS-ISG-2016-01. Compensatory measures have been implemented in accordance with these documents. The NRC Resident Inspector has been informed of this notification.|
|ENS 53395||10 May 2018 08:59:00||Nine Mile Point||NRC Region 1||GE-5||At 0248 (EDT), with the plant shutdown in Mode 4, Nine Mile Point Unit 2 experienced a partial loss of off-site power during relay testing that resulted in an automatic start of the Division 2 Emergency Diesel Generator. All systems responded as expected for the event. The cause is being investigated. The station responded in accordance with appropriate Special Operating Procedures and restored impacted systems. This event is being reported in accordance with 10CFR50.72(b)(3)(iv)(A) At the time of the report, the emergency diesel generators are loaded and supplying plant safety equipment. The licensee has notified the state of New York Emergency Management Agency and the NRC Resident Inspector.|
|ENS 53371||30 April 2018 14:53:00||Braidwood||NRC Region 3||Westinghouse PWR 4-Loop||At 1124 CDT, Braidwood Unit 1 experienced an automatic Reactor Trip. The cause of the Reactor Trip was a Turbine Trip with reactor power greater than P-8. The turbine trip was actuated as a result of a Turbine Motoring Generator Trip. The cause of the generator trip is unknown at this time and is under investigation. After the Reactor Trip occurred, the 1A Auxiliary Feedwater pump was manually started to provide feedwater flow to all four steam generators. The 1A Auxiliary Feedwater pump was subsequently secured and placed in standby when the Startup Feedwater pump was placed in service. Train A Main Control Room Ventilation Filtration system shifted to Makeup Mode due to a spurious actuation signal. No secondary relief valves lifted and no secondary steam was released as a result of the Reactor Trip. The Main Steam dump valves are in service to the Main Condenser to provide heat sink cooling. The plant is being maintained at normal operating pressure and temperature. AC power is being provided by Offsite Power with the Diesel Generators in standby and all safety systems available. There is no impact to Unit 2. This report is being made per 10 CFR 50.72(b)(2)(iv)(B) for a RPS actuation, 4-hr notification, and per 10 CFR 50.72(b)(3)(iv)(A) for a manual actuation of the Auxiliary Feedwater system, 8-hr notification. The licensee notified the NRC Resident Inspector and Illinois Emergency Management Agency.|
|ENS 53358||22 April 2018 22:40:00||Braidwood||NRC Region 3||Westinghouse PWR 4-Loop||On Sunday, April 22, 2018 at 1646 CDT, a valid actuation of Engineered Safety Feature (ESF) Bus 141 Undervoltage (UV) Relay occurred. At the time, Braidwood Station Unit 1 was performing a pre-planned 1A Diesel Generator (DG) Emergency Core Cooling System (ECCS) Actuation Surveillance, initiating the 1A DG to emergency start and sequence loads on a safety injection signal. Following the 1A DG solely supplying electrical power to Bus 141, the 1A DG lost voltage, resulting in an unplanned UV actuation of ESF Bus 141. The 1A DG output breaker was manually opened and local emergency stop of the 1A DG was attempted. The 1A DG continued to run at idle. Fuel supply was secured to the 1A DG and the engine stopped. Subsequently, operators restored power to ESF Bus 141 from the Unit 1 Offsite Power Source. Shutdown cooling was maintained throughout the event as the 1B Residual Heat Removal train was unaffected by the actuation. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for 'Any event or condition that results in valid actuation of any of the systems listed...', specifically 10 CFR 50.72(b)(3)(iv)(B)(8) for the 'Emergency ac electrical power systems, including: emergency diesel generators (EDGs)...'. The licensee notified the NRC resident inspector.|
|ENS 53354||20 April 2018 22:22:00||Braidwood||NRC Region 3||Westinghouse PWR 4-Loop||On Friday, April 20, 2018 at 1730 CDT, during the Braidwood Station Unit 1 refueling outage (A1R20), a scheduled ultrasonic test (UT) was performed on the top head to upper center disc weld of the Unit 1 reactor head. The UT identified 19 indications, 9 of which are not acceptable per ASME Section XI, 2001 Edition, 2003 Addenda, Paragraph IWB-3510. This event is reportable under 10 CFR 50.72(b)(3)(ii)(A) for 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded'. The licensee notified the NRC Resident Inspector.|
|ENS 53353||20 April 2018 17:57:00||Braidwood||NRC Region 3||Westinghouse PWR 4-Loop||On Friday, April 20, 2018 at 1042 CDT, Braidwood Station Unit 1 was at 0 percent power in Mode 6. The 1A Diesel Generator (DG) was inoperable with troubleshooting in progress. The 1B DG was being run for a normal monthly run in accordance with 1 BwOSR 184.108.40.206-2, 'Unit One 1B Diesel Generator Operability Surveillance,' and subsequently tripped. The trip was due to a failure of the overspeed butterfly valve actuator and springs, and not an actual overspeed condition. The unit entered Technical Specification (TS) 3.8.2, 'AC Sources - Shutdown,' Condition B for required DG inoperable. All required TS actions were met at the time of the 1B DG inoperability. The offsite power source remains available. At no time was residual heat removal lost. This event is reportable under 10 CFR 50.72(b)(3)(v)(B) 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 remove residual heat. The licensee has notified the NRC Resident Inspector.|