Semantic search
| Start date | Site | Title | Description | System | |
|---|---|---|---|---|---|
| ENS 57036 | 19 March 2024 15:30:00 | Browns Ferry | High Pressure Coolant Injection Isolation | The following information was provided by the licensee via fax or email: While performing a planned high pressure coolant injection (HPCI) system surveillance, an isolation signal was received based upon an exhaust rupture disc high pressure signal. This resulted in an unplanned inoperability of the HPCI system. All systems responded as expected, and the event is under investigation. No other systems were affected by this condition. This event is reportable as an 8-hour non-emergency notification under 10CFR50.72(b)(3)(v) as HPCI is a single train safety system. There was no impact to plant personnel or the public as a result of this condition. The NRC resident has been notified of this condition. | High Pressure Coolant Injection |
| ENS 57041 | 13 March 2024 10:48:00 | Cooper | High Pressure Coolant Injection (HPCI) Inoperable | The following information was provided by the licensee via email and phone: At 0548 CDT on March 13, 2024, during a planned (high pressure coolant injection) HPCI maintenance window, a condition was identified not associated with the planned maintenance which caused HPCI to be inoperable. Specifically, the HPCI auxiliary oil pump start stop pressure switch could not be adjusted into calibration. Further investigation found that the pressure switch was not mounted as designed. Since HPCI is a single train system, this is a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The condition was corrected prior to HPCI being declared operable on March 15, 2024. The reason for the delay in the event notification beyond 8 hours from the event time was due to not recognizing the need to report the condition while in a planned HPCI maintenance window. The NRC Senior Resident Inspector has been notified. | High Pressure Coolant Injection |
| ENS 57021 | 11 March 2024 17:37:00 | Hatch | Manual Reactor Trip | The following information was provided by the licensee via phone and email: On March 11, 2024, at 1337 EDT, with Unit 1 in Mode 1 at 35 percent power performing power ascension activities, the reactor was manually tripped due to the 'A' reactor feed pump (RFP) tripping on low suction pressure. Due to the power level at the time, the 'B' RFP had not been placed in service. Closure of containment isolation valves (CIVs) in multiple systems and actuation of high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) occurred as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. The 'B' RFP was placed in service and is controlling reactor water level. Decay heat is being removed by discharging steam to the main condenser using turbine bypass valves. Unit 2 is not affected. Due to the emergency core cooling system (ECCS) discharging into the reactor, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). Also, the Reactor Protection System actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, it is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs, RCIC and HPCI. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the 'A' RFP is under investigation. The reactor electric plant remains in a normal lineup with both emergency diesel generators available. There were no temperature or pressure technical specification limits approached. | Reactor Protection System Emergency Diesel Generator Reactor Core Isolation Cooling Emergency Core Cooling System Main Condenser |
| ENS 57004 | 4 March 2024 00:42:00 | Nine Mile Point | Automatic Reactor Scram Due to Main Turbine Trip on LOW Condenser Vacuum | The following information was provided by the licensee via email: On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram. Additionally, following the scram a low RPV (reactor pressure vessel) level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR (residual heat removal) shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system. This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration. | Shutdown Cooling Decay Heat Removal Control Rod |
| ENS 57002 | 1 March 2024 19:45:00 | Monticello | Inadvertent Siren Activation | The following information was provided by the licensee via email: At 1330 CST, on March 1, 2024, an equipment vendor was coordinating with Wright County performing maintenance on an emergency siren when the county operator mistakenly sent an alarm signal instead of cancel signal, activating all Wright County emergency sirens for approximately 17 seconds. At 1345 CST, the Monticello Nuclear Generating Plant (MNGP) emergency planning coordinator received a notification from the vendor and notified the duty shift manager (of the inadvertent activation). Wright County officials are planning to make a public notification via social media to local residents. No press release by the licensee is planned at this time. This event is reportable per 10 CFR 50.72(b)(2)(xi), 'News Release or Notification of Other Government Agencies.' This is a 4-hour Reporting requirement. The NRC Resident has been notified. | |
| ENS 56995 | 28 February 2024 14:39:00 | Monticello | Automatic Reactor Scram | The following information was provided by the licensee via fax and email: At approximately 0839 (CST) with Unit 1 in Mode 1 at 100 percent power, the reactor automatically scrammed due to the depressurization of the SCRAM air header caused by an invalid signal that (occurred) during system testing. The SCRAM was uncomplicated with all systems responding as expected. The cause and details of the event are under investigation. Containment isolation valves actuated and closed on a valid Group 2 signal. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B), and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group 2 isolation signal. Operations responded using the emergency operating procedure and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. State as well as Wright and Sherburne Counties will be notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Anticipated Transient Without Scram (ATWS) circuit was being tested when an invalid signal was sent to depressurize the SCRAM air header. | Reactor Protection System Main Condenser |
| ENS 56990 | 24 February 2024 08:19:00 | Browns Ferry | Automatic Start of Emergency Diesel Generator | The following information was provided by the licensee via phone and email: At 0219 CST on February 24, 2024, Browns Ferry Unit 3 was shut down in a refueling outage, while closing 4 kV shutdown board breaker 3EB-9, the 4 kV shutdown board normal feeder breaker tripped open resulting in a valid 4 kV bus under-voltage condition. Due to the under-voltage condition, the 3B emergency diesel generator (EDG) auto started and tied to the board. The cause of the breaker tripping open is unknown and an investigation is in progress. All systems responded as expected for the loss of voltage. This event requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No other safety related equipment was affected. The 3B EDG continues to supply the shutdown board pending further investigation. | Emergency Diesel Generator |
| ENS 56989 | 22 February 2024 17:03:00 | Cooper | Accident Mitigation - High Pressure Coolant Iinjection Declared Inoperable | The following information was provided by the licensee via email: At 1103 CST on February 22, 2024, a potential through-wall steam leak was identified on the high pressure coolant injection (HPCI) steam supply 1-inch drain line. As a result, HPCI was declared inoperable. Since HPCI is a single-train system, this is a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). Reactor core isolation cooling (RCIC) and low pressure emergency core cooling systems (ECCS) remain operable. Additional investigation is in progress. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Reactor Core Isolation Cooling Emergency Core Cooling System |
| ENS 56980 | 19 February 2024 15:45:00 | Peach Bottom | Loss of Reactor Building Ventilation | The following information was provided by the licensee via email: At 1045 EST, on 2/19/2024, during a maintenance activity, a loss of all reactor building ventilation occurred on Unit 2. With no flow past the ventilation radiation monitors, the radiation monitors were inoperable to support their ability to perform primary and secondary containment isolation functions or start the standby gas treatment system. Reactor building ventilation was restored within 15 minutes. Due to this inoperability, the radiation monitor system was in a condition that could have prevented fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector will be notified.
Upon further investigation, it was verified that the reactor building and the refueling floor radiation monitors are not needed to control the release of radiation for events described in chapter 14 of the updated Final Safety Analysis Report. For the analyzed loss of coolant accident (LOCA), the primary and secondary signals for this purpose were available and unaffected by this event. The radiation monitors provide a tertiary redundant method that is not credited within the station analysis. For all other analyzed accidents, the signal provided by the radiation monitors is not needed, as the secondary containment isolation function and start of the standby gas treatment system are not credited. Additionally, the fuel handling accident was not credible during the time of the event because no activities were in progress on the refueling floor. Therefore, the threshold for reporting the issue as an event or condition that could have prevented the fulfillment of a safety function was not met. The NRC Resident Inspector has been notified. Notified R1DO (Jackson) | Secondary containment Reactor Building Ventilation Standby Gas Treatment System |
| ENS 56977 | 19 February 2024 04:25:00 | Brunswick | Automatic Start of Emergency Diesel Generator | The following information was provided by the licensee via phone and email: At approximately 2325 EST on February 18, 2024, with Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 100 percent power, emergency diesel generator 2 automatically started due to the unexpected loss of AC power to emergency bus E2 during a planned transfer of E2 DC control power from normal to alternate for the 1B-1 battery. In addition, the unexpected loss of AC power to E2 resulted in Unit 1 primary containment isolation system (PCIS) partial Group 2 (i.e., drywell equipment and floor drain, residual heat removal (RHR), discharge to radioactive waste, and RHR process sample), Group 6 (i.e., containment atmosphere control/dilution, containment atmosphere monitoring, and post accident sampling systems), and partial Group 10 (i.e., air isolation to the drywell) isolations. Emergency diesel generator 2 automatically started and re-energized the E2 bus as designed when the loss of E2 signal was received. The PCIS actuations were as expected for the outage plant line up on Unit 1 at the time. The cause of the loss of electrical power to emergency bus E2 is under investigation at this time. 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 emergency diesel generator 2 and PCIS. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: This event will be entered into the plant's corrective action program. | Emergency Diesel Generator Primary Containment Isolation System Residual Heat Removal |
| ENS 56975 | 18 February 2024 16:33:00 | Hatch | Fitness for Duty | The following is a synopsis of information was provided by the licensee via email and phone call: A non-licensed supervisor had a confirmed positive during a random fitness for duty test. The supervisor's access to the plant has been terminated. | |
| ENS 56974 | 17 February 2024 13:37:00 | Brunswick | Primary Containment Degraded | The following information was provided by the licensee via email and phone call: At 0837 EST, on 02/17/2024, during a refueling outage at 0 percent power while performing local leak rate testing (LLRT) on the reactor core isolation cooling (RCIC) isolation valves, which is part of the containment boundary, it was determined that the Unit 1 primary containment leakage rate did not meet 10 CFR 50 Appendix J requirements specified in Technical Specification 5.5.12. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Core Isolation Cooling Primary containment |
| ENS 56959 | 11 February 2024 15:11:00 | Hatch | Primary Containment Degraded | The following information was provided by the licensee via email: At 1011 EST on 02/11/2024, during a refueling outage at 0 percent power, while performing local leakage rate testing (LLRT) of the feedwater check valves (part of the containment boundary), it was determined that the Unit 1 primary containment leakage rate did not meet 10 CFR 50 Appendix J requirements specified in Technical Specification 5.5.12. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Feedwater Primary containment |
| ENS 56957 | 9 February 2024 18:22:00 | Peach Bottom | Unanalyzed Condition - Inadequate Fuses for Fuel Pool Cooling | The following information was provided by the licensee via email: On 2/9/24 at 1322 EST, it was determined that the unit was in an unanalyzed condition. A review of DC feeder circuit protection schemes identified a circuit for the fuel pool cooling system is uncoordinated due to inadequate fuse sizing. This results in a concern that postulated fire damage in one area could cause a short circuit without adequate protection, leading to the unavailability of equipment credited for in 10 CFR 50 Appendix R, Fire Safe Shutdown. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The postulated event affects the following fire zones: fire areas 6S and 6N (within the Unit 2 reactor building). Compensatory actions for affected fire areas have been implemented. An extent of condition review is being performed. The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Fire watches have been established in the affected areas. These will be maintained until the protection scheme is revised.
The following updated information was provided by the licensee via email and phone call: On 03/08/24 at 1418, extent of condition reviews identified circuit(s) in the Units 2 and 3 Reactor Protection Systems (RPS) which are also uncoordinated due to improper fuse sizing. These circuits are not bounded by existing design and licensing documents for 10 CFR 50 Appendix R Fire Safe Shutdown and, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel. The postulated event affects the following fire areas: 32, 33, 38 and 39 (Units 2 and 3 Switchgear Rooms). In accordance with procedural requirements, compensatory actions for the affected fire areas have been implemented and will remain until the condition is resolved. The NRC Senior Resident Inspector has been notified. Notified R1DO (Arner)
On March 13, 2024, at 1350 EDT, extent of condition reviews identified a circuit in the Unit 2 reactor protection system (RPS) which is also uncoordinated due to improper fuse sizing. This circuit is not bounded by existing design and licensing documents for 10 CFR 50, Appendix R, Fire Safe Shutdown, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel. The postulated event affects fire area 57 (Switchgear Corridor, common to Units 2 and 3). In accordance with procedural requirements, compensatory actions for the affected fire areas have been implemented and will remain until the condition is resolved. Additionally, it was previously reported that fire area 6N contained a circuit which was not bounded by the Fire Safe Shutdown analysis; however, after further review it has been determined that compliance is maintained in this fire area and is therefore retracted from the scope of this report. The NRC Senior Resident Inspector has been notified. Notified R1DO (Jackson)
The following information was provided by the licensee via email: On 03/21/24 at 1211, extent of condition reviews identified an annunciator circuit for the Unit 3 emergency service water (ESW) and high pressure service water (HPSW) pump structure heating and ventilation panel that is also uncoordinated due to improper fuse sizing. This circuit is not bounded by existing design and licensing documents for 10 CFR 50 Appendix R Fire Safe Shutdown and, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel. The postulated event affects fire area 47 (Unit 3 pump structure for `B' ESW and `3A'-`3D' HPSW pumps) and the yard fire area (Manhole 026D). In order to restore immediate compliance, the cable has been de-energized to eliminate the possibility of the event of concern. This circuit will remain de-energized or other measures will be implemented until the condition is permanently resolved. The NRC Senior Resident Inspector has been notified. Notified R1DO (Ford) | Service water Reactor Protection System |
| ENS 56954 | 8 February 2024 16:05:00 | Quad Cities | Fitness for Duty Violation | The following information was provided by the licensee via email: A supervisor had a confirmed positive for alcohol during a random fitness for duty test. The supervisor's access to the plant has been terminated. | |
| ENS 56952 | 7 February 2024 21:00:00 | Hope Creek | Fitness for Duty Programatic Issue | The following information was provided by the licensee via email: A programmatic vulnerability, failure, or degradation was discovered within the fitness for duty (FFD) program that may permit undetected drug or alcohol use or abuse by individuals within the protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program. Public and plant safety have not been affected. The NRC Resident Inspector was notified. | |
| ENS 56941 | 1 February 2024 18:17:00 | Browns Ferry | Offsite Notification - Workplace Injury | The following information was provided by the licensee via email: On February 1, 2024, a contract worker was transported offsite for medical treatment due to a work-related injury that required the individual to be admitted to the hospital. The individual was free-released from the site prior to transport. The injury and hospitalization were reported by the contract worker's employer to OSHA per 29 CFR 1904.39(a)(2). Based upon that notification to another government agency, Tennessee Valley Authority is reporting this per 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified of this event.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous Event Notification, EN 56941, reported on 02/01/2024. On 02/01/2024, at 15:32 EST, Browns Ferry Nuclear Plant (BFN) made an Event Notification 56941 notifying the NRC of a notification to another government agency. During further review of NRC reporting guidance, BFN has concluded that the contract worker's employer report to OSHA was below the reporting threshold outlined in NUREG 1022, Revision 3. The NRC Resident Inspector has been notified. | |
| ENS 56936 | 29 January 2024 17:02:00 | Peach Bottom | Automatic Reactor Scram | The following information was provided by the licensee via email: At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram caused by a main turbine trip. Investigation is still ongoing. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods were fully inserted. The licensee indicated that the turbine trip may have been caused by a power load imbalance, however the cause of the incident is under investigation. The scram was not complex. Decay heat is currently being removed thru bypass valves dumping to the main condenser. Initially unit 2 lost the use of the bypass valves due to lack of condenser vacuum. Unit 2 used the high pressure coolant injection (HPCI) system in the condenser storage tank (CST) to CST mode to remove decay heat. Residual heat removal was used to keep the torus cool. Condenser vacuum was regained and unit 2 is back to removing decay heat with the turbine bypass valves. There was no impact to unit 3. The licensee confirmed there was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: Licensee adds 8-hour non-emergency 10 CFR 50.72(b)(3)(iv)(A) specified system actuation report to original 4-hour non-emergency 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation report. At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram by a main turbine trip. All control rods inserted. Reactor core isolation cooling system (RCIC) was manually initiated for level control. HPCI was manually initiated for pressure control. Primary containment isolation system (PCIS) Group II and III isolations occurred (specified system actuation). Investigation is ongoing. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Primary Containment Isolation System Reactor Core Isolation Cooling Residual Heat Removal Main Condenser Control Rod |
| ENS 56987 | 2 January 2024 04:33:00 | Brunswick | Invalid Actuation of Containment Isolation Valves | The following information was provided by the licensee via phone and email: This 60-day optional telephone notification is being made in lieu of a Licensee Event Report (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 2333 EST on January 1, 2024, an invalid actuation of group 6 primary containment isolation valves (PCIVs) (i.e., containment atmospheric control/monitoring (CAC/CAM) and post-accident sampling system (PASS) isolation valves) occurred. Reactor building ventilation isolated and standby gas treatment started per design. No manipulations associated with the isolation or reset logic were ongoing at the time. Troubleshooting determined that the group 6 isolation signal resulted from spurious relay contact actuation in the main stack radiation high-high isolation logic due to relay contact oxidation. The main stack radiation monitor is a shared component that sends isolation signals to Unit 1 and Unit 2. There were no Unit 1 actuations. Only the relay contacts associated with Unit 2 actuated. The relay has been replaced. 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. During this event the PCIVs functioned successfully, and the actuations were complete. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector had been notified. | Primary containment Reactor Building Ventilation |
| ENS 56988 | 28 December 2023 13:15:00 | Brunswick | Invalid Actuation of Emergency Diesel Generators | The following information was provided by the licensee via phone and email: This 60-day optional telephone notification is being made in lieu of a Licensee Event Report (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 0815 EST on December 28, 2023, an invalid actuation of the four emergency diesel generators (EDGs) occurred. It was determined that this condition was likely caused by spurious operation of the undervoltage relay for the startup auxiliary transformer feeder breaker to the `1D' balance of plant bus which was being fed by the unit auxiliary transformer at the time, per the normal lineup. This non-safety related EDG actuation logic was disabled, and additional investigation is planned during the upcoming refueling outage. 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. During this event, the four EDGs functioned successfully, and the actuations were complete. All emergency buses remained energized from offsite power and, therefore, the EDGs did not tie to their respective buses. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector had been notified. | Emergency Diesel Generator |
| ENS 56896 | 18 December 2023 07:23:00 | Hatch | High Pressure Coolant Injection System Inoperable | The following information was provided by the licensee email: At 0223 EST, on 12/18/2023, while Unit 2 was at 100 percent power in mode 1, the high pressure coolant injection (HPCI) outboard steam isolation valve closed resulting in the HPCI system being declared inoperable. The cause of the outboard steam isolation valve closing is under investigation. HPCI does not have a redundant system, therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The safety function was restored at 0512, on 12/18/23, and HPCI has been declared operable. Reactor core isolation cooling (RCIC) and low pressure emergency core cooling systems (ECCS) were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | High Pressure Coolant Injection Reactor Core Isolation Cooling Emergency Core Cooling System |
| ENS 56889 | 15 December 2023 00:39:00 | Hope Creek | Automatic Reactor Scram | The following information was provided by the licensee via phone call and email: On December 14, 2023, at 1939 EST, Hope Creek reactor scrammed following closure of turbine control valve number 4. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The outage control center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. | Feedwater Main Condenser Control Rod |
| ENS 56866 | 20 November 2023 15:56:00 | Dresden | HPCI Declared Inoperable | The following information was provided by the licensee via email: At 0956 (CST) on November 20, 2023, accumulated gas was identified in the Dresden Unit 2 high pressure coolant injection (HPCI) system discharge header. As a result, the HPCI system was declared inoperable. Since HPCI is a single-train system, this is a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The HPCI system was subsequently vented, and the accumulated gas has been removed, restoring the Dresden Unit 2 HPCI system to an operable status. All other emergency core cooling systems remained operable during this time period. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee administratively verified the isolation condenser was operable after declaring HPCI inoperable as required by technical specifications. The licensee stated there was no increase in plant risk. The cause of gas accumulating in the Dresden Unit 2 HPCI discharge header is under investigation, and this issue has been entered into the licensee's corrective action program. | High Pressure Coolant Injection Emergency Core Cooling System |
| ENS 56865 | 19 November 2023 02:20:00 | Quad Cities | Fitness for Duty Report | The following information was provided by the licensee via phone and fax: On November 18, 2023, the presence of alcohol was discovered inside the protected area. In accordance with the Constellation Fitness For Duty (FFD) Program, the individual has been escorted offsite and access to the plant denied pending the results of an investigation. This event is being reported under 10 CFR 26.719(b)(1) as it represents a significant FFD violation. The NRC Resident Inspector has been notified. | |
| ENS 56858 | 16 November 2023 14:06:00 | Brunswick | Failed Fitness for Duty Test | The following information was provided by the licensee via phone and email: At 0906 Eastern Standard Time (EST) on November 16, 2023, it was determined that a non-licensed employee supervisor failed a test specified by the Fitness for Duty (FFD) testing program. The individual's authorization for site access has been removed. The NRC Resident Inspector has been notified. | |
| ENS 56897 | 2 November 2023 01:11:00 | Hatch | 60 Day Notification for an Invalid Specified System Actuation | The following information was provided by the licensee via email and phone: At 2011 EDT on 11/01/23, with Unit 2 in Mode 3 at 0 percent power, Unit 2 received multiple spurious actuations. These actuations consisted of a partial group 1 and a partial group 5 primary containment isolation and a partial secondary containment isolation. The partial Group 1 isolation resulted in the closure of two main steam isolation valves (MSIVs); all other MSIVs were already closed. The partial group 5 isolation auto closed one of the reactor water cleanup (RWCU) isolation valves. The partial secondary containment isolation resulted in the closure of the inboard refueling floor and reactor building secondary containment isolation valves (SCIVs). Additionally, at 2238 EDT, Unit 2 again received multiple spurious actuations. These actuations consisted of a partial group 5 primary containment isolation and a partial secondary containment isolation. The partial group 5 isolation auto closed one of the RWCU isolation valves The partial secondary containment isolation resulted in the closure of the inboard refueling floor and reactor building SCIVs. And again, at 2354 EDT, Unit 2 received spurious actuations which consisted of a partial secondary containment isolation which resulted in the closure of the inboard refueling floor and reactor building SCIVs. The spurious actuations seen on 11/1/23 are triggered at -35 inches reactor water level (RWL) for group 5 and secondary containment isolations and at -101 inches RWL for group 1 isolations. It was determined that a combination of the RWL fluctuating above and below the wide range instrument reference leg tap, the reactor vessel pressure being lowered, and reactor core isolation cooling introducing colder water conditions near the reference leg tap of the wide range instrument caused the spurious actuations. Using multiple RWL indications for each of the instances mentioned above, the actuations were confirmed to be spurious as RWL was being controlled in a band of +55 inches to +85 inches at the time of the actuations. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an invalid actuation of a partial group 1, a partial group 5, and partial secondary containment logic. The NRC Resident has been notified. | Secondary containment Main Steam Isolation Valve Reactor Core Isolation Cooling Primary containment Reactor Water Cleanup |
| ENS 56826 | 1 November 2023 10:48:00 | Hatch | Manual Reactor Trip Due to Trip of Reactor Feed Pump | The following information was provided by the licensee via email: At 0648 EDT on 11/1/23, with Unit 2 in MODE 1 at 56 percent power, the reactor was manually tripped due to a trip of the 'B' reactor feed pump (RFP). The 'A' RFP had been previously isolated due to a leak. Closure of containment isolation valves (CIVs) in multiple systems and the actuation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) occurred as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all safety systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained with RCIC. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 was not affected. Due to the emergency core cooling system (ECCS) discharging into the reactor this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). Also, the reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, it is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs, RCIC and HPCI. There was no impact on the health and safety of the public or plant personnel. The Resident Inspector was notified. | High Pressure Coolant Injection Reactor Protection System Reactor Core Isolation Cooling Emergency Core Cooling System Main Condenser |
| ENS 56822 | 30 October 2023 16:00:00 | FitzPatrick | FITNESS-FOR-DUTY Report | The following information was provided by the licensee via phone call and email: A non-licensed supervisory employee had a confirmed positive test 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 56810 | 23 October 2023 00:48:00 | Nine Mile Point | Average Power Range Monitors Declared Inoperable | The following information was provided by the licensee via phone and email: On October 21, 2023, at 2048 EDT, reactor recirculation pump (RRP) 12 tripped. The cause for the trip is under investigation. Following the RRP trip, the average power range monitors (APRMs) flow bias trips were inoperable due to reverse flow through RRP 12. The APRMs were restored to operable on October 21, 2023, at 2058 EDT, when the RRP 12 discharge blocking valve was closed. This 8-hour non-emergency report is being made based upon requirements of 10CFR50.72(b)(3)(v)(A) which states: "Licensee shall notify the NRC of 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." The NRC Resident Inspector has been notified. | Reactor Recirculation Pump |
| ENS 56811 | 22 October 2023 16:49:00 | Cooper | Secondary Containment Momentarily Inoperable | The following information was provided by the licensee via fax and phone: On October 22, 2023, at 1149 CDT, with the reactor at 100 percent core thermal power and steady state conditions, the Cooper Nuclear Station secondary containment differential pressure exceeded the Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.1 limit of -0.25 inches water gauge. The condition existed for approximately 80 seconds until the reactor building ventilation system responded to restore differential pressure to normal. Investigations identified a hinged duct access hatch found open. The hatch was closed and latched, and ventilation system parameters were returned to normal. There were no radiological releases associated with this event. Declaring secondary containment inoperable as a result of not meeting TS SR 3.6.4.1.1 is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material and mitigate the consequences of an accident. The NRC Senior Resident Inspector has been informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At the time the licensee notified the NRC Headquarters Operations Officer, the cause of the hinged access duct being open had not been determined. This event has been added to the licensee's corrective action program. | Secondary containment Reactor Building Ventilation |
| ENS 56797 | 16 October 2023 02:56:00 | Brunswick | Unusual Event Due to Fire Not Verified to Be Extinguished within 15 Minutes | The following information was provided by the licensee: At 2256 EDT on October 15, 2023, Brunswick declared a Notification of Unusual Event due to a fire not extinguished within 15 minutes. The licensee received fire alarms and indication of a halon discharge in the basement of the emergency diesel generator building. Due to the delay in the entry into the area, the licensee was not able to verify that the fire was out within 15 minutes. Upon entry into the room, the licensee noted an acrid odor near a transformer, but there was not a fire in the room. The fire was declared out at 2310 EDT. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email).
The following information was provided by the licensee via email: Termination of Unusual Event due to verification of no fire in the basement of the emergency diesel generator building." The licensee terminated the Unusual Event at 0045 on 10/16/23. The licensee notified the NRC Resident Inspector. Notified R2DO (Miller), IR-MOC (Grant), NRR-EO (Felts), DHS-SWO, FEMA Ops Center, CISA Central, FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). | Emergency Diesel Generator |
| ENS 56786 | 10 October 2023 20:53:00 | Cooper | Offsite Notification - Spurious Alarm Actuation | The following information was provided by the licensee via fax: On October 10, 2023, at 1553 CDT, Cooper Nuclear Station (CNS) was notified of a spurious actuation of a single alert notification system siren in Nemaha, Nebraska. The CNS Emergency Alert System (EAS) was not activated. The actuation occurred during siren testing conducted at approximately 1545 CDT. No emergency conditions are present at Cooper Nuclear Station. A press release from Nebraska Public Power District is not planned at this time. This condition is reportable under 10CFR 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 NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Offsite notification was to local Nemaha County Emergency Management. | |
| ENS 56759 | 27 September 2023 15:41:00 | Monticello | Actuation of Reactor Protection and Containment Isolation Systems | The following information was provided by the licensee via fax: (On 09/27/2023) at 1041 CDT, with the plant at 75 percent power and main turbine control valve testing in progress, a reactor pressure transient resulted in a reactor steam dome high pressure scram and subsequent group 1 primary containment isolation of the main steam lines (MSL). All main steam isolation valves closed as a result of the group 1 isolation signal. Additionally, a group 2 containment isolation signal was received due to reactor pressure vessel (RPV) level less than plus 9 inches during the transient. Operations personnel responded and stabilized the plant. The high-pressure coolant injection (HPCI) system was placed in service to control RPV pressure. HPCI did not inject into the RPV and was not needed to control RPV water level. The cause of the initial pressure transient is under investigation. The NRC Resident Inspector has been notified.
The utility notified the State of Minnesota and Wright and Sherburne counties. Notified R3DO (Orlikowski) | Main Steam Isolation Valve Primary containment Reactor Pressure Vessel Main Steam Line |
| ENS 56753 | 20 September 2023 19:15:00 | Peach Bottom | FITNESS-FOR-DUTY Report | The following information was provided by the licensee via phone and email: A licensed operator had a confirmed positive test for alcohol during another entity's pre-access fitness-for-duty screening for unescorted access authorization. The individual's unescorted access at Peach Bottom Atomic Power Station has been denied. The NRC Resident Inspector has been notified. | |
| ENS 56710 | 2 September 2023 10:32:00 | Nine Mile Point | Automatic Reactor Scram Due to Feedwater Transient | The following information was provided by the licensee via email: On 9/2/2023 at 0632 EDT, a feedwater transient occurred resulting in an reactor protection system (RPS) automatic reactor scram on low level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 recirculation sample system isolation, Group 3 traveling in-core probe (TIP) isolation valve isolation, Group 6 and 7 reactor water cleanup isolation, and Group 9 containment purge isolations. All control rods inserted as expected. High pressure core spray and reactor core isolation cooling initiated and injected as expected. ECCS systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable and in Mode 3. These 4 hour and 8 hour non-emergency reports are 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). The NRC Resident was informed. There was no impact on Unit 1. | Feedwater Reactor Protection System Reactor Core Isolation Cooling High Pressure Core Spray Reactor Water Cleanup Control Rod |
| ENS 56707 | 1 September 2023 14:00:00 | Monticello | Controlled Substance Found in the Protected Area | The following information was provided by the licensee via email: On 9/1/23 at 0900 hours, site personnel identified a 1 ounce bottle of vanilla extract in the protected area. Alcohol was identified as an ingredient on the ingredients label. It was determined that the alcohol by volume (ABV) for vanilla extract was 35 percent, above the 0.5 percent ABV considered low alcohol content. | |
| ENS 56685 | 20 August 2023 20:00:00 | Fermi | Technical Specification Required Shutdown | The following information was provided by the licensee via email: On 8/20/2023 at 1600 EDT, during plant walkdowns in the drywell while in mode 3 to identify a cause of increasing unidentified leakage rate, reactor coolant system pressure boundary leakage (approximately 2 gpm) was identified on the reactor recirculation sample line between the reactor recirculation sample line inboard isolation valve (B3100F019) and where the sample line taps off the B reactor recirculation jet pump riser. This requires entry into technical specification 3.4.4 condition C, identification of pressure boundary leakage with a required action to be in mode 3 in 12 hours and mode 4 in 36 hours. At 1630 EDT, a technical specification required shutdown to mode 4, cold shutdown, was initiated. A press release by DTE is anticipated. This event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i), a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi), and an eight-hour, non-emergency notification 10 CFR 50.72(b)(3)(ii)(A) for the degraded condition of the pressure boundary. Investigation into the cause of the reactor coolant system pressure boundary leakage is still ongoing. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. | Reactor Coolant System |
| ENS 56676 | 11 August 2023 08:29:00 | Quad Cities | Automatic Reactor Scram | The following information was provided by the licensee via phone and email: At 0329 (CDT) on August 11, 2023, with Unit 2 in Mode 1 at 90 percent power, the reactor automatically tripped due to a turbine trip. The trip was uncomplicated with all systems responding normally post-trip. The cause and details of the event are under investigation. Containment isolation valves actuated closed in multiple systems on a valid Group II signal. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B), and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group II isolation. Operations responded using the emergency operating procedure and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Protection System Main Condenser |
| ENS 56667 | 8 August 2023 04:00:00 | FitzPatrick | Fitness for Duty Test | The following information was provided by the licensee via email: A licensed (non-active) individual failed to comply with fitness for duty testing policies. The individual's unescorted access was terminated. | |
| ENS 56652 | 3 August 2023 14:03:00 | Nine Mile Point | Failed Fitness for Duty Test | The following information was provided by the licensee via email: A licensed employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Senior Resident Inspector was notified. | |
| ENS 56650 | 1 August 2023 13:55:00 | Fermi | Active Seismic Monitoring System Inoperable | The following information was provided by the licensee via email: On 08/01/2023 at 0955 EDT, the Fermi 2 active seismic monitoring system provided indication of a potential seismic activity event. Plant abnormal procedures were entered, and compensatory measure were met and remain in place. Neither the (United States Geological Survey) (USGS) nor the next closest nuclear power plant could confirm or validate the readings obtained at Fermi. The seismic monitoring system was declared nonfunctional to validate the calibration of the system. Femi 2 has two active seismic monitors: one on the reactor pressure vessel pedestal and one in the high-pressure core injection (HPCI) room. Only the HPCI room accelerometer was declared inoperable. The HPCI accelerometer is the sole 'trigger' for the seismic recording system, which outputs peak accelerations experienced during a seismic event. This is used in assessment of the magnitude of an earthquake for EAL HU 2.1. The loss of the active seismic monitoring system is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). No seismic activity has been felt onsite and the USGS recorded no seismic activity in the area. The NRC Resident Inspector has been notified. | Reactor Pressure Vessel |
| ENS 56606 | 3 July 2023 20:30:00 | Quad Cities | Offsite Notification | The following information was provided by the licensee via phone and email: At 1530 (EDT) on July 3, 2023, Constellation Generation Company, LLC reported to the Illinois Environmental Protection Agency and the Illinois Emergency Management Agency and Office of Homeland Security (IEMA-OHS) that tritium concentrations in existing monitoring wells in a known (10 CFR) 50.75(g) recovery area were found higher than normal (at Quad Cities Nuclear Generating Station). This higher than normal tritium concentration did not exceed any NRC regulations or reporting criteria and there is no indication of a liquid release beyond the site boundary. This notification is being made solely as a four-hour, non-emergency notification for a notification of other government agency in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56570 | 13 June 2023 03:33:00 | Fermi | Accident Mitigation - High Pressure Coolant Injection Declared Inoperable | The following information was provided by the licensee via email: At 2333 EDT on June 12, 2023, the division 2 Mechanical Draft Cooling Tower (MDCT) Fan `D' was declared inoperable due to a trip of the fan while running in high speed. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the division 2 Emergency Equipment Cooling Water (EECW) system. The EECW system cools various safety related components, including the High-Pressure Coolant Injection (HPCI) system room cooler. An unplanned HPCI inoperability occurred based on a loss of the HPCI room cooler. The cause of MDCT Fan `D' trip is currently unknown with trouble shooting being developed for remediation of the condition. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. The NRC Senior Resident Inspector has been notified.
The purpose of this notification is to retract a previous event notification (EN) 56570 reported on June 13, 2023, at 0602 EDT. The cause of the fan trip was a failed vibration switch. At 0429 EDT on June 14, 2023, the vibration switch was replaced, the MDCT fan "D" was tested satisfactory for operability, and the UHS, emergency diesel generator 13/14, and MDCT were declared operable. Following the initial EN, further analysis of the condition was performed utilizing a previously performed gothic analysis model (to perform HPCI room heat-up calculations) which bounded this condition. Based on the initial conditions at the time of the indication loss, specifically HPCI room and suppression pool temperature, it was determined that the resulting worst case post-accident room temperature was sufficiently low enough to provide margin to HPCI operability without the room cooler in service for the required mission time. No other concerns were noted during the event. HPCI remained operable and there was no loss of safety function. The fan trip did not involve a 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 under 10 CFR 50.72(b)(3)(v)(D). Therefore, the NRC non-emergency 10CFR50.72(b)(3)(v)(D) report was not required and the NRC report 56570 can be retracted, and no licensee event report under 10 CFR 50.73(a)(2)(v)(D) is required to be submitted. The licensee notified the NRC Resident Inspector. Notified R3DO (Nguyen) | High Pressure Coolant Injection Emergency Diesel Generator Emergency Equipment Cooling Water |
| ENS 56542 | 25 May 2023 04:00:00 | Fermi | Part 21 Interim Evaluation - Mechanical Draft Cooling Tower Fan Brakes Design Flaw | The following is a summary of the information provided by the licensee via email: As previously reported under Fermi LER 2023-001-00, submitted on May 22, 2023, at 1145 EDT on March 23, 2023, it was determined that all mechanical draft cooling tower (MDCT) fan brakes would not perform their design function during a tornado due to the speed switch not functioning over its published voltage and frequency ranges. The MDCT fan brakes are required to prevent fan overspeed from a design basis tornado. On May 25, 2023, Fermi completed its 10 CFR Part 21 discovery process and determined the need to perform a 10 CFR Part 21 evaluation. The vendor (Engine Systems Inc. (ESI)) was contacted and the purchaser (Fermi) assumed responsibility for performing the Part 21 evaluation for the supplied mechanism. This Part 21 evaluation is being tracked by Fermi CARD 23-20075. It has been determined the direct cause of the event was due to the Dynalco speed switch model SST-2400A-1, supplied by ESI, not functioning over its published voltage and frequency ranges. Corrective actions were taken to develop a design change to correct MDCT fan speed control system returning the MDCT fans, ultimate heat sink, and the service water subsystems to service on March 24, 2023. The root cause evaluation is ongoing, and written follow-up will be provided in 30 days by providing a supplement to the original LER by June 24, 2023. No new commitments are being made in this submittal. | Service water |
| ENS 56535 | 22 May 2023 23:38:00 | Monticello | Notification to Another Government Agency | The following information was provided by the licensee via phone and fax: On 5/22/23, Xcel Energy performed a notification to the state of Minnesota Duty Officer, in accordance with Minnesota Statute 115.061, regarding 300-600 gallons of pumped ground water that overflowed from a holding tank and returned to the ground area from which it was pumped. The groundwater being pumped is related to recovery activities associated with the event reported on November 22, 2022 (EN 56236). This notification is being made solely as a four-hour, non-emergency report for notification to other government agency. An update is being provided to the Monticello community and published on Xcel Energy's website. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56531 | 20 May 2023 08:15:00 | Browns Ferry | Automatic Reactor Scram | The following information was provided by the licensee via phone and email: On 5/20/2023 at 0315 CDT, Browns Ferry Unit 1 was at 80 percent reactor power performing, 'Turbine control valve fast closure turbine trip and RPT (recirculation pump trip) initiate logic testing'. During performance of this test, Unit 1 received a full reactor scram. An investigation is in progress to determine the cause of the scram. All systems responded as expected, and Unit 1 is stable at zero percent power in mode 3. All control rods fully inserted into the core. Main steam isolation valves remained open with main turbine bypass valves controlling pressure. Reactor feedwater pumps remained in service to control reactor water level. Primary containment isolation signals groups 2, 3, 6, and 8 were received with expected system actuations. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. The event is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. The NRC Resident has been notified. | Feedwater Reactor Protection System Main Steam Isolation Valve Primary containment Control Rod |
| ENS 56513 | 9 May 2023 18:55:00 | Peach Bottom | Loss of Emergency Preparedness Capabilities | The following information was provided by the licensee via phone and email: At 1455 (EST) on Tuesday May 9, 2023, Peach Bottom Atomic Power Station (PBAPS) technical support center (TSC) ventilation system lost power. Power loss was caused by a tree down on the 361 transmission line. Power was not able to be restored within an hour. At 1639 (EST), power was restored to TSC ventilation, and capability was restored. This report is being submitted pursuant to 10 CFR 50.72(b)(3)(xiii) as a major loss of emergency preparedness capabilities due to a reduction in the effectiveness of the onsite TSC. NRC Resident has been notified. | |
| ENS 56509 | 8 May 2023 07:07:00 | Cooper | Shutdown Required by Technical Specifications Due to Inoperable Core Spray | The following information was provided by the licensee via fax: At time 0207 CDT, Cooper Nuclear Station (CNS) entered Technical Specification (Limiting Condition for Operation) LCO 3.0.3 due to declaring core spray subsystems A and B inoperable. This declaration was based on an issue with relays installed from the same manufacturing batch. The ability of the relays to function correctly to annunciate loss of logic power was called into question and they were declared inoperable. The plant has initiated actions to repair/replace affected relays. This event is reportable under 10 CFR 50.72(b)(2)(i) as an initiation of any nuclear plant shutdown required by Technical Specifications. In addition, this event Is reportable under 10 CFR 50.72(b)(3)(v) as a condition that could have prevented the fulfillment of a safety function for the core spray systems. NRC Resident Inspector was notified.
The following information was provided by the licensee via email: Technical Specification LCO 3.0.3 was exited at 0805 CDT on May 8, 2023. A reasonable expectation of operability was developed for the core spray subsystems A and B. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Shutdown was initiated and power was reduced approximately 45 percent. Reactor power is currently at 55 percent at the time of notification. Notified R4DO (Werner) via email.
The following information was provided by the licensee via email: CNS is retracting the 8-hour 10 CFR 50.72(b)(3)(v) non-emergency notification, for a condition that could have prevented the fulfillment of a safety function, made on May 8, 2023, at 0207 CDT (EN# 56509). Subsequent evaluation concluded that the core spray subsystems remained operable in accordance with the Technical Specifications Requirements 3.5.1, ECCS - Operating. As a result of the core spray system remaining operable, no loss of safety function occurred. The NRC Senior Resident Inspector has been notified. Notified R4DO (Werner). | Core Spray |
| ENS 56505 | 5 May 2023 01:34:00 | Browns Ferry | Fitness for Duty Report | The following information was provided by the licensee via email: On 05/04/2023 at 2034 CDT, a Browns Ferry Nuclear Plant non-licensed employee supervisor had a confirmed positive drug test identified during random fitness-for-duty medical testing. Employee's unescorted access has been suspended. A review of the employee's work has been completed. The (NRC) Resident Inspector has been notified. |