Semantic search
| Entered date | Site | Region | Reactor type | Event description | Topic | |
|---|---|---|---|---|---|---|
| ENS 57438 | 23 November 2024 02:42:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone or email: At 1937 EST on 11/22/2024, it was discovered that both trains of the control room emergency air temperature control system (CREATCS) were simultaneously inoperable; 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 has been notified. The following additional information was obtained from the licensee in accordance with headquarters operations officers report guidance: Technical specification 3.7.11 conditions A and C were entered as a result of this event. The 'B' train of CREATCS was restored at 0130 EST on 11/23/24 and the plant exited condition C. The 'A' train remained out of service at the time of notification. Although CREATCS is a common system for both Units 1 and 2, Unit 1 was defueled and outside the mode of applicability during the timeframe of this event. | |
| ENS 57351 | 29 September 2024 08:40:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 05:56 EDT on 09/29/2024, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip. The motor driven auxiliary feedwater (MDAFW) level control valves (LCV) for loop 1 failed to respond from the main control room. All others systems responded normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the auxiliary feedwater (AFW) and steam dump systems. Unit 2 is currently stable in Mode 6 for a maintenance outage and was not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour, non-emergency notification per 10 CFR 50.72 (b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. The AFW system started automatically and is operating as designed with the exception of the MDAFW LCVs for loop 1. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 57298 | 30 August 2024 18:30:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via phone and email: At 1051 CDT on 8/30/2024, during transfer of 4KV shutdown bus 1 to support Unit 1 shutdown activities, the alternate feeder breaker failed to close resulting in 4KV shutdown boards 'A' and 'B' experiencing an under voltage condition. This resulted in 'A' and 'B' diesel generators automatically starting and tying to their respective boards. This condition also caused a loss of reactor protection system (RPS) channel 'A' on Units 1 and 2, resulting in invalid actuation of primary containment isolation system Groups 2, 3, 6, and 8. The failure of the board to transfer was identified during preparation for the evolution, contingency actions were prepared and implemented as planned. The breaker failure to close has been corrected and 4KV shutdown bus 1 is energized on alternate. 4KV shutdown boards 'A' and 'B' have been restored to offsite power supplies and the diesel generators are secured. 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 has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The change in reactor power from 70 percent to 40 percent was not as a result of the failed breaker, rather Browns Ferry Unit 1's change in reactor power was due to a scheduled reactor shutdown which was in progress. In regards to the Unit 2 loss of channel 'A' RPS, this was not a specified system actuation. The actuation of the 'A' and 'B' diesel generators were the specified system actuation. Although the 'A' and 'B' diesels are common to both Units 1 and 2, only Unit 1 credits these specific diesel generators for accident mitigation. As such, this event is only reportable from Unit 1. Unit 2 did not experience a specified system actuation. | |
| ENS 57285 | 23 August 2024 13:44:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 1219 EDT, with unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the auxiliary feedwater (AFW) and steam dump systems. Unit 2 is currently in a refueling outage (U2R26) and was not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non- emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. The AFW system started automatically and is operating as designed. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 57313 | 10 September 2024 08:28:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: This 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A). The event was an invalid actuation of the Unit 1 containment ventilation isolation (CVI) system. On August 3, 2024, at 0840 EDT, with Unit 1 at 100 percent power, train 'A' of the CVI system actuated due to an invalid signal from 1-RM-90-130, containment purge air exhaust monitor. 1-RM-90-130 was out of service for maintenance testing at the time of the invalid signal. The cause of the signal was determined to be the result of an installed multimeter timing out, creating a short in the actuation circuitry. The train 'A' CVI signal was a full actuation of that train and the system functioned as designed. Prior to and following the CVI alarm, all other radiation monitors were stable at their normal values; therefore, the CVI (actuation) was invalid. Control room operators performed appropriate checks and confirmed that all required automatic actuations occurred as designed. Subsequent completion of the maintenance instruction was successful. This event was entered into the corrective action program as CR 1948103. The NRC Resident Inspector was notified. | |
| ENS 57336 | 23 September 2024 21:46:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 2127 EDT, on 8/01/2024, with Unit 1 in mode 1 at 98 percent power, a complete actuation of the 'A' train containment ventilation isolation (CVI) occurred. The 'A' train CVI resulted from the failure of a radiation monitor providing input to the isolation circuitry. The CVI removes containment purge from operation should it be in service and secures other radiation monitors which measure reactor coolant system leakage. In accordance with the station's procedures and technical specifications, a restoration from the CVI was made. Troubleshooting revealed that replacement of this obsolete radiation monitor was justified; a design change to perform this replacement is in progress. This report is being made under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an invalid actuation of the 'A' train CVI. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified of the event. | |
| ENS 57253 | 30 July 2024 18:52:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1641 EDT, with Unit 2 in Mode 1 at 94 percent power and increasing in power after a forced outage, the reactor automatically tripped due to an electrical trouble turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the auxiliary feedwater (AFW) and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. | |
| ENS 57245 | 25 July 2024 20:48:00 | Browns Ferry | NRC Region 2 | The following information was provided by the licensee via phone and email: Tennessee Valley Authority (TVA) completed an engineering evaluation for a Fike Metal Products 16-inch rupture disc, part number 16-CPV-C, which had failed in March 2024 during an event previously reported to the NRC as Event Notification 57036 and Licensee Event Report 260/2024-002-00. The evaluation determined that the failure of the rupture disc constituted a failure to comply by a basic component which resulted in a substantial safety hazard. The rupture disc was procured as a non-safety related item from Fike Corporation and commercially dedicated by Paragon Energy Solutions. The disc was supplied to TVA in a satisfactory condition meeting all acceptance criteria. During a routine flowrate surveillance test, the high-pressure coolant injection (HPCI) inner rupture disc developed a hole which caused the Unit 2 HPCI turbine to trip. This resulted in (Browns Ferry Unit 2) entering Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.1 Condition `C', which is a 14-day shutdown LCO. Per HPCI system design criteria, turbine casing protection disc rupture pressure shall be at 175 psig plus 1 or minus 10 psig and the rupture discs shall be sized for a flow capacity of 600,000 pounds per hour at 200 psig, minimum. The failed HPCI inner rupture disc did not experience pressures above 45 psig since being installed; therefore, the HPCI turbine inner rupture disc did not meet its technical requirements. On July 23, 2024, the Browns Ferry Nuclear Plant Site Vice President was notified of the requirement to report this event under 10 CFR 21.21. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i). A written notification in accordance with 10 CFR 21.21(d)(3)(ii) will be provided within 30 days. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Both Fike Corporation and Paragon Energy Solutions have been informed of the HPCI inner rupture disc not meeting technical requirements. Known potentially affected plants include Browns Ferry Units 1, 2, and 3.
The following information was provided by the licensee via phone and email: The purpose of this notification is to retract a previous event notification, EN 57245, reported on 7/25/24. Continued evaluation has concluded that the failure of the disc was not the result of a failure to comply by a basic component, therefore, the NRC non-emergency 10 CFR 21.21 (d) report was not required and the NRC EN 57245 can be retracted. The licensee has notified the NRC Resident Inspector. Notified R2DO (Masters) and Part 21/50.55 Reactors group (Email). | ||
| ENS 57324 | 16 September 2024 16:30:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1248 EDT on July 22, 2024, with Unit 1 in mode 1 at 100% power, a complete actuation of the 'A' train containment ventilation isolation (CVI) occurred. The 'A' train CVI resulted from the failure of a radiation monitor providing input to the isolation circuitry. This radiation monitor was subsequently repaired and a restoration from the CVI was made. The CVI removes containment purge from operation should it be in service and secures other radiation monitors which measure reactor coolant system leakage. This report is being made under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an invalid actuation of the 'A' train CVI. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified of the event. | |
| ENS 57214 | 8 July 2024 18:24:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1521 EDT on July 8, 2024, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The (reactor) trip was not complex with all systems responding normally post trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the steam dump system and the auxiliary feedwater (AFW) system. Unit 2 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the AFW system (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The specific cause of the turbine trip is under investigation by the licensee. | |
| ENS 57126 | 13 May 2024 16:40:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0917 EDT on May 13, 2024, a control room operator erroneously rendered the `B train of the Unit 2 residual heat removal (RHR) system inoperable. This occurred while the `A train of the Unit 2 RHR system was out of service for preplanned maintenance. RHR serves as the low head safety injection (LHSI) subsystem for the emergency core cooling system (ECCS) and because of this, Unit 2 was without a required train of ECCS from 0917 EDT to 0921 EDT. No other equipment issues were identified. The LHSI subsystem is credited by the analysis for a large break loss of coolant accident at full power. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D). The NRC resident inspector has been notified. There is no release of radioactive material associated with this event. | |
| ENS 57090 | 25 April 2024 02:22:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 4/24/2024 at 2215 CDT, Browns Ferry Unit 1 experienced an automatic reactor scram. The cause of the scram is currently under investigation. The main steam isolation valves (MSIVs) remain open with the main turbine bypass valves controlling reactor pressure. The reactor feedwater pumps are in service to control reactor water level. Primary containment isolation systems (PCIS) Groups 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all components actuated as required. Following the reactor scram, due to reactor water level reaching minus 45 inches, both high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) initiation signals were received, and both initiated as designed. All safety systems operated as expected. This event requires a 4-hour report per 10 CFR 50.72(b)(2)(iv)(A), `Any event that results or should have resulted in emergency core cooling system (ECCS) discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This event requires a 4-hour report 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. This event requires an 8-hour report per 10 CFR 50.72(b)(3)(iv)(A), `Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B): 1) Reactor protection system (RPS) including: reactor scram or reactor trip. 2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). 4) ECCS for boiling water reactors (BWRs) - high-pressure coolant injection (HPCI). 5) BWR reactor core isolation cooling system (RCIC). All safety systems operated as expected. At no time was public health and safety at risk. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Units 2 and 3 were not affected. | |
| ENS 57070 | 10 April 2024 11:23:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email and phone call: A non-licensed employee supervisor had a confirmed positive during a random fitness-for-duty test. The employees access to the plant has been terminated. The NRC Senior Resident Inspector has been notified. | Fitness for Duty |
| ENS 57036 | 19 March 2024 18:19:00 | Browns Ferry | NRC Region 2 | GE-4 | 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. | |
| ENS 57006 | 5 March 2024 04:14:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0132 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main feedwater isolation signal which resulted in steam generator lo-level reactor trip. The reactor trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the main feedwater isolation is being investigated. | |
| ENS 56990 | 24 February 2024 09:27:00 | Browns Ferry | NRC Region 2 | GE-4 | 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. | |
| ENS 56970 | 16 February 2024 02:05:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 2224 EST on February 15, 2024, with both units 1 and 2 in mode 1 at 100 percent power, an actuation of the emergency diesel generator (EDG) system on 1A-A, 1B-B, and 2B-B EDGs occurred while removing clearances. The 2A-A EDG did not start because it was still under a clearance. The reason for the emergency diesel generator system auto-start was clearance removal sequencing errors. The emergency diesel generator system automatically started as designed when the common emergency start signal was received. 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 emergency diesel generator system. 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: In accordance with NUREG-1022, Section 2.8 and Section 4.2.3, Watts Barr is retracting the previous report EN 56970 pursuant to 10 CFR 50.72(b)(3)(iv)(A). The start signal for the 1A-A, 1B-B, and 2B-B emergency diesel generators (EDG)s was from activation of the common emergency start of the 2A-A EDG. The actuation was not from a loss of offsite power (LOOP) to any shutdown board or from any parameters that would initiate a safety injection (SI) signal, for which the EDG is designed to provide a design basis safety function. Also, the starts were not from intentional manual actuation. Starting the EDGs did not make them inoperable and each EDG was able to perform its design safety function. The common emergency start relay for each diesel is not safety related. It is an anticipatory and redundant circuit to start other EDGs in the event of a LOOP or SI related to the specific EDG. With the 2A-A EDG out of service, the associated common emergency circuit would not be required to perform any function. The starts were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the system. Since the starts were not initiated via an automatic signal from a LOOP, SI, or traditional operator action, the signal is not a valid actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). Therefore, EN 56970 is being retracted. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Miller) | |
| ENS 57077 | 15 April 2024 14:38:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 2224 EST on February 15, 2024, with both units 1 and 2 in Mode 1 at 100 percent power, an invalid start of the emergency diesel generator (EDG) system on 1A-A, 1B-B, and 2B-B EDGs occurred while removing clearances. The 2A-A EDG did not start because it was still under a clearance. The 1A-A, 1B-B, and 2B-B EDGs started and functioned successfully. The start signal for the 1A-A, 1B-B, and 2B-B EDGs was generated from the common emergency start of the 2A-A EDG. The signal was not from a loss of offsite power (LOOP) to any shutdown board or from any parameters that would initiate a safety injection (SI) signal, for which the EDG is designed to provide a design basis safety function. Also, the starts were not from intentional manual actuation. Starting the EDGs did not make them inoperable and each EDG was able to perform its design (basis) safety function. The common emergency start relay for each diesel is not safety related. It is an anticipatory and redundant circuit to start other EDGs in the event of a LOOP or SI related to the specific EDG. With the 2A-A EDG out of service, the associated common emergency circuit would not be required to perform any function. The starts were not initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the system. This event was originally reported under EN 56970 on February 16, 2024, at 0205 EST in accordance with 10 CFR 50.72(b)(3) (iv)(A) as an event that results in a valid actuation of the emergency diesel generator system. This EN was retracted on February 21, 2024, at 1549 EST. This event is being reported in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) as an event that results in an invalid actuation of the emergency diesel generator system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 56963 | 13 February 2024 17:10:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a synopsis of information provided by the licensee via email: On February 13, 2024, a non-licensed supervisor violated the station's FFD policy. The employee's access at Sequoyah Nuclear Plant has been terminated. The NRC resident inspector has been notified. | Fitness for Duty |
| ENS 56941 | 1 February 2024 15:32:00 | Browns Ferry | NRC Region 2 | GE-4 | 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 56935 | 27 January 2024 23:39:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 2141 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. The licensee notified the NRC Resident Inspector. | |
| ENS 56910 | 28 December 2023 18:55:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: Plant alignment caused an unanalyzed condition regarding unit 1 and unit 2 Appendix R procedures. (Watts Bar Nuclear) (WBN) unit 1 and unit 2 Appendix R procedures require manual operator action times including (volume control tank) (VCT) isolation. They are calculated with an assumed hydrogen cover gas constant at 20 psig. This is to preclude hydrogen ingestion into the charging pumps with an operator action time of 70 minutes. Due to recent lower hydrogen concentration in the (reactor coolant system) (RCS), (unit 2) VCT hydrogen regulator set point was increased to 28 psig. This increased pressure set point invalidated the initial assumptions made in the Appendix R calculations for manual operator action times. WBN unit 1 VCT hydrogen regulator was also verified high out of band at 22 psig. WBN has restored unit 1 and unit 2 VCT hydrogen regulators to the required specification. The NRC Resident Inspector has been notified of this condition. | Unanalyzed Condition Operator Manual Action |
| ENS 56888 | 13 December 2023 09:45:00 | Sequoyah | NRC Region 2 | The following information is a synopsis of information provided by the licensee via phone and email: On December 11, 2023, Tennessee Valley Authority Sequoyah Nuclear Plant completed an internal 10 CFR Part 21 evaluation concerning Siemens 6.9kV, 1200A vacuum circuit breakers, Model No. 7-HKR-50-1200-130. Three separate breakers were found with issues including loose wires terminated incorrectly and the mechanism-operated control switch clevis pin missing a cotter key. Additionally, the mastic insulating pads were found defective on all three lower primaries by way of separation. The affected breakers were never installed in a safety related application. The NRC Resident Inspector will be notified. A written notification will be provided within 30 days. The manufacturer, Siemens, was notified of the defects. The only plant known to be affected at the time of the report is the Sequoyah Nuclear Plant. | ||
| ENS 56850 | 12 November 2023 22:02:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On November 12, 2023, at 0300 EST, a Watts Bar contractor was transported offsite for medical treatment due to a work-related injury. Upon arrival at an offsite medical facility, medical personnel determined the injury required the individual to be admitted into the hospital and will be kept overnight. The individual was inside of the Radiological Controlled Area, however was free released with no contamination. The injury and hospitalization were reported to the Occupational Safety and Health Administration (OSHA) under 29 CFR 1904.39(a)(2). The contracting agency informed OSHA at 1319 EST. Watt Bar Operations personnel were officially notified by the contracting agency of the report made to OSHA at 1945 EST. This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. | |
| ENS 56845 | 9 November 2023 15:55:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a summary of information provided by the licensee via email: A controlled substance was found in the protected area. The NRC Resident Inspector has been notified. | |
| ENS 56809 | 21 October 2023 09:25:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax and email: Fire potentially degrading the level of safety of the plant. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At 0907 EST, the licensee declared a notification of unusual event, under emergency action level HU.4, due to multiple fire alarms and CO2 discharge in the emergency diesel building. When the plant fire brigade entered the building, there was no indication of fire or damage to any plant equipment. The cause of the multiple alarms is under investigation. State and local authorities were notified and no offsite assistance was requested. Both units remain at 100 percent power. The NRC Resident Inspector has been notified. 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).
At 1007 EDT, Watts Bar terminated the notification of unusual event. The basis for termination was that no fire or damaged plant equipment was found. The NRC Resident Inspector has been notified. Notified R2DO (Miller), IR-MOC (Crouch), 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).
Watts Bar Nuclear Plant (WBN) is retracting Event Notice 56809, Notice of Unusual Event, based on the following additional information, not available at the time of the initial notification. Specifically, in accordance with the emergency preparedness implementing procedures, WBN reported a condition that was determined to meet emergency action level (EAL) HU4, Initiating criteria number 1, receipt of multiple (more than 1) fire alarms or indicators and the fire was within any Table H2 plant area, which includes the diesel generator building. It was further determined that multiple fire detection zones actuated (spurious and invalid) enabling the discharge of installed fire suppression (CO2) into the space. Upon entry by the site fire brigade, it was determined that no smoke or fire existed and reported to the Shift Manager at 0930 EDT. All fire alarms were reset. Troubleshooting activities are in progress to determine the cause. A fire watch has been established and CO2 has been isolated. The required compensatory measures for the affected areas will remain in place until completion of the investigation, and CO2 suppression is restored to functional. Notified R2DO (Miller), IR-MOC (Crouch), NRR-EO (Felts), DHS-SWO (email), FEMA Ops Center (email), CISA Central (email), FEMA NWC (email), CWMD Watch Desk (email), DHS NRCC THD Desk (email), and DHS Nuclear SSA (email). | Fire Watch |
| ENS 56660 | 4 August 2023 20:51:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1746 EDT on 08/04/2023, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to number 2 steam generator low low level. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the number 2 steam generator low low level is being investigated. | |
| ENS 56593 | 27 June 2023 19:04:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1626 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The (reactor) trip was not complex with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. | |
| ENS 56592 | 27 June 2023 11:52:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0831 (EDT) on June 27, 2023, Sequoyah Nuclear Plant reported an oil discharge into the plant intake located on the Tennessee River to the (Department of Transportation) National Response Center (report number 1371356). The source of oil was from a broken hydraulic hose from equipment in use on the intake. This oil spill is minor and did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable 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 56541 | 25 May 2023 17:02:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1345 EDT on May 25, 2023, it was determined that a fire barrier for area 737-A1B was not installed, and would render the 2A Emergency Diesel Generator (EDG) not operable in the event of a fire on the Unit 2 side of elevation 737 in the Auxiliary Building. The 2A EDG is the credited power source for fire safe shutdown for a fire located in this area. Without the credited source of power, this places WBN U2 (Watts Bar Nuclear Unit 2) in an unanalyzed condition. A fire watch has been established in the area until the issue is resolved. Therefore, this event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. | Safe Shutdown Unanalyzed Condition Fire Barrier Fire Watch |
| ENS 56531 | 20 May 2023 07:45:00 | Browns Ferry | NRC Region 2 | GE-4 | 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. | |
| ENS 56505 | 5 May 2023 16:00:00 | Browns Ferry | NRC Region 2 | GE-4 | 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. | Fitness for Duty |
| ENS 56411 | 15 March 2023 04:27:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 2257 (CDT) on 3/14/2023 during the 2R22 refueling outage on Browns Ferry Nuclear Plant Unit 2, it was determined there was RCS boundary leakage from five of eight sensing lines that pass through containment penetrations X-30 and X-34 that did not meet the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code. The condition will be resolved prior to plant startup. 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.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous Event Notification, EN 56411 reported on 3/14/23. Following the initial notification, further analysis of the condition was performed. It was determined that the leaking pipe weld was ASME Section XI Code Class 2 piping which falls under the requirements of ASME Section XI Subsection IWC and not Subsection IWB. Therefore, this condition does not represent a serious degradation of the nuclear power plant, including its principle safety barriers. Based upon the above, the leaks identified on the ASME Section XI Code Class 2 equivalent Main Steam sense lines are not reportable under 10 CFR 50.72(b)(3)(ii). Therefore, the NRC non-emergency 10 CFR 50.72(b)(3)(ii) report was not required and the NRC report 56411 can be retracted and no Licensee Event Report under 10 CFR 50.73(a)(2)(ii) is required to be submitted. Notified R2DO (Miller) | |
| ENS 56385 | 2 March 2023 17:52:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 1312 CST on March 2, 2023, Browns Ferry Nuclear Plant Units 1, 2, and 3 initiated voluntary communication to the state of Alabama and local officials as part of the Nuclear Energy Institute (NEI) Groundwater Protection Initiative (GPI), after receiving analysis results for leakage from a demineralized water storage tank that contained activity above the GPI voluntary communication threshold. All these results are significantly less than the limits established by the Nuclear Regulatory Commission (NRC) and Environmental Protection Agency (EPA) for effluents from the station. Further samples obtained of the water prior to entering the Tennessee River were less than detectable. The leakage source has been isolated and additional corrective actions are in progress. This condition did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a notification of other government agency. This event is reportable 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 56371 | 18 February 2023 11:25:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On February 17, 2023 during the planned U2R22 outage on Browns Ferry Nuclear Plant Unit 2, personnel entered the Unit 2 drywell for leak identification. Personnel discovered a cracked weld on the 2A recirculation pump discharge isolation valve drain line. At 0439 CST on February 18, 2023, following engineering evaluation, this drain line was determined to be ASME Code Class 1 piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. | |
| ENS 56321 | 24 January 2023 08:43:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: At 0121 CST on 01/24/2023, it was discovered that the Unit 1 High Pressure Coolant Injection System (HPCI) was inoperable; therefore, the condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), as an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. 1-FCV-073-0006B, HPCI Steam Line Condensate Outboard Drain Valve, failed closed during normal plant configuration. This valve is normally open. The HPCI steam line is not being drained with the valve in the current position. The Unit 1 Nuclear Unit Senior Operator entered Unit 1 Technical Specifications LCO 3.5.1 Condition C with required actions C.1 to immediately verify by administrative means that the Reactor Core Isolation Cooling (RCIC) system is operable and C.2 to restore HPCI to operable status in 14 days. RCIC has been verified operable by administrative means. There was no impact to the safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Time of Discovery |
| ENS 56257 | 3 December 2022 13:05:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 12/2/2022 at 2330 (CST) during the planned F311 outage on Browns Ferry Nuclear Plant Unit 3, personnel entered the Unit 3 drywell for leak identification. Personnel discovered a through-wall piping leak on a 0.75 inch test line between the two test line isolation valves. This 0.75 inch test line is located on the residual heat removal (RHR) loop 1 shutdown cooling and RHR return line to the reactor vessel. On 12/3/2022 at 1000 CST, Engineering determined this location is classified as ASME Code Class 1 piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. | |
| ENS 56274 | 15 December 2022 12:52:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: This 60-day telephone notification is being submitted in accordance with paragraphs 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A) to report an invalid Containment Ventilation Isolation (CVI) actuation at Watts Bar Nuclear Plant (WBN) Unit 1. On November 24, 2022, at 1621 Eastern Standard Time (EST), the Train B CVI actuated due to an invalid high radiation signal from 1-RM-90-131, Containment Purge Air Exhaust Monitor. Upon investigation, the high radiation signal was caused by a failed power supply. Corrective action included replacing the power supply, 1-RM-90-131 ratemeter, and restoring the system to service. Prior to and following the invalid high radiation alarm, all radiation monitors except 1-RM-90-131 were stable at their normal values; therefore, the CVI was invalid. Control room operators performed appropriate checks and confirmed that all required automatic actuations occurred as designed. This event has been entered into the corrective action program as Condition Report 1819098. The NRC Resident Inspector was notified. | |
| ENS 56168 | 18 October 2022 22:08:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 10/18/2022 at 1440 CDT, Browns Ferry Unit 3 declared both trains of standby liquid control (SLC) inoperable due to acceptance criteria failure of 3-SI-3.1.7.6, 'Standby Liquid Control System ATWS Equivalency Calculation for Newly Established Pump Flow Rate.' The purpose of this surveillance is to ensure the anticipated transient without scram (ATWS) calculation criteria is met after each pump flow test. Chemistry performed the surveillance following pump flow testing and the requirement for equivalency calculation failed low with a result of less than 1.0. CR 1810303 documents this condition in the corrective action program. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(A), 10 CFR 50.72(b)(3)(v)(C), and 10 CFR 50.72(b)(3)(v)(D). This condition is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(v)(A),10 CFR 50.73(a)(2)(v)(C), and 10 CFR 50.73(a)(2)(v)(D). The NRC Resident Inspector has been notified of this event. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officer's report guidance: The plant entered an 8 hour limiting condition for operation based on the above. The condition was resolved at 2053 CDT when the system was restored to normal operation. | |
| ENS 55992 | 12 July 2022 17:25:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via fax or email: At 0917 CDT on 7/12/2022, during the performance of U1 (Unit 1) High Pressure Coolant Injection (HPCI) rated flow test, the 1-FCV-73-19 (HPCI governor valve) failed to operate as expected. This condition results in U1 HPCI being inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The Automatic Depressurization System (ADS) and Reactor Core Isolation Cooling (RCIC) system remain operable. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. U1 entered TS LCO 3.5.1 Condition C, 14-day Shutdown LCO (Limiting Condition for Operation), due to the HPCI inoperability. | |
| ENS 56322 | 25 January 2023 13:22:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information is a synopsis of information provided by the licensee via fax and phone: On May 23, 2022, Framatome informed Tennessee Valley Authority (TVA) of a deviation of breakers purchased under contract. On January 23, 2023, TVA determined that a defect of the basic component could create a substantial safety hazard. Framatome Inc. identified a deviation in the Siemens medium voltage vacuum circuit breaker where a failure to electrically charge or electrically close could occur. Framatome Inc. identified this as a departure from the technical requirements included in the procurement document. It is noted that the ability to electrically trip the circuit breaker would not be affected by the condition. TVA was notified by Framatome under 10 CFR 21.21(b) to evaluate the application of the breaker for a substantial safety hazard. The TVA evaluation identified these breakers as intended for use in safety related Class 1E applications where a loss of the closure function would impact mitigation of design basis accidents and transients. During the Framatome dedication testing/inspection of Siemens medium voltage vacuum breakers, a hi-pot test failure on one circuit breaker was encountered. Troubleshooting and inspection found damage to charging motor wiring. It was determined that the cause of the damage was due to the manner in which control wiring was routed and connected to the internal bracket in close proximity to a bracket edge. This edge caused damage to wiring after significant number of cycles were applied to the breaker prior to dedication testing. TVA received nine medium voltage vacuum circuit breakers at an offsite warehouse facility. While located at that facility, TVA, with assistance from Framatome, examined the affected breakers for the wire routing condition. The wiring harnesses of certain breakers were corrected. Framatome is to examine medium voltage vacuum circuit breakers that may be purchased under this contract for the wiring condition and correct as necessary before delivery. The NRC Senior Resident Inspector has been notified. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i).
The following information is a synopsis of information provided by the licensee via phone: The Sequoyah site licensing manager requested via phone call to the HOO that the model number for the basic component with the defect be listed in the Part 21 event narrative in addition to the official Part 21 report. The component discussed is a Siemens 6.9kV, 1200A, 125VDC Vacuum Circuit Breaker, Model No.: 7-HKR-50-1200-130. Notified R2DO (Miller) and the Part 21 Reactors Group (Email). | |
| ENS 55866 | 29 April 2022 00:19:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following is a summary of information provided by the licensee via telephone: On 04/28/22, at 2355 EDT, with both Sequoyah Unit 1 and 2 in Mode-1, 100 percent, a Notice of Unusual Event was declared due to receiving two seismic alarms and security feeling ground movement. Additionally, security in a tower heard an explosion. Both units remain in Mode-1, 100 percent and they are investigating the validity of the seismic alarms before proceeding with the Abnormal Operating Procedure required shutdown. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee will notify the NRC Resident Inspector. The state of Tennessee and the Tennessee Valley Authority were notified. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS NRCC THD Desk(email), and DHS Nuclear SSA (email).
The following is a summary of information provided by the licensee via telephone: On 4/29/22, at 0406 EDT, Sequoyah Unit 1 and Unit 2 terminated the Notice of Unusual Event. The Civil Engineers determined that the alarms were due to a failed seismic indicator channel. Through interviews, only one security officer felt ground movement for a couple of seconds and heard a faint rumbling sound. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee will notify the NRC Resident Inspector. The state of Tennessee and the Tennessee Valley Authority were notified. Notified R2DO (Miller), NRR EO (Miller), and IR MOC (Gott) via email. Additionally, notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS NRCC THD Desk(email), and DHS Nuclear SSA (email).
The following information was provided by the licensee via email: SQN (Sequoyah Nuclear Plant) is retracting the previous NOUE (Notice of Unusual Event) declaration made on 4/28/22 at 2355 (EDT) based on Emergency Action Level HU2 for a seismic event greater than Operating Basis Earthquake levels. Following the declaration of the NOUE, the station reviewed all available indications and determined that a seismic event had not occurred. The instrumentation failure was documented under Event Notification #55867. Notified R2DO (Miller), and IR MOC (Gott), NRR EO (Miller) via email. | Operating Basis Earthquake |
| ENS 55867 | 29 April 2022 07:04:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax: On 4/28/2022, at 2338 EDT, Sequoyah received an unexpected alarm for seismological recording initiated. At 2341 EDT, unexpected alarm 1/2 Safe Shutdown Earthquake response spectra exceeded was received. The National Earthquake Information Center was contacted to confirm there was no seismic activity, and this was also confirmed on the U.S. Geological Survey website. The alarms were determined to be invalid, and they occurred due to a failure in the seismic monitoring system. This failure results in loss of ability to assess the Emergency Action Level for Initiating Condition HU2 `Seismic event greater than Operating Basis Earthquake (OBE) levels' per procedure EPIP-1, `Emergency Plan Classification Matrix.' If an actual seismic event had occurred, HU2 could not be assessed. However, compensatory measures have been implemented and include assessing OBE criteria based on alternative criteria contained in procedure AOP-N.05, `Earthquake,' which provides conservative guidance when seismic instruments are unavailable. This is an eight-hour, non-emergency notification for an event resulting in a major loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii). There is 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 faulty detector was removed from service, so the remaining detector provides conservative detection as the only source to make-up the logic for a seismological alarm. | Safe Shutdown Earthquake Operating Basis Earthquake Earthquake |
| ENS 55818 | 2 April 2022 15:10:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via fax: At 1345 CDT, Browns Ferry declared a Notification of Unusual Event due to a fire at the 3B Reactor Feedwater Pump within the Turbine Building which was not extinguished within 15 minutes. Subsequently, the fire was extinguished at 1402 CDT. Unit 3 remains in Mode 1 at approximately 9.5 percent rated thermal power (RTP). Unit 1 and 2 remain at 100 percent RTP and unaffected. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The fire began at 1332 CDT. It is believed that the fire was in the oil system of the Feedwater Pump. The fire was extinguished by the on-site fire brigade. No off-site assistance was requested. The Unusual Event was declared under Emergency Action Level HU-4. The licensee notified the NRC Resident Inspector and required State and local government agencies. Unit 3 is currently stable. Notified DHS-SWO, FEMA Ops Center, and CISA Central Watch Officer, and FEMA NWC, DHS NRCC THD Desk, and DHS NuclearSSA via email.
The Notification of Unusual Event was exited at 1544 CDT. Notified R2DO (Miller), IR-MOC (Kennedy), NRR-EO (Felts), DHS-SWO, FEMA Ops Center, and CISA Central Watch Officer. | |
| ENS 55741 | 16 February 2022 16:42:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax or email: At 1159 EST, on 2/16/2022, the Watts Bar Nuclear, Shift Manager was notified that Tennessee Valley Authority (TVA) attempted to notify Tennessee Emergency Management Agency (TEMA) regarding routine siren testing at 0750 EST. TVA was unable to reach TEMA via telephone land line or the Emergency Communication and Notification System (ECNS). TEMA Watch Point staff were located at their back-up facility. TVA subsequently notified TEMA via cell phone that there were communication issues with the primary and backup notification methods. It was determined that the TEMA back-up facility was not able to receive incoming calls. At 0820 EST, TEMA positioned personnel at their primary facility in order to respond to notifications. This restored primary and backup means of notifying the state because the primary facility was not affected by the communication issues. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Offsite Communications Capability because it affected TVA's ability to notify the State of TN. The licensee has notified the NRC Resident Inspector. | Siren |
| ENS 55742 | 16 February 2022 17:01:00 | Sequoyah | NRC Region 2 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax or email: At 1128 EST on 2/16/2022, the SQN (Sequoyah Nuclear) Shift Manager was notified that TVA (Tennessee Valley Authority) attempted to notify Tennessee Emergency Management Agency (TEMA) regarding routine siren testing at 0750. TVA was unable to reach TEMA via telephone land line or the Emergency Communication and Notification System (ECNS). TEMA Watch Point staff were located at their back-up facility. TVA subsequently notified TEMA via cell phone that there were communication issues with the primary and backup notification methods. It was determined that the TEMA back-up facility was not able to receive incoming calls. At 0820, TEMA positioned personnel at their primary facility in order to respond to notifications. This restored primary and backup means of notifying the state because the primary facility was not affected by the communication issues. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Offsite Communications Capability because it affected TVA's ability to notify the State of TN. The licensee has notified the NRC Resident Inspector. | Siren |
| ENS 55706 | 16 January 2022 06:41:00 | Browns Ferry | NRC Region 2 | GE-4 | The following information was provided by the licensee via email: On 1/15/2022 at 2320 (EST) during the planned F108 outage on Browns Ferry Nuclear Plant Unit 1, personnel entered the Unit 1 Drywell for leak identification. Personnel discovered a through-wall piping leak on a 3/4 inch test line upstream of the test valve. This 3/4 inch test line is located on the RHR Shutdown Cooling & RHR Return Line to the reactor vessel and is classified as ASME Code Class 1 Piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC resident has been notified. | |
| ENS 55590 | 18 November 2021 02:21:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | At 2238 Eastern Standard Time (EST), on 11/17/2021, a Watts Bar Nuclear Plant contractor was transported offsite for treatment at an offsite medical facility. The offsite medical facility notified Watts Barr Nuclear Plant at 2310 EST that the individual had been declared deceased. The fatality was not work-related and the individual was inside of the Unit 1 Radiological Controlled Area. The individual was confirmed not to be contaminated. This is a four-hour notification, non-emergency for a notification of other government agency. This event is reportable 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. The licensee will notify OSHA. | Fatality |
| ENS 55550 | 28 October 2021 14:19:00 | Watts Bar | NRC Region 2 | Westinghouse PWR 4-Loop | At 1340 EDT on October 28, 2021, Watts Bar Nuclear Plant (WBN) Units 1 and 2 initiated voluntary communication to the State of Tennessee and local officials as part of the Nuclear Energy Institute (NEI) Groundwater Protection Initiative (GPI), after receiving analysis results for two on-site monitoring wells that indicated tritium activity above the GPI voluntary communication threshold. The suspected source, a permitted release line, has been isolated, and additional corrective actions are in progress. This condition did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable 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 55660 | 16 December 2021 14:57:00 | Browns Ferry | NRC Region 2 | GE-4 |
This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of the Reactor Protection System (RPS). On October 20, 2021, at approximately 0705 hours Central Daylight Time (CDT), Browns Ferry, Unit 1, 1B RPS bus unexpectedly lost power. The loss of the bus resulted in a half scram, automatic Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolations, and Trains A, B, and C SBGT (Stand-By Gas Treatment) and A CREV (Control Room Emergency Ventilation system) started. All systems responded as expected. At 0720 hours CDT, the bus was placed on the alternate power supply and the half scram and PCIS isolations were reset. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the RPS bus loss was a trip of the underfrequency relay due to drift of the relay setpoint. The relay was replaced and 1B RPS bus was returned to the normal power supply on October 21, 2021, at 0510 hours CDT. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1729592. The NRC Resident Inspector has been notified of this event. | Reactor Vessel Water Level Half scram |