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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 55964 | 25 June 2022 03:38:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Main Turbine Trip | The following information was provided by the licensee via email: At 2338 EDT, on June 24, 2022, with the unit in Mode 1 at 100 percent power, the reactor automatically scrammed due to an RPS actuation following a Main Turbine Trip. The cause of the turbine trip is not known at this time. The scram was not complex, with systems responding normally post-scram. Operations responded and stabilized the plant. Reactor water level has been recovered and maintained at the normal level. Decay Heat is being removed by the Main Steam system to the main condenser using the Turbine Bypass Valves. All Control Rods inserted into the core. The transient occurred with no surveillances or activities in progress. Investigation into the cause of the Turbine Trip is in progress. 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 low reactor water level caused an isolation of Primary Containment (Groups 4/13/15) as expected. The Primary Containment Isolation Event is being reported under 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. | Reactor Protection System Primary containment Main Condenser Control Rod Main Steam | |
| ENS 55732 | 4 February 2022 22:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | The following information was provided by the licensee via email: At 1700 EST, on February 4, 2022 with the unit in Mode 1 at 58 percent power, the reactor automatically scrammed due to low Reactor water level due to a transient on the Feedwater System while preparing to shutdown for a refueling outage. The scram was not complex, with systems responding normally post-scram. Operations responded and stabilized the plant. Reactor water level has been recovered and maintained at normal level. Decay Heat is being removed by the Main Steam system to the main condenser using the Turbine Bypass Valves. All Control Rods inserted into the core. The transient occurred while in the process of removing the South Reactor Feed Pump from service. While reducing speed on the South, the North Reactor Feed Pump increased in speed and tripped on low suction. The plant was preparing to shut down for a refueling outage when the trip occurred. 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). Additionally, in preparation of plant shutdown, Primary Containment De-Inerting was in progress. The low Reactor water level caused an isolation of Primary Containment (Groups 4/13/15). The Primary Containment Isolation Event is being reported under 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. | Feedwater Reactor Protection System Primary containment Main Condenser Control Rod Main Steam | |
| ENS 54761 | 2 July 2020 03:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Offsite Power - Auto Initiation of Emergency Diesel Generator | At 2305 EDT on July 1, 2020, while in Mode 5 for Refueling Outage 20 with no core alterations in progress, Fermi 2 experienced a loss of Division 2 offsite power (345 kV) which resulted in a valid automatic initiation of the Division 2 Emergency Diesel Generators (EDG) 13 and 14. EDG 13 and 14 started as expected to supply their associated busses. Division 1 offsite power remains operable and powering the Division 1 Residual Heat Removal (RHR) system in Shutdown Cooling (SDC) mode of operation. Division 1 EDGs 11 and 12 remain operable and available. The cause of the loss of Division 2 offsite power is under review and has preliminarily been determined to be caused by Mayfly accumulation in and around the Division 2 (345 kV) switchyard. Actions have been put in place to minimize and deter Mayflies from gathering near plant switchyards. All systems responded as expected for the loss of Division 2 offsite power and no loss of SDC occurred. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The event is reportable pursuant to 10 CFR 50.72(b)(3)(iv)(A), as a valid specified system actuation. | Emergency Diesel Generator Shutdown Cooling Residual Heat Removal | |
| ENS 53336 | 14 April 2018 14:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Scram and Emergency Core Cooling System Injection | At 1040 EDT, Fermi 2 automatically scrammed on RPV (Reactor Pressure Vessel) Level 3 following a loss of the Division 1 Station System Transformer (SST) #64. All control rods fully inserted. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically started as designed on Reactor Water Level (RWL) 2 and restored RWL. The lowest RWL reached was 101.8 inches (above Top of Active Fuel). HPCI injected for approximately 35 seconds. RWL is currently being maintained in the normal level band with RCIC. No Safety Relief Valves (SRVs) actuated. All isolations and actuations for RWL 3 and 2 occurred as expected. Investigation into loss of SST #64 continues. At the time of the scram, all Emergency Core Cooling Systems (ECCS) and Emergency Diesel Generators (EDGs) were operable, and no safety related equipment was out of service. This report is being made in accordance with 10CFR50.72(b)(2)(iv)(A), any event that results in ECCS discharge into the reactor coolant system as a result of a valid signal and 10CFR50.72(b)(2)(iv)(B), any event that results in the actuation of the Reactor Protection System (RPS) when the reactor is critical. Following the loss of power and reactor scram, the Division 2 EECW (Emergency Equipment Cooling Water) Temperature Control Valve (TCV) controller was in Emergency Manual and maintaining max cooling. Operators placed the controller in Auto and the TCV is controlling normally. The NRC Senior Resident has been notified. Decay heat is being removed via Division 2 steam dumps to the condenser. The plant is in a modified shutdown electric lineup with offsite power available and stable. Emergency diesel generators did automatically start and load.
This update provides additional clarification of the applicable reporting criteria for this event associated with Primary Containment Isolation Actuations. All isolations and actuations for RWL (Groups 4, 13, and 15) and RWL 2 (Groups 2, 10, 11, 12, 14, 16, 17, and 18) occurred as expected. This report is also being made in accordance with 10CFR50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any systems listed in paragraph (b)(3)(iv)(B): RPS, HPCI, and RCIC. RPV pressure is being maintained by the bypass valves to the main condenser. All actuations that occurred were fully completed and restored. The licensee notified the NRC Resident Inspector. Notified R3DO (Stone).
This update provides additional information regarding the specified system actuations and an additional applicable reporting criteria. The loss of Division 1 Station System Transformer (SST) #64 at 1040 EDT on 4/14/2018 resulted in the automatic initiation of Emergency Diesel Generators (EDG) 11 and 12. The EDGs started as expected and continue to supply their associated busses. This is reportable pursuant to 10CFR50.72(b)(3)(iv)(A), as an event or condition that resulted in a valid actuation of any system listed in paragraph (b)(3)(iv)(B), including EDGs. In addition, the loss of the Division 1 SST #64 resulted in the expected transfer from the normal to alternate power source for the Low Pressure Coolant Injection (LPCI) swing bus, rendering LPCI loop select inoperable. The alternate power source continued to energize the LPCI swing bus throughout the event until the system was realigned to the normal power source at 1239 EDT on 4/14/2018. This condition is reportable pursuant to 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified R3DO (Stone). | Reactor Coolant System Reactor Protection System Emergency Diesel Generator Primary containment Emergency Core Cooling System Safety Relief Valve Main Condenser Control Rod Low Pressure Coolant Injection | |
| ENS 51391 | 14 September 2015 03:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Manual Scram Due to Loss of Turbine Building Closed Cooling Water | At 2305 EDT on September 13, 2015, a manual scram was initiated in response to a loss of all Turbine Building Closed Cooling Water (TBCCW). All control rods fully inserted. The lowest Reactor Water Level (RWL) reached was 137 inches. All isolations and actuations for RWL 3 occurred as expected. Decay heat was initially being removed through the Main Turbine Bypass System to the Main Condenser, however, as a result of the loss of TBCCW, the Main Feed Pumps lost cooling and had to be secured. At 2310, Standby Feedwater was initiated and Main Feedwater was secured. The loss of TBCCW also caused all Station Air Compressors (SACs) to trip on loss of cooling. The loss of SACs caused the Instrument Air header pressure to degrade to the point at which the Secondary Containment isolation dampers drifted closed. This resulted in the Reactor Building vacuum exceeding the Technical Specification limit. At 2325, operators started the Standby Gas Treatment system and manually initiated a Secondary Containment isolation signal. Secondary Containment vacuum was promptly restored to within Technical Specification limits. Additionally, Operators were monitoring for expected MSIV drift due to the degraded Instrument Air header pressure. When outboard MSIVs were observed to be drifting, Operators closed the outboard and inboard MSIVs at 2345. At 2352, Safety Relief Valves (SRVs) reached the Low-Low Setpoint and began cycling to control reactor pressure. RWL is currently being maintained in the normal level band with the Standby Feedwater and Control Rod Drive systems. Reactor Pressure is being controlled with Safety Relief Valves. Operators are currently in the Emergency Operating Procedure for Reactor Pressure Vessel control. Investigation into the loss of TBCCW continues. No safety-related equipment was out of service at the time of the event. All offsite power sources were adequate and available throughout the duration of the event. The NRC resident inspector has been notified.
At 0409 EDT the Reactor Core Isolation Cooling (RCIC) system was placed in service due to identification of an unisolable leak in the Standby Feedwater System. Reactor water level and pressure is now being controlled though the RCIC system and Safety Relief Valves. This event update is reportable as a valid manual initiation of a specified safety system under 10CFR50.72(b)(3)(iv)(A). The NRC resident inspector has been notified. The leak rate was reported as approximately 5-10 gallons per minute from a weld on the standby feedwater pump header drain valve F326. The licensee reported the leak stopped once RCIC was placed into service. The licensee is still investigating the issue. Notified the R3DO (Pelke), IRD Manager (Grant), NRR EO (Morris).
At 1847 EDT on September 14, 2015, a valid automatic Reactor Protection System (RPS) actuation occurred due to Reactor Water Level 3 while shutdown in MODE 3. Operators were manually controlling Reactor Pressure Vessel (RPV) level and pressure with Reactor Core Isolation Cooling (RCIC) and Safety Relief Valves (SRV). While operators were cycling SRVs, the RPV level went below the Level 3 setpoint. Operators promptly restored RPV level by manual operation of RCIC. The Level 3 actuation and associated isolations were verified to operate properly. The scram signal has been reset. Fermi 2 remains in MODE 3 controlling RPV Level and Pressure through manual operation of RCIC and SRVs. This is the second occurrence of a valid specified safety system actuation reportable under 10CFR50.72(b)(3)(iv)(A) for this ongoing event. The NRC Resident Inspector has been notified. Notified R3DO (Riemer), IRD Manager (Grant), and NRR EO (Morris)
This update provides clarification of the applicable reporting criteria for this Event associated with primary containment isolation actuations. Upon the manual reactor scram at 2305 EDT on September 13, 2015, Reactor Protection System (RPS) Level 3 actuated and Primary Containment Isolation System (PCIS) Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for these actuations is 10 CFR 50.72(b)(3)(iv)(A). The applicable reporting criterion for the manual closure of the inboard and outboard main steam isolation valves at 2345 EDT on September 13, 2015, is also 10 CFR 50.72(b)(3)(iv)(A). In addition, the manual closures of all MSIV lead to a loss of condenser vacuum which resulted in the actuation of PCIS Group 1 at 0001 EDT on September 14, 2015, as expected. The applicable reporting criterion for this actuation is also 10 CFR 50.72(b)(3)(iv)(A). Upon reaching Level 3 at 1847 EDT on September 14, 2015, PCIS Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for this actuation is 10 CFR 50.72(b)(3)(iv)(A). The licensee informed the NRC Resident Inspector. Notified the R3DO (Stone). | Feedwater Secondary containment Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Primary containment Main Turbine Reactor Pressure Vessel Standby Gas Treatment System Safety Relief Valve Main Condenser Control Rod | |
| ENS 47868 | 26 April 2012 14:12:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of the Reactor Protection System During Testing | At 1012 EDT on April 26, 2012, during the Reactor Pressure Vessel Hydrostatic Test, a valid high pressure reactor scram occurred due to issues related to controlling pressure near rated values. This actuation of the Reactor Protection System was not part of the pre-planned testing sequence. All control rods were fully inserted at the time of the scram. This report is being made in accordance with 10CFR50.72 (b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' The reactor scram was reset after reactor pressure was lowered. The licensee has notified the NRC Resident Inspector. | Reactor Protection System Reactor Pressure Vessel Control Rod | 05000341/LER-2012-002 |
| ENS 42738 | 29 July 2006 19:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram and Eccs Injection Due to Loss of Feedwater | At 15:50 EDT on 7/29/06, a Level 3 reactor scram occurred due to a loss of feedwater. The loss of feedwater was caused by a loss of Division 1 electrical power. All control rods fully inserted into the core. The lowest reactor vessel water level reached was 110 inches. HPCI and RCIC auto initiated on Level 2 and injected into the reactor pressure vessel. The Division 1 Emergency Diesel Generators auto initiated and supplied the Division 1 ESF buses. Level 3 and Level 2 isolations occurred as expected. Reactor water level is now being controlled in the normal water level band using Standby Feedwater. No SRVs lifted and RPV pressure is being controlled by the Turbine Pressure Regulator with the main condenser available as the heat sink. At the time of the scram, work was being performed on the 120kV mat which resulted in a loss of Bus 101. Group 13 (Drywell Sumps) isolated on Reactor Water Level 3. Group 10 (Reactor Water Cleanup Inboard), Group 11 (Reactor Water Cleanup Outboard), Group 12 (Torus Water Management System), Group 17 ( Reactor Recirc Pump Seals and Primary Containment Radiation Monitoring), and Group 18 (Primary Containment Pneumatic Supply) isolated on Reactor Water Level 2. The licensee notified the NRC Resident Inspector. The licensee stated that HPCI injected for 2 minutes and was then secured as Standby Feedwater was started. At the time of the notification, Bus 101 had been re-energized, and preparations were being made to restore normal power to the Division 1 buses and return the Emergency Diesel Generators to standby. The licensee is investigating the exact cause of the loss of power.
The purpose of this report is to update the information provided at 19:19 ET on 7/29/2006. This event was originally reported under reporting criteria 50.72(b)(2)(iv)(A) as an ECCS injection. It has subsequently been determined that both the HPCI and RCIC systems auto-started in response to a reactor low water level 2 (Level 2) injection signal, however, only the RCIC system injected into the vessel. The Level 2 signal was only present for about 2.7 seconds until reactor water level recovered above Level 2. The HPCI injection logic is such that the Level 2 signal must be present until HPCI startup has completed. This includes time for the hydraulic pressure from the HPCI Auxiliary Oil Pump to develop enough pressure to open the HPCI turbine steam isolation valve (E4100F067) and time to stroke open the motor operated HPCI turbine steam isolation valve (E4100F001). It took about 12 seconds before steam was admitted to the HPCI turbine. Thus, the HPCI main pump outlet valve (E4150F006) did not open due to the short duration of the Level 2 signal. This is consistent with the HPCI system design. Therefore the event was not reportable as an event that resulted in or should have resulted in an ECCS injection into the reactor vessel. The event remains reportable under criteria 50.72(b)(2)(iv)(B) and 50.72(b)(3)(iv)(A). Additional clarification of the cause of the scram is also provided. The loss of bus 101 resulted in the loss of power to the operating south reactor feed pump (SRFP) turbine lube oil pump resulting in a loss of feedwater flow from the SRFP. The north reactor feed pump continued to operate. The plant is designed with an automatic runback of the recirculation system to allow continued operation following the loss of a single feed pump. However, the loss of bus 101 also resulted in the locking of the reactor recirculation pump speeds (scoop tube lock), disabling the runback feature. This led to a reactor scram on reactor low water level 3 (Level 3) since a single feed pump is not able to maintain reactor water level at 100% power operation. When south reactor feed pump lubrication pressure recovered, feedwater flow from the SRFP recovered. Recovering feedwater injection from the SRFP following the scram caused a rapid increase in reactor water level and a high reactor water level 8 (Level 8) shutdown of the HPCI, RCIC and reactor feedwater pumps. The standby feedwater system was subsequently started and used to maintain reactor level. The plant is restarting, and is in Mode 2 with reactor temperature at approximately 508�F and reactor pressure at approximately 817 psi at the time of this report. Based on this update, ECCS injection was removed from CFR Section of the report. The licensee notified the NRC Resident Inspector. Notified R3 DO (K. O'Brien) | Feedwater | 05000341/LER-2006-003 |