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	Entered date	Site	Scram	Region	Reactor type	Event description
ENS 57021	11 March 2024 15:46:00	Hatch	Manual Scram	NRC Region 2	GE-4	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.
ENS 56846	10 November 2023 03:14:00	Susquehanna	Manual Scram	NRC Region 1	GE-4	The following information was provided by the licensee via email: At 0118 EST, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually scrammed due to degrading main condenser vacuum. The scram was not complex, with all systems responding normally post-scram. The main turbine bypass valves opened automatically to maintain reactor pressure. Operations responded and stabilized the plant. Reactor water level is being maintained via feedwater pumps. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not impacted. Due to Reactor Protection System actuation while critical, this event is being reported as a four-hour and eight-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Unit 1 reactor is currently stable in mode 3. An investigation is in progress into the cause of the degrading condenser vacuum. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 56826	1 November 2023 09:38:00	Hatch	Manual Scram	NRC Region 2	GE-4	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.
ENS 56478	20 April 2023 05:24:00	Brunswick	Automatic Scram	NRC Region 2	GE-4	The following information was provided by the licensee via phone and email: At 0148 Eastern Daylight Time (EDT) on April 20, 2023, with Unit 1 in Mode 1 at 100% power, the reactor automatically tripped due to a turbine trip. Turbine Bypass valves did not open on the trip due to Turbine Protection system power supply failure; the Safety Relief Valves (SRVs) opened automatically to control reactor pressure. Reactor Pressure reached approximately 1095 psig on the trip; exceeding the 1060 psig RPS trip setpoint. Operations responded and stabilized the plant. Operations was able to transition from SRVs to main steam line drains to the condenser. Reactor water level is being maintained via the Condensate / Feedwater system. Decay heat is being removed by discharging steam to the main condenser using the main steam line drains. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Reactor water level reached low level 1 (LL1) following the reactor trip. The LL1 signal causes Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves), and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the valid Primary Containment Isolation System (PCIS) actuation and RPS actuation from the reactor pressure signal, this event is also being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 2 is not affected by this event. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 56446	31 March 2023 15:54:00	Hatch	Manual Scram	NRC Region 2	GE-4	The following information was provided by the licensee via email: At 1432 EDT on 03/31/23, with Unit 2 in mode 1 at 97 percent power, the reactor was manually tripped due to a loss of both recirculation pumps. The cause of the recirculation pump trips is under investigation. Additionally, closure of CIVs in multiple systems occurred during the trip as a result of reaching the actuation setpoint on reactor water level as designed. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via condensate / feedwater. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. 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). It is also reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 56278	17 December 2022 04:03:00	Cooper	Manual Scram	NRC Region 4	GE-4	The following information was provided by the licensee via email: On December 16, 2022 at 2351 CST, with the Unit in Mode 1 at 13 percent power, a manual scram was inserted due to lowering Reactor Pressure Vessel (RPV) pressure, which occurred following an unexpected opening of Main Turbine Bypass Valve 1. All control rods fully inserted. Following actuation of the manual scram, RPV pressure lowered, resulting in an automatic Primary Containment lsolation (PCIS) Group 1 isolation (expected response). The main steam isolation valves and steam line drain valves all closed. The Group 1 (isolation) has been reset allowing RPV pressure control with steam line drains to the main condenser. All systems responded as designed. The plant is stable in Mode 3. Investigation of the bypass valve opening is ongoing. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation and 50.72(b)(3)(iv)(A) Specified System Actuation. There was no impact on health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified.
ENS 55964	25 June 2022 01:00:00	Fermi	Automatic Scram	NRC Region 3	GE-4	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.
ENS 55909	23 May 2022 21:01:00	Susquehanna	Automatic Scram	NRC Region 1	GE-4	The following information was provided by the licensee via email: At 1716 hours EDT on May 23, 2022, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed. Unit 1 reactor was being operated at approximately 100 percent (Rated Thermal Power) RTP. The Control Room received indication that both divisions of (Reactor Protection System) RPS actuated from (Reactor Pressure Vessel) RPV high pressure signals and all control rods fully inserted. The Main Turbine bypass valves opened automatically to control reactor pressure. Reactor water level lowered to -42 inches causing Level 3 and Level 2 isolations. (High Pressure Coolant Injection) HPCI (Emergency Core Cooling System) ECCS actuation occurred as designed at -38 inches and injected to the Reactor Vessel. No other ECCS system actuations occurred. (Reactor Core Isolation Cooling) RCIC automatically initiated as designed at -30 inches. The Operations crew subsequently maintained reactor water level at the normal operating band using Feedwater pumps. The reactor is currently stable in Mode 3. An investigation is in progress into the cause of the Automatic SCRAM. The NRC Senior Resident Inspector was notified. A voluntary notification to (Pennsylvania Emergency Management Agency) PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10 CFR 50.72(b)(2)(iv)(A) & 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A).
ENS 55899	16 May 2022 19:51:00	Peach Bottom	Automatic Scram	NRC Region 1	GE-4	The following information was provided by the licensee via fax: Unit 2 experienced multiple electrical transients resulting in a Group I Primary Containment Isolation Signal (PCIS) isolation and subsequent unit reactor scram. Low reactor water level during the automatic scram caused PCIS Group II and III isolation signals. Following the PCIS Group I isolation, all main steam lines isolated. All control rods inserted and all systems operated as designed. The following additional information was obtained from the licensee via phone in accordance with Headquarters Operations Officers Report Guidance: Peach Bottom Unit 2 automatically scrammed from 100 percent power due to an electrical transient and subsequent PCIS Group I isolation (Main Steam Isolation Valve closure). Unit 2 lost main feedwater due to the PCIS Group I isolation, however, all other systems responded as expected following the scram. High Pressure Coolant Injection is maintaining pressure control while Condensate Pumps are maintaining inventory. The unit is currently stable and in Mode 3. Peach Bottom Unit 3's Adjustable Speed Drives were impacted by the electrical transients and the unit reduced power to 98 percent power. The NRC Resident Inspector was notified.
ENS 55894	11 May 2022 22:25:00	Fermi	Manual Scram	NRC Region 3	GE-4	The following information was provided by the licensee via email: During performance of High Pressure Coolant Injection (HPCI) Pump and Valve Operability surveillance in accordance with procedure 24.202.01, the turbine tripped without operator action. The plant was operating in Mode 1 at 10 percent power with the HPCI turbine running in a test mode at 5100 gpm with all surveillance criteria met. The surveillance was near completion at the point where the HPCI turbine is manually tripped. Before the manual trip was performed, the HPCI turbine tripped without operator action. Prior to performance of the surveillance, HPCI was provisionally operable with only satisfactory completion of Post Maintenance Testing (PMT) surveillance remaining to declare HPCI operable. HPCI surveillance testing was performed at low reactor pressure (165 psig) in Mode 2 satisfactorily. Investigation into the cause of this trip is in progress. HPCI has been declared inoperable from the time of release of the surveillance. Reactor Coolant Isolation Cooling (RCIC) was verified to be operable prior to and after the surveillance in accordance with Technical Specifications 3.5.1 condition E.1. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55679	29 December 2021 19:16:00	Hatch	Manual Scram	NRC Region 2	GE-4	This following information was conveyed by the licensee via phone and email: At 1552 EST on 12/29/21, with Unit 1 in Mode 1 at 90 percent power, the reactor was manually tripped due to reactor pressure perturbations. The cause of the reactor pressure perturbations is under investigation. Additionally, closure of (containment isolation valves) CIVs in multiple systems occurred during the trip as a result of reaching the actuation setpoint on reactor water level. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via condensate / feedwater. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. 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). It is also reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55575	14 November 2021 08:50:00	Peach Bottom	Manual Scram	NRC Region 1	GE-4	At 0525 EST, November 14, 2021, "Unit 2 was manually scammed by operations due to lowering main condenser vacuum. This resulted in PCIS (primary containment Isolation system) Group II/III isolation signals. All control rods inserted, and all systems operated as designed. Unit 3 is unaffected and remains at 100 percent power in Mode 1. The Resident Inspector was notified.
ENS 55514	11 October 2021 17:05:00	Susquehanna	Automatic Scram	NRC Region 1	GE-4	At 1321 EDT on October 11, 2021, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed due to a trip of the Main Turbine. Unit 2 reactor was being operated at approximately 95 percent RTP (rated thermal power) with no evolutions in progress. The Control Room received indication of a Main Turbine trip with both divisions of RPS (Reactor Protection System) actuated and all control rods inserted. Turbine bypass valves opened automatically to control reactor pressure and subsequently failed open causing the reactor to depressurize. When reactor pressure reached approximately 560 psig, the operations crew manually closed the Main Steam Isolation Valves (MISVs) to stop the depressurization. Reactor water level lowered to -31 inches causing Level 3 (+13 inches) isolations. No (automatic) ECCS (Emergency Core Cooling System) actuations occurred. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) were manually initiated to control reactor water level. The Operations crew subsequently maintained reactor water level at the normal operating band using RCIC and reactor pressure was controlled with HPCI in pressure control mode and main steam line drains. The Reactor Recirculation Pumps tripped as designed on EOC-RPT (end of cycle recirculation pump trip). The reactor is currently stable in Mode 3. An investigation into the cause of the turbine trip is underway. The NRC Resident Inspector was notified. A voluntary notification to PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A).
ENS 54996	13 November 2020 05:32:00	Limerick	Automatic Scram	NRC Region 1	GE-4	At 0245 EST on November 13, 2020, the Limerick Unit 1 reactor automatically scrammed on a valid Reactor High Pressure signal (1096psig). The Reactor High Pressure signal was caused by the closure of the 1B Inboard Main Steam Isolation Valve (MSIV), causing reactor pressure to rise, exceeding the Reactor Protection System (RPS) setpoint of 1096psig. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Steam Bypass Valves to the Main Condenser and normal level control using the Feedwater System. The closure of the 1B Inboard MSIV appears to have been caused by a loss of Primary Containment Instrument Gas (PCIG) pneumatic supply to the valve. The licensee notified the NRC Resident Inspector, and will be notifying Berks, Chester, and Montgomery Counties, as well as the Pennsylvania Emergency Management Agency.
ENS 54976	1 November 2020 09:34:00	Cooper	Manual Scram	NRC Region 4	GE-4	On November 1, 2020, at 0534 CST the reactor was manually scrammed due to an un-isolable leak on the Turbine High Pressure Fluid System. Initial power level when the leak was identified was 100 percent. Power was lowered commencing at 0525 in accordance with shutdown procedures. The Reactor Operator scrammed the reactor at 0534 from approximately 75 percent power. Following the scram, Reactor vessel water level lowered to approximately -20 inches on the Wide Range Instruments, and was subsequently recovered to normal post scram range (approximately 36 inches) using the Reactor Feedwater system. Group 2 Isolation occurred due to Reactor vessel level reaching the isolation setpoint (3 inches). The plant is stable in MODE 3 and proceeding to cold shutdown. The Main Condenser remained available throughout the evolution and condenser vacuum is currently being maintained by the Mechanical Vacuum Pumps. Pressure is being controlled using the steam line drains to the main condenser. All control rods fully inserted and there were no complications. All systems responded as designed. The Turbine High Pressure Fluid System has been secured. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) due to RPS Actuation-Critical and 50.72(b)(3)(iv)(A) Valid Specified System Actuation. The licensee has notified the NRC Resident Inspector.
ENS 53336	14 April 2018 14:34:00	Fermi	Automatic Scram	NRC Region 3	GE-4	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. * * UPDATE ON 4/14/2018 AT 1838 EDT FROM JEFF MYERS TO HOWIE CROUCH * * * 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). * * UPDATE ON 4/15/2018 AT 1950 EDT FROM KELLEY BELENKY TO DAVID AIRD * * * 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).
ENS 53319	7 April 2018 12:10:00	Brunswick	Automatic Scram	NRC Region 2	GE-4	On April 7, 2018, at 0836 EDT, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped during testing of the stator cooling system. The trip was uncomplicated with all systems responding normally. No safety-related equipment was inoperable at the time of the event. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Operations responded using Emergency Operating Procedures and stabilized the plant in Mode 3. Reactor water level being maintained via normal feedwater system. Decay heat is being removed through the bypass valves. Reactor water level reached low level 1 (LL1) as a result of the reactor trip. The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the Primary Containment Isolation System (PCIS) actuation, this event is also being reported as an eight-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PCIS. Unit 2 was not affected. There was no impact on the health and safety of the public or plant personnel. The safety significance of this event is minimal. The automatic reactor trip was not complicated and all safety-related systems operated as designed. Investigation of the cause of the Reactor Protection System actuation is in progress. The licensee notified the NRC Resident Inspector.
ENS 53269	18 March 2018 16:16:00	Browns Ferry	Automatic Scram	NRC Region 2	GE-4	At 1158 CDT on March 18, 2018, the Unit 1 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from High Reactor Steam Dome Pressure in response to Turbine Control Valve Closure. The reactor had been operating at 100 percent power. Investigation is in progress. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Steam Relief Valves (MSRVs) operating on the initial transient as expected. Main Turbine Bypass Valves are currently controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. All safety system operated as expected. At no time was public health and safety at risk. This event is reportable within 4 hours 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.' It is also reportable within 8 hours 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) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation' and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 53162	10 January 2018 13:53:00	Browns Ferry	Automatic Scram	NRC Region 2	GE-4	At 0928 CST on January 10, 2018, the Unit 3 reactor automatically scrammed due to a Reactor Protection System (RPS) signal generated from Turbine Control Valve Emergency Trip System pressure low. The reactor had been operating near 73 percent power for an emergent issue for Turbine Control Valve (TCV) No. 3. With TCV No. 3 out of service and closed, the unit was operating with RPS in a half scram condition. A subsequent failure of the TCV No. 2 sensing line resulted in RPS coincidence logic being met for TCV fast closure SCRAM. The investigation of the TCV No. 2 sensing line failure continues. All control rods fully inserted into the core. Main Steam Isolation Valves remained open with Main Turbine Bypass Valves controlling reactor pressure. Reactor Feedwater pumps remained in service to control reactor water level. Primary Containment Isolation Signals Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. This event is reportable within 4 hours 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 preplanned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident inspector has been notified.
ENS 52795	8 June 2017 19:10:00	Susquehanna	Automatic Scram	NRC Region 1	GE-4	At 1527 hrs (EDT) on June 8, 2017, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a loss of Main Turbine Electro-Hydraulic Control (EHC) logic power causing a High Flux Reactor Power RPS (Reactor Protection System) trip. All control rods (fully) inserted and both reactor recirculation pumps tripped due to reaching reactor water level 2. Reactor water level lowered to -49 inches causing Level 3 (+13 inches) and Level 2 (-38 inches) isolations. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically initiated and were overridden by control room operators after RPV (Reactor Pressure Vessel) water level was restored to the normal band with feedwater. HPCI and RCIC injected to the Reactor Coolant System during reactor level stabilization. All isolations and initiations occurred as expected. No main steam relief valves opened. Pressure was controlled via main turbine bypass valve operation. All safety systems operated as expected. Secondary Containment Zone 1, 2, and 3 differential pressure lowered to 0 inch WG (Water Gauge) due to a trip of the Reactor Building Ventilation system that resulted from Unit 1 Level 2 isolation. Differential pressure was restored to Zones 1, 2, and 3 by the initiation of Standby Gas Treatment System on the Unit 1 Level 2 initiation. Unit 1 reactor is currently stable in Mode 3. Investigation into the loss of Main Turbine EHC logic power is underway. The NRC Resident Inspector has been notified. A voluntary notification to PEMA and press release will occur. The suspected cause of the loss of power to the EHC logic circuit is ongoing maintenance on the system.
ENS 52696	20 April 2017 05:57:00	Hatch	Automatic Scram	NRC Region 2	GE-4	On 04/20/2017 at 0302 EST during a reactor startup, a reactor scram resulted from upscale spike on two Intermediate Range Monitors (IRMs), 1C51K601A and 1C51K601B. IRM A, 1C51K601A is in Reactor Protection System Channel A and IRM B, 1C51K601B is in Reactor Protection System Channel B. All control rods fully inserted. No PCIS (Primary Containment Isolation System) actuations occurred and none were expected to occur based upon plant condition following the reactor scram. Investigation is in progress. Condition was not due to a true flux event. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) as an event or condition that resulted 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. CR 10356172 The NRC Resident has been notified. The reactor was at 0.5% (percent) power at the time of the event and will remain in Hot Shutdown pending the results of the root cause investigation.
ENS 52648	29 March 2017 23:36:00	Browns Ferry	Manual Scram	NRC Region 2	GE-4	At 1844 CDT on 3/29/2017, Unit 2 initiated a manual scram due to multiple rods inserting. At 1842 during Unit 2 start-up, Intermediate Range Monitor (IRM) 'G' drifted low. The operator adjusted the range down one position with no immediate reaction. At 1844, a spike on IRM 'G' caused a half scram on Reactor Protection System (RPS) 'A' trip system. The half scram was being reset after evaluating no trip condition was present. As the operator reset groups 2 and 3, a trip signal from IRM 'F' was received on the RPS 'B' trip system, resulting in rod insertion for groups 1 and 4. When the operator identified multiple rods inserting, the actions of procedure 2-AOI-100-1 were followed and a manual scram was inserted. Investigation is ongoing. All safety systems remained in standby readiness configuration. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary Containment Isolations Systems did not receive an actuation signal and performed as designed. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the 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.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) 'any event or condition that results in valid actuation of systems listed in paragraph (b)(3)(iv)(B) Reactor Protection System(RPS) including reactor scram and reactor trip'. This event requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified.
ENS 52042	24 June 2016 16:06:00	FitzPatrick	Manual Scram	NRC Region 1	GE-4	At 1215 (EDT) on 6/24/2016, James A. FitzPatrick (JAF) was at 100% power when Breaker 710340 tripped and power was lost to L-gears L13, L23, L33, and L43. These provide non-vital power to Reactor Building Ventilation (RBV), portions of Reactor Building Closed Loop Cooling (RBCLC), and 'A' Recirculation pump lube oil systems. Off-site AC power remains available to vital systems and Emergency Diesel Generators (EDG) are available. Due to the loss of RBV, Secondary Containment differential pressure increased. At 1215 (EDT), Secondary Containment differential pressure exceeded the Technical Specifications (TS) Surveillance Requirement SR-3.6.4.1.1 of greater than or equal to 0.25 inches of vacuum water gauge. The Standby Gas Treatment (SBGT) system was manually initiated and Secondary Containment differential pressure was restored by 1219 (EDT). The 'A' Recirculation pump tripped at 1215 (EDT) and reactor power decreased to approximately 50%. 'B' Recirculation pump temperature began to rise due to the degraded RBCLC system. At 1236 (EDT), a manual scram was initiated. Reactor Pressure Vessel (RPV) water level shrink during the scram resulted in a successful Group 2 isolation. All control rods have been inserted. The RPV water level is being maintained with the Feedwater System and pressure is being maintained by main steam line bypass valves. A cooldown is in progress and JAF will proceed to cold shutdown (Mode 4). Due to complete loss of RBCLC system, the Spent Fuel Pool (SFP) cooling capability is degraded but the Decay Heat Removal system remains available. SFP temperature is slowly rising and it is being monitored. The time (duration) to 200 degrees is approximately 117 hours. The initiation of reactor protection systems (RPS) due to the manual scram at critical power is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The general containment Group 2 isolations are reportable per 10 CFR 50.72(b)(3)(iv)(A). In addition, the temporary differential pressure change in Secondary Containment is reportable per 10 CFR 50.72(b)(3)(v)(C), as an event that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector and the State of New York.
ENS 51968	1 June 2016 10:40:00	Limerick	Manual Scram	NRC Region 1	GE-4	Limerick Unit 2 was manually scrammed from 100 (percent) power at 0900 (EDT) on 6/1/2016 in accordance with plant procedure OT-112 'Unexpected/Unexplained change in core flow' when both 2A and 2B Recirculation Pump Adjustable Speed Drives (ASDs) tripped due to an electrical fault. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Steam Bypass valves to the Main Condenser and normal level control using Feedwater. The Manual RPS actuation is reportable under 10 CFR 50.72 (b)(2). All rods inserted fully on manual scram and the plant is in a normal shutdown electrical line up. Unit 1 was not affected by this event. The licensee plans to issue a press release. The licensee notified local counties and Pennsylvania Emergency Management Agency (PEMA). The NRC Resident Inspector has been notified.
ENS 51925	13 May 2016 05:00:00	Susquehanna	Manual Scram	NRC Region 1	GE-4	At approximately 0110 hours (EDT) on May 13, 2016, Susquehanna Steam Electric Station Unit Two reactor was manually scrammed by plant operators due to a sustained loss of AC power to essential plant loads. Power to MCC 2B246 was lost at 2355 on May 12, 2016, resulting in a loss of Drywell cooling. Drywell pressure increased to 1.3 psig when operators placed the mode switch to the shutdown position to manually SCRAM the reactor. All rods inserted as expected. Reactor water level lowered to -27 inches and was immediately restored by normal feedwater level control. Level 3 (+13 inch) PCIS isolations occurred, along with an initiation of the RCIC system (-30 inches). Once adequate level was verified, RCIC was overridden. Pressure was controlled with turbine bypass valves, and subsequently main steam line drains. All safety systems functioned as expected. The power loss also tripped Reactor Building HVAC, causing a loss of secondary containment differential pressure resulting in a loss of safety function. Due to the loss of drywell cooling, high drywell pressure actuations and a second reactor SCRAM signal, this signal was automatic, occurred at 0314 hours. HPCI (which automatically initiated on high drywell pressure) was subsequently overridden and declared inoperable, resulting in a loss of safety function. (HPCI did not inject into the vessel). The reactor is currently stable in Mode 3. Initial reports from the field indicate a phase to phase fault on the MCC 2B246 bus bars. The licensee has notified the NRC Resident Inspector and will be issuing a press release.
ENS 51715	7 February 2016 13:46:00	Brunswick	Manual Scram	NRC Region 2	GE-4	At 1346 EST the licensee reported that at 1326, Brunswick Unit 1 declared an Alert under EAL HA 2.1 due to an explosion/fire in the Balance of Plant 4 kV switchgear bus area. Prior to the Alert declaration, the operators initiated a manual SCRAM due to an unexpected power decrease from 88% to 40%. The licensee has visually verified that there is no ongoing fire and is investigating the initial cause of the event. Offsite power is available to the site, but EDGs 1 and 2 are running and supplying Unit 1 loads. The MSIVs shut and HPCI/RCIC are being used to maintain vessel level. At 1412 EST, NRC decided to remain in Normal Mode. At 1704 EST the licensee reported the following: At 1313 hours Eastern Standard Time (EST) a manual reactor scram was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. At this time, a Startup Auxiliary Transformer (SAT) experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. Emergency Diesel Generators (EDGs) 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The power interruption resulted in closure of the Main Steam Isolation Valves, per design. The manual scram also resulted in closure of Group 2, 6, and 6 Containment Isolation Valves. The Reactor Core Isolation Cooling (RCIC) system was manually started and is being used to control reactor water level. The High Pressure Coolant Injection (HPCI) system was manually started and is being used for pressure control. The Plant response to the event was per design. Unit 2 is not directly affected by the event, however, due to the shared electrical distribution system is in a Technical Specification Action statement due to the Inoperable Unit 1 SAT. The public health and safety is not impacted by this event. At 1751 EST, the licensee reported that the emergency declaration had been downgraded to an Unusual Event at 1730 because the plant no longer meets the criteria for an Alert, but does meet the criteria for an Unusual Event due to a "loss of all offsite power to Emergency 4 kV buses E1 (E3) and E2 (E4) for greater than or equal to 15 minutes." The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email). * * UPDATE FROM MARTY IRWIN TO DANIEL MILLS AT 1825 ON 2/07/16 * * * At 1814 EST the emergency declaration was terminated because offsite power was restored. The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified R2DO (Musser), NRR EO (Morris), IRD MOC (Stapleton), R2RA (Haney), NRR ET (Lubinski), NRR ET (Dean), DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
ENS 51680	24 January 2016 01:10:00	FitzPatrick	Manual Scram	NRC Region 1	GE-4	At 2241 (EST) on 1/23/2016, James A FitzPatrick inserted a manual scram from 89 percent power due to lowering intake level. Following the successful scram, a residual transfer occurred, resulting in a loss of the non-vital busses, loss of all Circulating Water Pumps, and a manual closure of the Main Steam Isolation Valves (MSIVs). The cause of the residual transfer is unknown. RPV (Reactor Pressure Vessel) level shrink during the scram resulted in a successful Group 2 isolation. Reactor Vessel (level) and pressure are being maintained with the High Pressure Coolant Injection System which was manually started. A cooldown is in progress. FitzPatrick will proceed to Mode 5 until the cause is identified and corrected. The Emergency Diesel generators auto started as a result of the loss of power to the non-vital busses. Offsite power remained available throughout the event. Operators are controlling pressure manually via the relief valves. FitzPatrick will notify the Public Service Commission of the event. The NRC Resident Inspector was notified.
ENS 51614	19 December 2015 10:38:00	Limerick	Automatic Scram	NRC Region 1	GE-4	At 0702 EST on 12/19/2015, the Unit 2 reactor automatically scrammed on a valid reactor low level signal (12.5 inches). The reactor low level signal was caused by the trip of the in-service 2A Reactor Feedwater Pump, causing reactor level to lower, exceeding the low reactor level setpoint of 12.5 inches. The shutdown was normal. The plant is stable in Hot Shutdown with normal reactor level, pressure control via the Main Steam Bypass Valves to the Main Condenser and normal level control using the Condensate System. High level trip of the 2A Reactor Feedwater Pump was caused by high reactor level of +54 inches following opening of bypass valves during reactor start-up and pressurization. The cause is still being investigated. All systems functioned as expected following the reactor scram. The licensee notified the NRC Resident Inspector.
ENS 51430	28 September 2015 22:49:00	Hope Creek	Automatic Scram	NRC Region 1	GE-4	On September 28, 2015 at 2046 EDT, the Hope Creek reactor scrammed following a trip of both reactor recirculation pumps. 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. The licensee notified Lower Alloways Creek township of the event.
ENS 51391	14 September 2015 02:46:00	Fermi	Manual Scram	NRC Region 3	GE-4	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. * * UPDATE AT 0555 EDT AT 09/14/15 FROM CHRIS ROBINSON TO JEFF HERRERA * * * 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). * * UPDATE PROVIDED BY CHRIS ROBINSON TO JEFF ROTTON AT 2135 EDT ON 09/14/2015 * * * 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) * * UPDATE FROM BRETT JEBBIA TO JOHN SHOEMAKER AT 1446 EST ON 2/27/16 * * * 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).
ENS 50981	13 April 2015 09:56:00	Limerick	Manual Scram	NRC Region 1	GE-4	Following a pre-planned scram for entry into the U2 Refueling Outage, the 'B' Reactor Protection system (RPS) was unable to be reset due to a deficiency with the '2H' Intermediate Range Power Monitor. Due to this failure to reset RPS, a manual full scram was initiated as required by plant procedures. The licensee notified the NRC Resident Inspector.
ENS 50973	11 April 2015 01:52:00	Susquehanna	Manual Scram Automatic Scram	NRC Region 1	GE-4	At 2346 EDT on April 10, 2015, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed due to a main turbine trip caused by loss of turbine steam seals and degrading main condenser vacuum. Unit 2 reactor was being shutdown for a refueling outage. At approximately 37 percent power, turbine steam seals were lost resulting in a degrading vacuum. The vacuum degraded quickly, resulting in a main turbine trip before the reactor operator could insert a manual scram. At 37 percent power, the turbine trip caused an automatic scram. This occurred during a transfer from normal steam seal supply to the auxiliary boiler supply. All control rods (fully) inserted. Reactor water level lowered to +2 inches causing Level 3 (+13 inches) isolation. No ECCS actuations occurred. The Operations crew subsequently maintained reactor water level at the normal operating band using RCIC (Reactor Core Isolation Cooling). No steam relief valves opened. The reactor recirculation pumps tripped on EOC-RPT due to the turbine trip at power. The reactor is currently stable in Mode 3. Investigation into the cause of the loss of turbine steam seals is underway. The NRC Resident Inspector was notified. A voluntary notification to PEMA (Pennsylvania Emergency Management Agency) and press release will occur. Unit 2 is in a normal shutdown electrical lineup. Turbine steam seals were restored to the normal steam supply and condenser vacuum was restored. Decay heat is being removed via the steam bypass valves to the condenser. Unit 2 is proceeding with their cooldown to support the scheduled refueling outage.
ENS 50903	19 March 2015 10:51:00	Fermi	Automatic Scram	NRC Region 3	GE-4	At 0702 EDT on March 19, 2015, Fermi 2 received an automatic scram due to actuation of the Reactor Protection System (RPS) function of Oscillation Power Range Monitor (OPRM) Upscale. The plant had recently transitioned to Single Loop Operation after securing the 'A' Reactor Recirculation Pump due to loss of normal and emergency cooling water supply. The lowest reactor water level was 134 inches above top of active fuel. Reactor water level is being maintained in the normal band by the Feedwater and Control Rod Drive Systems. No Safety Relief Valves (SRV) actuated. Reactor pressure is being maintained via the Main Turbine Bypass Valves and Main Condenser. Reactor Pressure Vessel Level 3 isolation occurred. No additional safety system actuations occurred. All off-site power sources were available throughout the event. The plant is currently in Mode 3 and in a stable condition. Investigation into the cause of the event is ongoing. This event is being reported under the four hour Non-Emergency reporting criteria of 10CFR50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified.
ENS 50847	24 February 2015 01:02:00	Limerick	Automatic Scram	NRC Region 1	GE-4	At 2140 EST on 02/23/2015, Unit 1 reactor automatically scrammed on a valid reactor high pressure signal (1096#). The reactor high pressure signal was caused by the closure of the 1C inboard main steam isolation valve (MSIV), causing reactor pressure to rise, exceeding the reactor protection system (RPS) setpoint of 1096# pressure. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the main steam bypass valves to the main condenser and normal level control using the feedwater system. The closure of the 1C inboard MSIV appears to have been caused by a loss of primary containment instrument gas (PCIG) pneumatic supply to the valve. Instrument air was aligned to the remaining MSIV's. Limerick Unit 1 will remain in Hot Shutdown until repairs can be made. All rods inserted into the core during the scram. No relief or safety valves actuated during the transient. The electric grid is stable and supplying all plant loads. There was no affect on Unit 2. The licensee notified the NRC Resident Inspector, Pennsylvania Emergency Management Agency, and Berks, Chester and Montgomery counties.
ENS 50404	26 August 2014 21:24:00	Browns Ferry	Automatic Scram	NRC Region 2	GE-4	At 1730 CDT on August 26, 2014, Browns Ferry Unit 1 experienced a turbine trip resulting in an automatic reactor scram. The cause of the turbine trip was a control valve fast closure signal that was generated by a turbine trip on generator neutral over voltage signal. The Main Steam Isolation Valves (MSIVs) remained 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 required components actuated as required with the exception of Standby Gas Treatment (SBGT) train A, which is under a clearance for planned maintenance. Neither High Pressure Coolant Injection (HPCI) nor Reactor Core Isolation Cooling (RCIC) initiation signals were received. Initially, three Main Steam Relief Valves (MSRVs) opened to control the pressure surge and subsequently reclosed. 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 also 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, and (2) General containment Isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).' The NRC Resident Inspector has been notified. Service Request 926468 was initiated in the Corrective Action Program. The plant is in its normal shutdown electrical lineup. The licensee is investigating the cause of the generator neutral overvoltage signal. There was no impact on units 2 and 3.
ENS 50090	6 May 2014 13:27:00	Browns Ferry	Automatic Scram	NRC Region 2	GE-4	At 0830 (CDT) on 05/06/2014, the Unit 3 reactor automatically scrammed due to low reactor water level as a result of a trip of both recirculation pumps. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. The Reactor Feedwater System controlled and maintained water level above the level 2 initiation setpoint. Prior to the Scram, the reactor was operating at 100% power. A Core and Containment Cooling Systems Analog Trip Unit Functional Test was in progress. The cause of the recirculation pump trip is under investigation. This event is reportable within 4 hours per 10CFR 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.' It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. U1 and U2 remained at 100% power and were unaffected. * * UPDATE AT 1302 EDT ON 05/09/14 FROM TODD BOHANAN TO DONG PARK * * * Investigation revealed that a failed power supply caused an Anticipated Transient Without Scram/Alternate Rod Insertion (ATWS/ARI) signal to be generated when a level 2 Reactor Water Level was simulated on one instrument. All systems responded to the ATWS/ARI signal as designed. This signal opened the Recirc Pump Trip breakers for both Recirculation Pumps and opened the ARI valves to bleed air from the Reactor Protection System (RPS) scram air header. The resulting transient caused reactor water level to dip below the RPS trip setpoint (level 3 Reactor Water Level), a normal plant response, and the automatic scram signal occurred. At the time of the RPS scram signal, all rods were inserting and reactor power was approximately 2-3% and lowering. The NRC Resident Inspector has been notified. Notified R2DO (Bonser).
ENS 49871	5 March 2014 02:26:00	Limerick	Manual Scram	NRC Region 1	GE-4	At 2334 EST on 3/4/14 Unit 1 was manually scammed during a Rapid Plant Shutdown. The Rapid Plant Shutdown was initiated due to an Electro Hydraulic (Control) (EHC) System failure resulting in all Low Pressure Turbine lntercept Valves failing closed. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Steam Bypass Valves to the Main Condenser and normal level control using Feedwater. The licensee informed both State and local agencies and the NRC Resident Inspector. A press release will be issued by the licensee.
ENS 49608	5 December 2013 05:40:00	Hope Creek	Automatic Scram	NRC Region 1	GE-4	While operating at 76% power on 12/5/13 at 0325 EST, the main turbine tripped on moisture separator hi level. The reactor scrammed along with the main turbine trip. All safety systems responded as designed and expected. There was no radiological release. There were no injuries. During the scram, all rods inserted into the core. Plant is stable in Mode 3 in its normal S/D (shutdown) electrical line up. Decay heat is being removed via the turbine bypass valves dumping steam to the main condenser. At 0505 EST while securing from cooldown in an attempt to start a recirc pump, BPVs (Bypass Valve) opened causing reactor level swell and subsequent shrink. During this time, RPV (Reactor Pressure Vessel) level lowered to below RPV level 3 and caused a RPS (Reactor Protection System) actuation. RPV level was recovered and is now stable in normal band. The licensee has notified the NRC Resident Inspector. * * UPDATE ON 12/5/13 AT 1000 EST FROM LINDSAY KOBERLEIN TO DONG PARK * * * This update to ENS #49608 adds reporting criterion 10CFR50.72(b)(3)(iv)(A) for the RPS actuation at 0505 EST during post-scram recovery. The licensee notified the NRC Resident Inspector and the Lower Alloways Creek township. The licensee will be making a press release. Notified R1DO (Cook).
ENS 49592	1 December 2013 10:02:00	Hope Creek	Automatic Scram	NRC Region 1	GE-4	While operating at 100% power on 12/01/2013, at 0613 EST, the main turbine tripped on moisture separator hi level. The reactor scrammed along with the main turbine trip. All safety systems responded as designed and expected. There was no radiological release. There were no injuries. During the scram, all rods inserted into the core. The plant is stable in mode 3 in its normal shutdown electrical line up. Decay heat is being removed via the turbine bypass valves dumping steam to the main condenser. The licensee notified the NRC Resident Inspector and will be notifying Lower Alloways Creek township.
ENS 49344	15 September 2013 11:52:00	Susquehanna	Manual Scram	NRC Region 1	GE-4	At 1123 EDT, (on 9/15/13), Susquehanna Unit 2 received a Division 2 RHR (Residual Heat Removal) room Flooded alarm. Plant operators reported 3 inches of water in the room with water coming from the 2B RHR pump suction relief valve. The Suppression Pool Isolation valve was closed to isolate the leaking relief valve from the Suppression Pool. This action stopped Suppression Pool level from lowering. The Leak is isolated at this time. Further water is being added to the room as the system piping is being drained. The cause of the relief valve lifting is unknown at this time and is under investigation. At the time of the event, Division 2 RHR system had been declared inoperable for unrelated equipment issues. At 1222 EDT, the licensee confirmed that leakage from the 2B RHR pump suction relief valve had stopped. Susquehanna Unit 2 remains stable and there was no impact on Susquehanna Unit 1. The licensee notified the NRC Resident Inspector, State, and Local authorities. Notified DHS SWO, FEMA, DHS NICC, and Nuclear SSA (email only). * * UPDATE FROM MARTIN LICHTNER TO JOHN SHOEMAKER AT 1641 EDT ON 9/15/13 * * * U2 SSES (Susquehanna Steam Electric Station) has exited the Unusual Event for Division 2 RHR room flooded as of 1552 EDT. The leak has been isolated, water removed from the room and Division 1 RHR is operating in shutdown cooling proceeding to Mode 4 (Cold Shutdown). A press release for this event was authorized at 1613 and issued at 1628 (on 9/15/13). The licensee has notified the NRC Resident Inspector. Notified R1DO (Cahill), NRR (Dorman), IRD (Gott), DHS SWO, FEMA, DHS NICC, and Nuclear SSA (email only).
ENS 49342	14 September 2013 05:15:00	Susquehanna	Manual Scram	NRC Region 1	GE-4	At approximately 0330 hours on September 14, 2013, Susquehanna Steam Electric Station Unit 2 reactor was manually scrammed while transitioning the 'A' reactor feed pump from flow control mode to discharge pressure mode. Reactor water level rose to +54 inches causing a trip of reactor feedpumps. Subsequently the mode switch was taken to shutdown to manually scram the Unit 2 reactor. All control rods inserted. Reactor water level lowered to approximately +18 inches. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. No containment isolations occurred. All safety systems operated as expected. RCIC system was manually initiated for level control until a reactor feedpump was recovered, then RCIC was manually shutdown. The cause of the feedwater flow transient and trip of the reactor feedwater pumps is under investigation. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a 4 hour report, and 10CFR50.72(b)(3)(iv)(A) for an 8 hour report. Decay heat is being removed via the turbine bypass valve to the condenser. Offsite power remains stable, and there was no impact on Unit 1. The licensee has notified the NRC Resident Inspector. The Pennsylvania Emergency Management Agency will be notified, and the licensee will be making a press release.
ENS 49108	12 June 2013 16:59:00	Hope Creek	Manual Scram	NRC Region 1	GE-4	This is a report of a manual RPS actuation and manual RCIC actuation per 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). At 1332 (EDT), on 6/12/13, the 'B' Circulating Water Pump tripped with a stuck open discharge valve resulting in a vacuum transient. Operators lowered reactor power from 100% in an effort to stabilize condenser vacuum. When vacuum reached 6.5 inches, the operators inserted a manual reactor scram at 1333 (EDT). All control rods inserted as required. No automatic ECCS or RCIC initiations occurred. No primary or secondary containment isolations occurred. The plant is stable in OP CON 3 HOT SHUTDOWN with the condensate pumps in service. The Reactor Recirculation Pumps are in service. At the time of the event, a RCIC surveillance was in progress, but did not contribute to the event. The RCIC pump was secured and subsequently placed in service for inventory control. The only safety-related equipment out of service at the time of the scram was the C Service Water Pump, which was tagged for scheduled maintenance. No personnel injuries occurred. No radiation releases occurred. The NRC Resident Inspector has been informed.
ENS 49099	7 June 2013 13:50:00	Susquehanna	Manual Scram	NRC Region 1	GE-4	At approximately 1203 (EDT) on June 7, 2013, Susquehanna Steam Electric Station Unit One reactor was manually scrammed during reactor startup. Pressure setpoint was being adjusted to the normal operating setpoint, from 750 psig to 934 psig, when all turbine bypass valves unexpectedly opened. Reactor Feed Pumps, Main Turbine, HPCI and RCIC tripped on the high level setpoint Level 8 (+54 inches) due to the resultant reactor level swell. The reactor operator then inserted a manual scram. All control rods inserted. Reactor water level lowered to approximately -10 inches causing Level 3 (+13 inches) isolations. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. All safety systems operated as expected. The cause of the Turbine Bypass valve opening is under investigation. Unit 2 was unaffected. The licensee notified the NRC Resident Inspector and will notify the Pennsylvania Emergency Management Agency.
ENS 48936	17 April 2013 02:11:00	Limerick	Manual Scram	NRC Region 1	GE-4	During outage main turbine stop valve RPS logic surveillance testing, an invalid RPS actuation occurred due to an error in executing main turbine surveillance testing procedures. A Turbine Stop Valve closure RPS signal occurred due to an error in the restoration sequence of restoring the RPS bypass signal and a subsequent manual trip of the main turbine. This resulted in a full scram and a trip of both reactor recirculation pumps. The site post-scram response procedure was entered, which required that the mode switch be placed in the locked SHUTDOWN position. This caused an expected but valid RPS actuation. No control rod motion occurred due to all control rods were inserted at the time of the invalid RPS actuation and subsequent valid RPS actuation. The license has notified the NRC Resident Inspector.
ENS 48829	19 March 2013 08:37:00	Browns Ferry	Manual Scram	NRC Region 2	GE-4	At 0402 (CDT) on 03/19/2013, the Unit 1 reactor was manually scrammed due to lowering main condenser vacuum. The cause of the loss of vacuum was a significant leak on the 1C feedwater heater level control line. The leak appeared as a steam/water leak near the penetration to the main condenser. As extraction steam was isolated, condenser vacuum deteriorated and was approaching the turbine trip setpoint, at which time the reactor was manually scrammed. Condenser vacuum recovered following the scram. MSIVs (Main Steam Isolation Valves) are open, main turbine bypass valves are controlling reactor pressure and reactor feedwater pumps are being used to control reactor water level. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated as required. In response to the scram, all plant equipment responded as designed. The reactor had been operating near 95% power for several hours due to the 1C3 heater isolating at 2334 (hrs. CDT) on 3/18/13. This event is reportable within 4 hours per 10CFR50.72(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.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). All rods inserted into the core during the scram. No safety relief valves lifted during the transient. The plant is in its normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector.
ENS 48782	25 February 2013 17:49:00	Browns Ferry	Automatic Scram	NRC Region 2	GE-4	At 1313 (CST) on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 (CST) with positive indication of isolation, and isolation valves deactivated at time 1514 (CST) to satisfy TS LCO 3.6.1.3 required actions. This event is reportable within 4 hours per 10CFR50.72(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'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). At 1415 (CST), Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 (CST), water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2.
ENS 48738	10 February 2013 09:11:00	Hatch	Manual Scram	NRC Region 2	GE-4	During normal power operations, the crew observed condensate/feedwater conductivity begin to increase at approximately 0530 EST on 02/10/13. The crew responded to the associated alarm response procedures and entered abnormal operating procedure 34AB-N61-001-1 due to degrading reactor water chemistry parameters. A power reduction (from 100%) was initiated at 0555 EST in accordance with station procedures for responding to a suspected condenser tube leak. At 0700 EST, a manual reactor SCRAM (from approximately 47%) was inserted due to the elevated reactor water conductivity in accordance with station abnormal operating procedures. All rods inserted completely and no complications were encountered following the reactor SCRAM, normal feedwater injection remained available. Following the SCRAM, a Group 2 Primary Containment Isolation Signal (PCIS) was received as a result of reactor water level lowering below +3 inches. The lowest reactor water level observed was (minus) 2 inches and was restored to normal operating levels utilizing normal feedwater injection. Following restoration of reactor water level to the normal operating level, the Group 2 PCIS signal was reset. No ECCS injection systems actuated as a result of the reactor SCRAM. The SCRAM was uncomplicated and the plant is stable. Decay heat removal is to the main condenser via the turbine bypass valves. The plant is in a normal offsite electrical power shutdown alignment. Efforts are in progress to isolate the condenser in-leakage. There was no impact on Unit 2. The licensee has notified the NRC Resident Inspector.
ENS 48623	22 December 2012 16:39:00	Browns Ferry	Automatic Scram	NRC Region 2	GE-4	On 12/22/2012 at 1152 CST, the Unit 2 reactor automatically scrammed due to actuation of the Reactor Protection System (RPS) from loss of power to RPS. At 1134 CST, the D 4kV Shutdown Board unexpectedly de-energized during performance of post-maintenance testing for the 3D Diesel Generator paralleling circuitry, resulting in loss of power to the 2B RPS subsystem. Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations were received along with automatic initiation of A, B, and C Standby Gas Treatment subsystems and A Control Room Emergency Ventilation subsystem due to loss of power to the 2B RPS subsystem. While attempting to reenergize the 2B RPS subsystem, the 2A RPS subsystem was inadvertently de-energized resulting in Unit 2 reactor automatic scram. All affected safety systems responded as expected for the loss of RPS and reactor scram. Due to the loss of RPS, the Main Steam Isolation Valves (MSIVs) closed. Reactor pressure did not rise to the automatic initiation set point for Safety Relief Valve (SRV) actuation. Reactor Core Isolation Cooling System (RCIC) and High Pressure Coolant Injection System (HPCI) reactor water level initiation set point of -45" was reached and RCIC and HPCI automatically initiated as designed to restore water level above the initiation set point. Both Recirculation Pumps also tripped on reactor water level of -45". Reactor pressure control was established by manually operating one SRV and water level control established with RCIC. HPCI was returned to standby readiness. The scram was reset, MSIVs were opened, and the Main Condenser was established as a heat sink. The scram event from critical is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B). It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector was notified. The 2A and 2B RPS subsystems were returned to service. The electrical grid is stable and supplying shutdown loads on Unit 2. Unit 1 and Unit 3 were unaffected and continue to operate at 100% power.
ENS 48607	19 December 2012 20:29:00	Susquehanna	Automatic Scram	NRC Region 1	GE-4	At approximately 17:31 hours on December 19, 2012, Susquehanna Steam Electric Station Unit Two reactor automatically scrammed on low RPV level (Level 3, +13 inches) while transitioning the 'A' reactor feed pump from discharge pressure mode to flow control mode. All control rods inserted and both reactor recirculation pumps tripped. Reactor water level lowered to approximately -29 inches causing Level 3 (+13 inches) isolations. An automatic trip of the reactor recirculation pumps occurred, but is not expected at this RPV level. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. All safety systems operated as expected. The cause of the loss of feed water flow and trip of the reactor recirculation pumps is under investigation. This report is being made per 10CFR50.72(b)(2)(iv)(B) for a 4 hour report, and 10CFR50.72(b)(3)(iv)(A) for an 8 hour report. Decay heat is removed via steam to the main condenser using the bypass valves . On-site electrical power is in the normal configuration. The Unit 2 reactor is currently stable in Mode 3. Unit 1 was not affected and operates at 99% power. The licensee will inform the Commonwealth of Pennsylvania and make a press release. The NRC Resident Inspector was notified.
ENS 48501	11 November 2012 06:08:00	FitzPatrick	Automatic Scram	NRC Region 1	GE-4	An unplanned, automatic reactor scram occurred at 0355 EST due to a Main Turbine trip signal. All safety systems operated and actuated as expected. Both the Main Transformer, T-1A and normal station services transformer T-4 activated their respective deluge systems. On-site fire brigade and offsite fire assistance have successfully extinguished the T-1A transformer fire. There is still an active fire in the T-1A bus ductwork. The plant will be taken to cold shutdown conditions. At 0545 EST the plant entered the emergency plan at the NUE level due to inability to successfully extinguish the fire. All control rods fully inserted following the reactor scram. MSIVs remain open with decay heat being removed via steam to the main condenser using the bypass valves. All electrical buses are powered from their normal offsite reserve source. The licensee notified the NRC Resident and appropriate State and local government agencies. Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email. * * UPDATE FROM JOHN WALKOWIAK TO DONALD NORWOOD AT 0642 EST ON 11/11/2012 * * * As of 0639 EST the fire in the T-1A bus ductwork has been extinguished. * * UPDATE FROM MARK HAWES TO DONALD NORWOOD AT 0747 EST ON 11/11/2012 * * * Local fire department is on-site. No radiological release and no protective actions required. Plant cooldown in progress. * * UPDATE FROM MARK HAWES TO DONALD NORWOOD AT 0810 EST ON 11/11/2012 * * * The Unusual Event (HU 6.1) has been terminated at 0801 EST. Cooldown in progress to cold condition. Reactor level at 206 inches and pressure is at 530 pounds. The licensee notified the NRC Resident and appropriate State and local government agencies. Notified R1DO (Dentel), NRR EO (McGinty), IRD (Gott). Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email.

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