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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56887 | 13 December 2023 07:02:00 | 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 phone and email: At 0102 CST, while operating at 100 percent (reactor) power, River Bend Station experienced an automatic reactor scram caused by a turbine trip signal. The cause of the turbine trip signal is not known at this time and is being investigated. At 0108, reactor core isolation cooling (RCIC) was initiated due to a loss of reactor feed pumps following feedwater heater string isolation. At 0114, reactor water level control was transferred back to feedwater and RCIC was secured. Reactor water level is being maintained by feedwater pumps and reactor pressure is being maintained by turbine bypass valves. The scram was uncomplicated and all other plant systems responded as designed. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) specified system actuation as result of expected post scram (reactor water) level 3 isolations and manual initiation of RCIC. No radiological releases have occurred due to this event from the unit. The NRC Senior Resident Inspector has been notified of this event. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the turbine trip, while still under investigation, was likely due to an electrical transient involving the main generator. Walkdowns in the switchyard post-scram identified damage to one of the output breaker disconnects. | Feedwater Reactor Protection System Reactor Core Isolation Cooling | |
| ENS 56863 | 18 November 2023 05:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram | The following information was provided by the licensee via phone and email: On November 17, 2023, at 2215 CST, River Bend Station (RBS) was operating at 30 percent reactor power performing plant startup activities when an isolation of low-pressure feedwater string `A' occurred. The team entered applicable alternate operating procedures and inserted control rods to exit the restricted region of the power to flow map. Feedwater temperature continued to lower until it challenged the prohibited region of the AOP-0007 graph requiring a reactor scram. The team inserted a manual reactor scram at 2355 from 24 percent reactor power. All control rods fully inserted and there were no complications. All systems responded as designed. Currently RBS Unit 1 is stable with reactor level being maintained 10 to 51 inches with feed and condensate, and pressure being maintained 500 to 1090 psig using steam drains. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) Specified System Actuation as result of Group 3 isolations. The NRC Senior Resident inspector has been notified. No radiological releases have occurred due to this event from the unit. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The electric plant is in a normal lineup for current plant conditions with all emergency diesel generators available. The cause of the initial isolation of low-pressure feedwater string "A" is still under investigation. | Feedwater Reactor Protection System Emergency Diesel Generator Control Rod | |
| ENS 55169 | 2 April 2021 15:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Turbine Trip | At 1017 CDT on April 2, 2021, while operating at 85 percent power, River Bend Station experienced an automatic reactor scram caused by a turbine trip signal. The cause of the turbine trip signal is not known at this time and is being investigated. Reactor water level is being maintained by feedwater pumps and reactor pressure is being maintained by turbine bypass valves. The scram was uncomplicated and all plant systems responded as designed. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) Specified System Actuation as result of expected post scram level 3 isolations. No radiological releases have occurred due to this event from the unit. The NRC Resident Inspector has been notified of this event. | Feedwater Reactor Protection System | |
| ENS 55154 | 25 March 2021 14:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Lowering Condenser Vacuum | On March 25, 2021 at 0901 CDT, River Bend Station Unit 1 (RBS) was operating at 93 (percent) reactor power (limited by 100 (percent) recirculation flow) when condenser vacuum began to lower due to ARC-AOV1A, Steam Jet Air Ejector Suction Valve, going closed. At 0918 CDT, a manual reactor SCRAM was inserted at approximately 80 (percent) reactor power due to condenser vacuum continuing to lower. After the SCRAM, all systems responded as designed and condenser vacuum was restored by starting a mechanical vacuum pump. The cause of the Steam Jet Air Ejector Suction Valve closure is unknown at this time and being investigated. Currently RBS is stable, and pressure is being maintained using Turbine Bypass Valves. The Main Steam Isolation Valves remained opened throughout the event. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72 (b)(3)(iv)(A) Specified System Actuation as result of expected post SCRAM level 3 isolations. No radiological releases have occurred due to this event from the unit. NRC Resident Inspector has been notified of this event. | Reactor Protection System Main Steam Isolation Valve Steam Jet Air Ejector | |
| ENS 54849 | 21 August 2020 14:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram | On August 21, 2020 at 0908 CDT, River Bend Station was operating at 100% reactor power when reactor recirculation pump 'B' tripped. At 0918 CDT, a manual reactor scram was inserted at 67% reactor power after receiving indications of thermal hydraulic instability as indicated by flux oscillations on the period based detection system (PBDS) and average power range monitors (APRMs). All control rods fully inserted and there were no complications. All systems responded as designed. Currently River Bend Station Unit 1 is stable and pressure is being maintained using turbine bypass valves. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical and 10 CFR 50.72 (b)(3)(iv)(A) Specified System Actuation as result of Group 3 isolations. NRC Resident Inspector has been briefed on this event. No radiological releases have occurred due to this event from the unit. | Reactor Protection System Reactor Recirculation Pump Control Rod | |
| ENS 54096 | 1 June 2019 04:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Low Reactor Water Level | At 2345 CDT at River Bend Station (RBS) Unit 1, a manual Reactor scram was inserted in anticipation of receiving an automatic Reactor Water Level 3 (9.7") scram due to the isolation of the 'B' Heater String with the 'A' Heater String already isolated. The 'B' heater string isolation caused loss of suction and subsequent trip of the running Feed Water Pumps 'A' and 'C'. All control rods fully inserted with no issues. Subsequently Reactor level was controlled by the Reactor Core Isolation Cooling (RCIC) system. Feed Water Pump 'C' was restored 4 minutes after the initial trip and the RCIC system secured. Currently RBS-1 is stable and is being cooled down using Turbine Bypass Valves. No radiological releases have occurred due to this event from the unit. The plant is currently under a normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector.
This amended event notification is being made to provide additional information that was not included in the original notification made on 6/1/19 at 0315 EDT. This event was reportable under 10 CFR 50.72(b)(3)(iv)(A) which was not annotated or described in the original report. Forty-two minutes after the Feed Water Pump 'C' was started, the pump tripped causing a Reactor Water Level 3 (9.7") RPS actuation. Feed Water was restored five minutes later using the Feed Water Pump 'A'. The NRC Resident Inspector has been notified. Notified the R4DO (Warnick). | Reactor Core Isolation Cooling Control Rod | |
| ENS 53732 | 10 November 2018 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram on High Reactor Pressure | At 0046 CST, River Bend Station experienced an automatic reactor scram on high reactor pressure. Initial indications are that the cause of the scram was an uncommanded closure of the #3 turbine control valve. The plant is stable with reactor water level in the normal level band of 10-51 inches being maintained with feedwater and condensate. Reactor pressure is in the prescribed band of 500-1090 psig, being maintained with turbine bypass valves and steam line drains. No injection systems were actuated either manually or automatically as a result of the event. The reactor scrammed on a Reactor Pressure High scram signal. A Reactor Level 3 signal resulted from the normal post-scram water level response. All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an automatic RPS actuation with the reactor critical. All control rods fully inserted. The Unit is in a normal shutdown electrical alignment. All control rods inserted properly and all systems functioned as designed. The licensee is investigating the cause of the event. The licensee notified the NRC resident inspector. | Feedwater Control Rod | |
| ENS 53192 | 1 February 2018 16:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram | At 1057 CST on February 1, 2018 with the unit in Mode 1 at approximately 27% power, a manual actuation of the Reactor Protection System (RPS) was initiated due to an unexpected trip of the B Recirc Pump with A Recirc Pump in fast speed. B Recirc Pump tripped during transfer from slow to fast speed resulting in single loop operation. Operators were unable to reconcile differing indications of core flow. This resulted in a conservative decision to initiate a manual scram. The cause of the B Recirc Pump trip and the apparent issues with core flow indication are under investigation. The plant is currently stable in Mode 3. The plant response to the scram was as expected. All control rods (fully) inserted as expected; the feedwater system is maintaining reactor vessel water level in the normal control band and reactor pressure is being maintained with steam line drains and main turbine bypass valves. The NRC Senior Resident (Inspector) has been notified.
This event was initially reported under 10 CFR 72(b)(2)(iv)(B) as a manual actuation of the RPS due to an unexpected trip of the B Reactor Recirculation Pump with the A Reactor Recirculation Pump running in fast speed (Single Loop Operations). Operations was unable to reconcile differing indications of core flow and made the conservative decision to perform a planned shutdown in accordance with normal operating procedures. Therefore, this event 'resulted from and was part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.73(a)(2)(iv)(A) and NUREG-1022 Section 3.2.6. Consequently, this event is not reportable as an actuation of RPS. The NRC Resident Inspector has been notified. R4DO (Groom) has been notified. | Feedwater Reactor Protection System Reactor Recirculation Pump Control Rod | |
| ENS 52915 | 19 August 2017 01:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Scram While at 100 Percent Power | At 2055 CDT on August 18, 2017, an automatic actuation of the reactor protection system occurred while the plant was operating at 100 percent power. No plant parameters requiring the actuation of the emergency diesel generators or the emergency core cooling system were exceeded. The main feedwater system remained in service following the scram to maintain reactor water level, and the main condenser remained available as the normal heat sink. The scram occurred after a planned swap of the main feedwater master controller channels in preparation for scheduled surveillance testing. When the channel swap was actuated, the feedwater regulating valves moved to the fully open position. The scram signal originated in the high-flux detection function of the average power range monitors, apparently from the rapid increase in feedwater flow. The cause of the apparent feedwater controller malfunction is under investigation. The NRC Resident Inspector has been notified. No safety relief valves opened. Decay heat is being removed via steam to the main condenser using the bypass valves and steam drains. The licensee intends to go to Cold Shutdown to investigate the malfunction. | Feedwater Reactor Protection System Emergency Diesel Generator Emergency Core Cooling System Safety Relief Valve Main Condenser | |
| ENS 52825 | 24 June 2017 01:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram on Main Generator Trip | While performing a scheduled generator voltage regulator test, River Bend Station experienced an automatic scram when the main generator tripped. It is unknown at this time why the main generator tripped. There were no equipment issues that materially impacted post scram operator response. The intention at this time is to go to cold shutdown while the cause of the trip is investigated. All rods inserted during the scram. Reactor water level is being maintained via normal feedwater with decay heat being removed via turbine bypass valves to the main condenser. The electrical grid is stable and supplying plant loads via the normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector. | Feedwater Main Condenser | |
| ENS 52602 | 10 March 2017 13:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Closure of the Main Turbine Control Valves | At 0714 CST on March 10, 2017, with the unit in Mode 1 at approximately 17% power, a manual actuation of the reactor protection system (RPS) was initiated due to rising reactor pressure caused by the closure of the Main Turbine Control Valves (MTCV's). The cause of the closure of the MTCV's is under investigation. The unit is currently stable in Mode 3. All control rods inserted as expected; water level control is stable in the normal control band and reactor pressure is being maintained with steam line drains (aligned to the main condenser). The NRC Senior Resident Inspector has been notified. | Reactor Protection System Control Rod | |
| ENS 51701 | 29 January 2016 21:18:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation After Loss of One Offsite Power Source | On January 29, 2016, at 1518 CST, with the plant in cold shutdown, power was lost on reserve station service (RSS) line no. 1. This is one of two sources of offsite power required by Technical Specifications. The power loss de-energized the Division 1 onsite AC safety-related switchgear, causing an automatic start of the Division 1 emergency diesel generator (EDG). The Division 1 reactor protection system (RPS) bus was also de-energized, causing a half-scram signal. Approximately 8 minutes later, a full actuation of the RPS occurred due to a high water level condition in the control rod drive hydraulic system scram discharge volume header. All reactor control rods were already fully inserted. The loss of Division 1 RPS also caused the actuation of the Division 1 primary containment isolation logic. The Division 1 isolation valves in the balance-of-plant systems closed as designed. Both trains of the standby gas treatment system actuated. The loss of RSS no. 1 occurred during post-modification testing on relays at the local 230kV switchyard. The exact cause of the event is under investigation. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). The unit remains in cold shutdown with 1 source of offsite power and all 3 (EDG) available. The (NRC) Resident Inspector has been notified. | Reactor Protection System Emergency Diesel Generator Primary containment Standby Gas Treatment System Control Rod | |
| ENS 51644 | 9 January 2016 08:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram on Main Steam Isolation Due to Electrical Fault | On 1/9/16 at 0237 (CST), River Bend Station sustained a reactor scram during a lightning storm. An electrical transient occurred resulting in a full main steam isolation (MSIV) (Group 6) and a Division II Balance of Plant isolation signal. During the scram, level 8 occurred immediately which tripped the feed pumps. A level 3 signal occurred also during the scram. Subsequent level 3 was received three times due to isolated vessel level control. The plant was stabilized and all spurious isolation signals reset, then the MSIVs were restored. The plant is now stable in Mode 3 and plant walkdowns are occurring to assess the transient. During the scram, all rods inserted into the core. The plant was initially cooled down using safety relief valves. Offsite power is available and the plant is in its normal shutdown electrical lineup. The licensee has notified the NRC Resident Inspector. | Safety Relief Valve Main Steam | |
| ENS 51568 | 27 November 2015 10:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Following Partial Loss of Offsite Power | At 0431 CST on November 27, 2015, an automatic reactor scram occurred following the trip of the main generator. The generator trip was apparently caused by a partial loss of offsite power, which resulted from a differential ground on the north bus of the local 230 kV switchyard. The ground signal caused the reserve station service line no. 1 to de-energize, which tripped the Division 1 offsite power source to station, as well as the main generator. The plant responded as designed as follows: The Division 1 emergency diesel generator started and tied to the bus restoring Division 1 emergency power. The Division 3 emergency diesel generator started and tied to the bus, restoring power on the Division 3 switchgear. The reactor protection system tripped as designed. Reactor water level was controlled normally with condensate and feed water. A level 3 reactor water level scram signal occurred as expected, and RPV (Reactor Pressure Vessel) water level was restored to normal level band. Reactor pressure was controlled by the bypass valve system, and a normal cool down was initiated. The reactor is being taken to cold shutdown pending an investigation of the event. The loss of power also resulted in a partial loss of normal service water cooling to the plant, and emergency service water cooling automatically initiated per design. At the time of event, the reactor protection system was aligned to the backup power supply, which was momentarily lost. This resulted in multiple system isolations including reactor water clean up, and outboard balance of plant isolations. These isolations were initiated due to loss of offsite power, and all responded as designed. The isolation resulted in a loss of the running decay heat removal pump for the spent fuel pool. The standby pump is available for service and being aligned for service. The plant is currently stable in hot shutdown. Transmission and distribution personnel are currently investigating the ground in the 230 kV switchyard. All control rods inserted. The licensee has notified the NRC Resident Inspector. | Service water Reactor Protection System Emergency Diesel Generator Decay Heat Removal Control Rod | |
| ENS 51112 | 2 June 2015 02:11:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to Low Reactor Water Level | At 2111 (CDT) River Bend Nuclear Station sustained an Automatic Reactor Scram due to low Reactor Water Level (Level 3). The plant is currently stable, with level being maintained in a normal band of 10 - 51 inches with Condensate and Feedwater. Reactor Pressure is in the prescribed band of 500-1090 psig. The plant is in Mode 3, Hot Shutdown, and will remain in Mode 3 until investigation of the scram is complete. The transient began with a trip of Reactor Feed Pump 'A', followed by a Reactor Scram and a trip of Reactor Feed Pump 'C'. Reactor water level was recovered with Reactor Feed Pump 'B' to a normal post scram level band. There was a problem noted with the Reactor Feedwater Master Level Controller; this was mitigated by the Operator placing the controller to manual. There was no subsequent Level transient. Reactor Pressure was stabilized in normal pressure band with Turbine bypass valves. During the transient, a Reactor Recirculating Flow Control Valve Runback was not received as expected. Reactor Recirculating Pump 'A' responded as expected to transient (switching to low pump speed), Reactor Recirculating Pump 'B' tripped during transient. A Level 3 isolation signal was received, all expected isolations occurred. The cause of the transient is currently under investigation. The reactor is stable in Mode 3 with decay heat being removed via turbine bypass valves, and a normal electrical line up. The NRC Resident Inspector has been notified.
At 2231 on 6/1/15, Reactor Water Cleanup System isolated on High Reactor Water Cleanup System Heat Exchanger room temperature due to loss of Turbine Building chill water during the initial transient. All Reactor Water Cleanup System Valves isolated as expected. Reactor Water Cleanup was the only system affected by this isolation signal. The licensee has notified the NRC Resident Inspector. Notified R4DO (Whitten). | Feedwater Reactor Water Cleanup | |
| ENS 50872 | 8 March 2015 03:40:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation Due to Loss of One Reserve Station Service Offsite Power Source | On March 7, 2015, at 2140 CST, with the plant in Mode 5, Refueling, the RSS#2, one of two Reserve Station Service offsite power sources, de-energized. This loss of RSS#2 caused the de-energization of the Division 2 Safety Bus, which caused a valid start signal to the logic of the Division 2 Emergency Diesel Generator. No start occurred, however, due to the diesel being in the maintenance mode. Division 2 Standby Service Water, which was being run for system fill and vent as well as surveillance testing, also lost power. At the time RSS#2 was lost, the Division 2 Diesel generator and Division 2 Standby Service Water were inoperable for the Refuel Outage 18 Division 2 maintenance window. Division 1 systems and RSS#1 were not affected by the power loss and continued to operate normally. There were 5 control building dampers and 1 floor drain air operated valve that repositioned due to the loss of power. The cause of RSS#2 de-energization is still under investigation. The licensee notified the NRC Resident Inspector. | Service water Emergency Diesel Generator | |
| ENS 50704 | 25 December 2014 14:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | At 0837 (CST) on 12/25/14, a loss of Reactor Protection System (RPS) 'B' occurred which resulted in a Division 2 RPS half SCRAM. This occurred concurrent with a Division 1 RPS half SCRAM which had been inserted for LCO 3.3.1.1 Action 'A' due to issues with the #2 turbine control valve RPS logic on 12/23/14. This resulted in a full RPS actuation and Reactor SCRAM. During the SCRAM, a reactor water Level '8' occurred which tripped the running reactor feed pump. Reactor water level peaked at 56 inches. This Level '8' is under investigation. Once reactor water level lowered below 51 inches the Level '8' signal was reset, and the team attempted to start the 'C' reactor feed water pump. The 'C' reactor feed pump failed to start upon attempt. The 'A' reactor feed pump was then started successfully. The startup feed regulating valve failed to open in automatic or manual mode, resulting in an RPV Level '3' signal (lowering to 8.1 inches). The operators manually aligned the 'C' feed water regulating valve and restored reactor water level to normal band. The plant is stable in Mode 3 pending investigation. The licensee notified the NRC Resident Inspector. | Reactor Protection System | |
| ENS 50546 | 17 October 2014 08:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram on High Average Power Range Monitor Flux | At 0303 (CDT), River Bend Nuclear Station sustained a reactor scram due to high Average Power Range Monitor (APRM) flux, suspected due to a malfunction of the Electrohydraulic Control System. Reactor recirculation pump 'B' tripped, reactor recirculation pump 'A' responded appropriately. All other systems responded appropriately except for loss of feed water due to low suction pressure trip from isolating the condensate demineralizers. Reactor water level did not get out of level band. Condensate demineralizers and feedwater were restored to service. Level 3 (isolation) was initiated due to scram. (One) system, Suppression Pool Cooling isolated accordingly due to level 3 signal. Currently the plant is in mode 3, hot shutdown. Plant will remain in mode 3 until investigation of scram is complete. During the scram, all rods inserted into the core. No relief valves lifted as a result of the transient. All safety equipment is available although reactor core isolation cooling is functional but inoperable due to an earlier issue discovered during a surveillance test. The reactor is at normal pressure and temperature for Mode 3. The cause of the high APRM flux and the identified anomalies are under investigation. The licensee has notified the NRC Resident Inspector.
The licensee is updating the notification to include an 8 hour notification for a specified system actuation due to the Level 3 isolation signal. Licensee is proceeding to cold shutdown to troubleshoot the EHC system. The licensee will notify the NRC Resident Inspector. Notified R4DO (Haire). | Feedwater Reactor Core Isolation Cooling Reactor Recirculation Pump | |
| ENS 47960 | 24 May 2012 18:48:00 | 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)(xiii), Loss of Emergency Preparedness | Manual Reactor Scram Due to a Loss of Feed as Result of a Loss of Switchgear | At 1348 CDT on 5/24/12 with the Reactor at 33% power, River Bend Station operators inserted a manual reactor scram based on loss of high pressure feed to the reactor following a loss of a 13.8 Kv switchgear. The Control Room team observed an electrical transient in the Control Room concurrent with the start of Reactor Feed Pump "B". The crew identified that no high pressure feed was aligned to the reactor and inserted a manual scram. Based on the configuration of the electrical plant during startup, all circulating water and Normal Service Water (NSW) was supplied from NPS-SWG1B. MSIVs were closed based on loss of circulating water and Standby Service Water (SSW) initiated automatically based on loss of NSW. EOP-0001, 'RPV Control' was entered on reactor high pressure and reactor low water level. EOP-0002, 'Primary Containment Control' was entered based on primary containment pressure high and suppression pool level high. EOP-0003, 'Secondary Containment Control', was entered on annulus pressure high. Reactor water level control is being maintained with Reactor Core Isolation Cooling (RCIC). High pressure core spray was manually started but was not required and was subsequently shut down. Pressure control is via RCIC and Safety Relief Valves (SRVs). Safety related busses are aligned to offsite power as normal. They were not affected by the electrical transient. Immediately after the scram at 1350, a report from the Turbine Building indicated smoke was seen around the Reactor Feed Pump 'B' termination cabinet. The Fire Brigade was activated. At 1358, the Fire Brigade reported that there was no fire. A review of the Emergency Action Levels (EALs) was performed. No emergency declaration was required. Initial investigation shows damage to cabling and circuit boards associated with Reactor Feedpump 'B' in the Turbine Building, but no fire was ever observed. In addition, the Technical Support Center (TSC) and Operations Support Center (OSC) lost power. At the time, both facilities continued to be in a state of readiness and emergency functions could be performed. At 1526, power was restored to both facilities, including the ventilation systems. All rods inserted into the core. The unit is stable at 230 psi and 391 degrees F. Reactor pressure is maintained by RCIC and decay heat removal via safety relief valves to the suppression pool. The unit is in a technical specification for suppression pool high level. There were no safety system failures. There is one non safety related 13.8 switchgear out of service due to this event and NNS-Switchgear 2A out of service from an event three days ago. Offsite assistance was not required. The NRC Resident Inspector has been notified. | Secondary containment Service water Reactor Core Isolation Cooling Primary containment High Pressure Core Spray Decay Heat Removal Safety Relief Valve | |
| ENS 47940 | 21 May 2012 19:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Resulting from a Low Condenser Vacuum-Turbine Trip | At 1452 CDT on 5/21/2012 with the reactor at 100% power, River Bend Station experienced a reactor scram resulting from an RPS actuation. Following the scram, reactor water level briefly lowered below level 3. The reactor is stable with pressure and temperature being controlled by the feed water system and main steam bypass valves, respectfully. The cause of the scram was due to a turbine trip/low condenser vacuum. The low vacuum condition resulted from a loss of non-safety related 4160V switchgear that powers two of four circulating water pumps. A suspected spurious Division II isolation of Reactor Core Isolation Cooling (RCIC) was observed. Restoration of RCIC to standby is in progress. A report of a fire in a manhole was received shortly after the scram . Fire Brigade was dispatched and noted a small active fire in a cable tray. The fire was extinguished with fire extinguishers. Power cables are routed through this manhole. The plant is conducting causal investigations to fully understand the cause of the turbine trip. As information becomes available, River Bend Station will provide additional information. All rods are inserted, and the plant is in a normal shutdown electrical lineup with the exception of NNS-Switchgear 2A being deenergized. Offsite assistance was not required to extinguish the fire. The licensee notified the NRC Resident Inspector. | Reactor Core Isolation Cooling Main Steam | |
| ENS 47549 | 23 December 2011 12:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to a Turbine Trip | At 0610 CST on 12/23/2011 with the reactor at 100% power, River Bend Station experienced an (automatic) reactor scram resulting from a RPS actuation. Following the scram, reactor water level briefly lowered below level 3, resulting in the automatic closure of containment isolation valves in the suppression pool cooling system. This isolation was confirmed to have occurred as designed. The reactor is stable with pressure and temperature being controlled by the feed water system and main steam bypass valves, respectively. The cause of the scram was due to a turbine trip. Initial indications are that the turbine tripped due to a loss speed sensor. All control rods inserted and Reactor Core Isolation Cooling was manually operated for approximately 1 minute and secured, The plant is conducting causal investigations to fully understand the cause of the turbine trip. As information becomes available River Bend Station will provide additional information. This event is being reported in accordance with 10CFR50.72(b)(iv)(B) as an automatic RPS actuation with the reactor critical. The safety relief valves momentarily lifted immediately following the scram. The plant electrical distribution system is in a normal shutdown configuration. The scram was uncomplicated. The licensee notified the NRC Resident Inspector. | Reactor Core Isolation Cooling Safety Relief Valve Control Rod Main Steam | |
| ENS 45388 | 29 September 2009 16:27:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Water from Upper Refuel Pool | On 9/29/2009, at 1127 hours, with the plant in Mode 5, Refuel, upper pool filled for refuel activities, and no fuel handling or control rod movement activities in progress, River Bend Station operators manually started Low Pressure Coolant Injection Pump Charlie to facilitate filling of the upper pool. This action was necessary to offset a loss of upper pool water through a main steam line and into the drywell. Approximately 5000 gallons of water drained from the upper pool into the drywell. Initial investigation found that the leakage was past a partially de-flated main steam line plug, through Safety Relief Valve flanges, which were not fully torqued at the time. The steam line plug became deflated during ECCS testing when service air isolated to containment. The service air system provides a backup to the mechanical seal. The drywell was evacuated and actions were initiated to close containment. The steam line plug was re-inflated which stopped leakage to the main steam line and drywell. At no time during the event was River Bend Station out of compliance with Technical Specification requirements for the upper pool. River Bend personnel are performing further investigations into the cause of the event. The licensee has notified the NRC Resident Inspector. | Main Steam Line Safety Relief Valve Control Rod Low Pressure Coolant Injection | |
| ENS 45369 | 20 September 2009 22:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Scram Due to Reactor Recirc Pump Trip | At 1747 hours on September 20, 2009, at River Bend Station, during plant shut down for a planned outage, an unplanned manual reactor scram was initiated by plant operators. As part of the planned shutdown, the reactor recirculation pumps were being transferred from fast to slow speed. This transfer did not occur as expected. Instead, the pumps tripped to off. After this occurred, the operators entered the manual reactor scram. Power level was approximately 23 percent at the time of the scram. All other plant equipment and systems performed as expected. Plant personnel are investigating the cause of the pump trip. The plant is proceeding with planned outage activities. All rods fully inserted. Decay heat is being removed via main steam drains and bypass valves to the condenser. Reactor pressure is at 200 psig. The electrical lineup is normal and all safety related equipment is available if required. No safety or relief valves lifted during the manual scram. The licensee has notified the NRC Resident Inspector. | Reactor Recirculation Pump Main Steam | |
| ENS 44034 | 5 March 2008 20:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Scram Due to Apparent Malfunction in Turbine Control System | At 2:43 p.m. CST, an unplanned automatic reactor scram occurred. The initial investigation determined that the event resulted from an apparent malfunction in the main turbine control system. This had the effect of causing reactor steam pressure to exceed the trip set point for the reactor protection system, initiating the reactor scram signal. All reactor control rods inserted, and the main feedwater system remained in service. Reactor safety-relief valves initially lifted as a result of the pressure transient. Reactor pressure is now being controlled by steam line drains with turbine bypass valves as a backup. There was no condition requiring the actuation of emergency diesel generators or emergency core cooling systems. NRC Resident Inspector has been notified. | Feedwater Reactor Protection System Emergency Diesel Generator Main Turbine Emergency Core Cooling System Control Rod | |
| ENS 43773 | 7 November 2007 09:06:00 | 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)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Scram as a Result of Loss of Normal Power 13.8 Kv Bus | With the plant in Mode 1 at (approximately) 75% power, the Auxiliary boiler and water treatment building 480 volt switchgear (NJS-SWG1J) faulted. The fault resulted in the loss of the NPS A bus (13.8 Kv normal supply), causing condensate and feed pumps to trip. Operators in the control room immediately responded and the plant was manually scrammed at 0306. Both the high pressure core spray (HPCS) and the reactor core isolation cooling (RCIC) systems responded automatically and injected into the vessel (valid ECCS signal). Safety systems responded as expected, including level 2 isolations. The licensee believes a transformer fault may have transferred up the line and caused the loss of normal power supply. RCIC is controlling reactor water level with primary plant pressure approximately 325 psia. Decay heat is being controlled through modulating the SRV's. The licensee has all systems available to place the unit in safe shutdown and cooldown. The licensee has one inoperable EDG and is not in any technical specification action statement at this time. The licensee notified the NRC Resident Inspector.
The high pressure core spray system was returned to its standby lineup at 0318 (all times are CST). Standby service water was being placed in service at 0701 to raise service water header pressure when standby service water pump 'C' started automatically. NPS 13.8kv switchgear 'A' was restored to service at 1245. The reactor core isolation cooling system, which automatically started at the time of the event, was shutdown at 1645. The Division 3 diesel generator, which automatically started at the time of the event, was restored to its standby lineup at 1429. Shutdown cooling was placed in service with residual heat removal pump 'A', at 1626. The plant entered Mode 4 (cold shutdown) at 1942. The electrical fault that initiated the event has been isolated to a 13.8kv/480v transformer in the turbine building. An investigation is ongoing. The licensee notified the NRC Resident Inspector. Notified the R4DO (Spitzberg) and NRR (Lubinski). | Service water Reactor Core Isolation Cooling Shutdown Cooling High Pressure Core Spray Residual Heat Removal | |
| ENS 43668 | 27 September 2007 03:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram During the Performance of Aprm Surveillance Testing | At approximately 2244 on September 26, 2007 an unplanned reactor scram occurred. An Average Power Range Monitor (APRM) surveillance was being performed at the time. This surveillance should not have caused a full scram and the cause of the scram is under investigation. The plant responded to the scram as expected, all control rods fully inserted, and there is no need for emergency injection system operation. Critical parameters are being maintained within prescribed bands and the plant has been stabilized. A level 3 signal was received during this scram which would have caused an isolation but all affected valves were already in the closed position. This event is being reported in accordance with 10CFR50.72(b)(2)(iv)(B) as an event that results in activation of the reactor protection system with the reactor critical. All emergency systems are available and in standby mode. The licensee notified the NRC Resident Inspector. | Reactor Protection System Control Rod | |
| ENS 43345 | 4 May 2007 17:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Loss of Cooling to Number 2 Main Transformer | At approximately 1256 CDT on May 4, 2007, a manual reactor scram was initiated following the loss of cooling to the no. 2 main transformer. Reactor power at the time of the scram was approximately 70 percent (initially 100% power). Following the scram, reactor water level briefly decreased below Level 3, resulting in the automatic closure of containment isolation valves in the suppression pool cooling system. This isolation was confirmed to have occurred as designed. Reactor pressure and water level control were promptly established. All control rods inserted, and no emergency injection system operation was required. This event is being reported in accordance with 10CFR50.72(b)(2)(iv)(B) as a condition resulting in a manual actuation of the reactor protection system. The NRC Resident Inspector was notified of this event by the licensee. | Reactor Protection System Main Transformer Control Rod | |
| ENS 42921 | 19 October 2006 22:56:00 | 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 Scram Following Spontaneous Feedwater Valve Closure | At 1756 CDT with the plant operating at 100% power, a reactor scram occurred in response to a reactor water level 3 signal from an apparent loss of feedwater. Both feedwater injection lines isolated when isolation valves were inadvertently closed, The cause of the isolation valve closure is under investigation. When reactor water level lowered to level 2, high pressure core spray (HPCS) initiated automatically and recovered water level. The reactor core isolation cooling system (RCIC) was tagged out for maintenance at the time of the event. Following the scram, main steam isolation valves isolated on low main steam header pressure. As a result, reactor pressure control was being controlled with the safety relief valves. SRV pressure control in turn led to EOP entry conditions on containment pressure and suppression pool level. Both feedwater lines were opened, and normal reactor level control was restored. The MSIV's were opened and pressure control was returned to the turbine bypass valves and the main condenser. Initial indications are that all plant equipment functioned as designed with the exception of the 'B' feed pump which experienced an apparent seal failure. The plant is stable in Mode 3. All plant conditions are understood. This event is being reported in accordance with 10CRF50.72(b)(2) as an RPS actuation and an injection of HPCS into the reactor vessel, and in accordance with 10CRF50.72(b)(3) as a loss of safety function of HPCS, as it was manually disabled during recovery from the event. The HPCS Injection valve was manually overridden closed for 76 minutes. In addition,' containment isolation valves in multiple systems actuated in response to the RPV level 2 signal. Reactor vessel water level lowered to below level 2. Decay heat is being removed by normal feedwater to the reactor vessel steaming to the main condenser. Offsite power is available and stable. Emergency Diesel Generators are available. The licensee notified the NRC Resident Inspector.
The closure signal to the main feedwater header isolation valves occurred when part of a chart recorder above the isolation valve control switches was dropped by an operator. The operator was attempting to adjust the paper drive mechanism in the recorder, and accidentally dropped the paper cartridge, which struck the 'CLOSE' pushbuttons on the isolation valve control switches. Following the scram, there was a delay in placing the reactor mode switch in the 'SHUTDOWN' position, which is an immediate action required by procedure. Placing the mode switch to 'SHUTDOWN' bypasses the reactor low steam pressure MSIV Isolation. Reactor steam pressure began dropping after the scram, until it reached the MSIV automatic closure setpoint, and the MSIVs isolated, In addition the licensee corrected one of the 10 CFR Section entries from "50.72(b)(3)(v)(A) POT UNABLE TO SAFE SD" to "50.72(b)(3)(v)(D) ACCIDENT MITIGATION." The licensee notified the NRC Resident Inspector. Notified R4 DO (D. Powers) and NRR EO (N. Chokshi) | Feedwater Emergency Diesel Generator Main Steam Isolation Valve Reactor Core Isolation Cooling High Pressure Core Spray Safety Relief Valve Main Condenser Main Steam | |
| ENS 42505 | 15 April 2006 21:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram from 100% Power | At 1615 CDT time on April 15th, 2006 with the Reactor at 100% power, River Bend Station experienced a reactor scram resulting from a RPS actuation. All plant systems performed as designed. The reactor is stable with pressure and temperature being controlled by feed water system and main steam bypass valves, respectively. The cause of the scram is not known at this time. The plant is conducting causal investigations to fully understand the cause of the plant scram. All control rods fully inserted, no safety valves lifted, and no Emergency Safety Feature actuations occurred on the scram. The electrical grid is stable. The licensee notified the NRC Resident Inspector.
River Bend Station status on 04/16/2006: The unit is in Mode 3, Hot Shutdown, and stable. Investigation of the plant trip found an optical isolator in the control circuit for the end-of-cycle recirculation pump trip failed causing both recirculation pumps to shift to slow speed. The pump downshift placed reactor power and flow into the exclusion region of the power to flow map. The reactor received trip signals from APRM E & H heat flux exceeding their corresponding trip set points. The optical isolator is being replaced. Plant restart will commence pending successful replacement of the optical isolator. The licensee informed the NRC Resident Inspector. Notified R4DO (Pruett). | Control Rod Main Steam | |
| ENS 41339 | 15 January 2005 08:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Receiving a Main Generator Alarm | At 0212 plant operators inserted a manual reactor scram and tripped the main turbine while at 100% power due to indications and alarms of a generator field ground fault. All rods fully inserted and there were no indications of any relief valves lifting. The reactor water level is being controlled within the normal band with feed pump "C" and reactor pressure is being controlled with bypass valves at 800-1090 psig. Also the "A" heater drain pump tripped during the reactor scram due to the water level in the third port heater. The licensee is investigating the cause. The NRC Resident Inspector was notified. | Main Turbine | |
| ENS 41252 | 10 December 2004 19:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram Due to Loss of Vital Instrument Bus | At 1317 (hrs CST) on December 10, 2004, an automatic actuation of the reactor protection system (RPS) occurred resulting in a reactor scram. The apparent cause of the event was loss of a vital instrument bus due to a fault in a nonsafety related vital inverter. This inverter provides power to selected control room instrumentation and controls. This resulted in the loss of feed water level control. Reactor level is being maintained by the High Pressure Core Spray System. The feed water system is not available. Reactor pressure is being controlled through the main turbine steam bypass system to the condenser. The condenser is available and being used as the heat sink. The residual heat removal system was operated in suppression pool cooling mode to provide a means of rejecting water from the suppression pool (water input from High Pressure Core Spray System minimum flow line). The plant is currently stable, and being maintained in hot shutdown. Systems responded as expected based on the initiating event. Reactor Core Isolation Cooling is not being used pending evaluation of a system alarm that is currently being investigated. Investigation of the initiating fault is being pursued in order to recover the vital bus and feed water level control. It has been preliminarily determined that the loss of instrument power resulted in the Main Feedwater regulating valve failing as-is and the "B" Reactor Recirculation Pump shifting down in speed. The reduction in reactor power with constant feed flow resulted in a high reactor vessel water level, producing a direct reactor scram signal at the High Level 8 setpoint. The licensee has notified the NRC Resident Inspector.
At 1657 CST (on 12/10/04), reactor level control was restored to the normal Feedwater and Condensate Systems. The High Pressure Core Spray System was restored to the normal standby lineup. Investigation into the cause of the reported RPS actuation continues. Investigation into the Reactor Core Isolation Cooling System alarms has resulted in declaring this system inoperable. The licensee will notify the NRC Resident Inspector. Notified R4 DO (L. Smith), NRR EO (M. Tschiltz) and IRD Manager (S. Frant)
The final determination of the cause of the scram was determined to be due to the B recirc pump downshift and subsequent power to flow scram on APRM flux. The licensee will notify the NRC Resident Inspector. Notified R4DO (L Smith)
The Reactor Core Cooling System was returned to available status at 0343 on 12/11/2004 and was restored to operable status at 2200 on 12/11/2004." The licensee notified the NRC Resident Inspector. Notified R4DO (Kriss Kennedy). | Feedwater Reactor Protection System Reactor Core Isolation Cooling Main Turbine Reactor Recirculation Pump High Pressure Core Spray Residual Heat Removal | |
| ENS 41165 | 1 November 2004 22:44:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Automatic Start | At 1644 (11/01/04), with River Bend Station in Mode 5 (Refueling), voltage on offsite power line Reserve Station Service (RSS) No. 2 was lost. This offsite power line is the 230 KV power supply to the Division 2, 4160 volt safety related electrical bus. The Division 2 emergency diesel generator (EDG) automatically started and loaded on a loss of voltage to the 4160 volt safety bus. At the time of the event, Division 2 Residual Heat Removal (RHR) was in the fuel pool cooling assist mode and the pump secured due to the loss of power. It was subsequently restarted and continues to provide the shutdown cooling function. The following systems were isolated as a result of this event: Reactor Water Clean-up, Alternate Decay Heat Removal, and Floor Drains. Investigation of the cause of the event is ongoing. There was no impact on spent fuel pool level or temperature. At the event reporting time, power to the bus was still being provided by the Division 2 EDG with plans to restore normal power shortly. The licensee informed the NRC Resident Inspector. See also similar event EN # 41164 of 10/31/04. | Emergency Diesel Generator Shutdown Cooling Residual Heat Removal Decay Heat Removal | |
| ENS 41164 | 1 November 2004 03:56:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Plant Had Auto Start of the Division 1 Emergency Diesel Generator | Undervoltage conditions were experienced on Division 1 emergency switchgear with subsequent start and load of the respective diesel generator (EDG). A preliminary investigation indicates that an unexpected undervoltage signal was generated when technicians inadvertently contacted the wrong terminals during preparations for the respective division ECCS surveillance testing. All systems and equipment responded as required. Power is still being provided (as of event reporting time) by the Division 1 EDG while they are working to restore normal power. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 41082 | 1 October 2004 12:30:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Trip Due to Main Generator Load Rejection | At 7:30 a.m. (CDT) on October 1, 2004, an automatic reactor scram occurred as a result of an electrical fault on the main generator output lines that caused a main generator trip and turbine trip. All control rods inserted. Approximately 13 minutes prior to the fault, a loss of one station service transformer had occurred. This resulted in an automatic start of the Division 1 diesel generator and a loss of power to some plant auxiliaries, including the feedwater level regulation isolation valves. The loss of reserve station service no. 1, combined with the trip of the main generator, caused a loss of power to two condensate pumps and one main feedwater pump. The remaining two feedwater pumps tripped on low suction pressure. The reactor containment isolation cooling pump (RCIC) steam supply isolated during the scram transient, so the control room operators manually started the high-pressure core spray system (HPCS) pump for level control. The injection valve was closed as level had already reached the high water level isolation setpoint for that valve. It was later reopened manually as level approached the low level setpoint (level 2), which would have automatically opened the valve. The level 2 setpoint was reached briefly after the valve was already open. Reactor pressure is being controlled manually with safety relief valves (SRV's). The main steam isolation valves (MSIV's) were manually closed due to lowering pressure from steam loads in the plant that could not be immediately isolated because of loss of power to their valves. RCIC is also running in CST (condensate storage tank) to CST mode, to augment pressure control. The electrical load center which supplies power to the instrumentation and valves needed for feedwater operation was cross-tied to an alternate power source, and feedwater was restored to operation and is presently controlling reactor water level. During the event, standby service water also initiated. This is the presently known information. Further information will be provided as the investigation continues. The licensee has notified the NRC Resident Inspector. | Feedwater Service water Main Steam Isolation Valve Core Spray Safety Relief Valve Control Rod | |
| ENS 40957 | 15 August 2004 09:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Following Partial Loss of Offsite Power | At 0405 on August 15, 2004, River Bend Station had a loss of 230Kv Line #353 (RSS #2, Port Hudson) feeding the plant's switchyard. This resulted in a generator trip and a reactor scram. Feedwater tripped on a Reactor vessel Level 8 signal. RCIC is operating to control reactor vessel level. SRVs were used to manually control pressure. Mechanical vacuum pump (ARC A) would not start to control condenser vacuum. The MSIVs were manually isolated in anticipation of a Group 6 isolation (low vacuum). There was no ECCS initiation. Power was restored to the mechanical vacuum pump ARC B. Vacuum was restored and main steam lanes are in use for pressure control. We are still investigating the root cause. The following ESF actuations occurred: - Division II Diesel Generator - Standby Gas Treatment System - Annulus Mixing System - Control Building Ventilation System. The licensee will inform the NRC Resident Inspector.
River Bend Station is currently in Mode 3, Hot Shutdown. During the event, offsite power was maintained through Reserve Station Service Line #1 to Division 1 safety related equipment and 'A' balance of plant (BOP) non-safety related equipment. The 230 KV line from the plant's switchyard (RSS Line #2) that was lost during the event has been recovered and the Division 2 bus has been restored to its normal power supply. The Division 2 diesel generator has been secured from operation (and has been placed in standby) During, the event, main condenser vacuum was restored from the main control room by starting the 'B' mechanical vacuum pump. Feedwater Pump 'A' remained available during the event. The circulating water supply to the main condenser remained in operation throughout the event. RCIC (Reactor Core Isolation Cooling) remains in service for reactor vessel level control. The ESF (Emergency Safety Feature) systems have been secured and placed in standby. A root cause team has been assembled to address the cause of the initial failure of the 230 KV Line #353. The licensee has notified the NRC Resident Inspector Notified R4DO (Linda Smith) and NRR EO (Cynthia Carpenter) | Feedwater Standby Gas Treatment System Main Condenser Main Steam | |
| ENS 40191 | 23 September 2003 03:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Received During Main Turbine Control Valve Testing | At 2243 on 9/22/2003 a Reactor scram occurred while performing Main Turbine control valve testing. Preliminary investigations indicate that high Reactor pressure was the signal that caused the scram. As a result of the scram, a Reactor low level 3 (9.7 inch) signal was received as expected from a high power scram due to the rapid coolant shrinkage. All expected actions for the low reactor level 3 (9.7 inch) signal occurred as expected. The plant is stable in Mode 3 (hot shutdown) with level being controlled with the Feedwater System and Reactor pressure being controlled with the Main Turbine Bypass valves. All control rods fully inserted, no ECCS actuations were received, and the electrical grid is stable. The licensee notified the NRC Resident Inspector. | Feedwater Control Rod |