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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56894 | 16 December 2023 09:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Scram Due to Turbine Trip | The following information was provided by the licensee via email: On December 16, 2023, at 0350 CST, Grand Gulf Nuclear Station was operating in mode 1 at 81 percent power when an automatic scram occurred due to a turbine trip signal. Before the scram the unit was performing a rod sequence exchange, and no critical work was underway. The cause of the turbine trip signal is not known at this time and is being investigated. All control rods fully inserted, there were no complications, and all plant systems responded as designed. Reactor water level is being maintained by main feedwater and condensate. Reactor pressure is being maintained with main turbine bypass valves. No radiological releases have occurred due to this event. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of the reactor protection system when the reactor is critical and specified system actuation due to expected reactor water level 3 isolation signals on a reactor scram. 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: Group 2 and Group 3 isolations occurred on the Level 3 isolation signal. | Feedwater Reactor Protection System Control Rod | |
| ENS 56282 | 20 December 2022 03:01:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram Due to Loss of Feedwater Pump | The following information was provided by the licensee via email: At 2101 (CST) on December 19, 2022, a manual reactor scram was initiated at Grand Gulf Nuclear Station (GGNS). Following the reactor scram, the high pressure core spray (HPCS) system was used to maintain reactor water level. The manual (reactor protection system) RPS actuation is being reported in accordance with 10 CFR 50.72(b)(2) and the HPCS actuation is being reported in accordance with 10 CFR 50.72(b)(3). At 2058, GGNS experienced a loss of a condensate booster pump. At 2101, the `A' reactor feedwater pump tripped and the reactor was manually scrammed. All control rods were fully inserted into the core. At 2104, the `B' reactor feedwater pump tripped and HPCS was manually started. HPCS was manually injected to maintain reactor water level at 2121. The `A' reactor feedwater pump was successfully restarted at 2126. GGNS is currently in Mode 3. Reactor level is being maintained with the `A' reactor feedwater pump and pressure is being maintained with the turbine bypass valves. The NRC Resident Inspector was notified. | Feedwater High Pressure Core Spray Control Rod | |
| ENS 55973 | 30 June 2022 19:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram Due to Loss of Transformer | The following information was provided by the licensee via phone and email: At 1445 (CDT) on June 30, 2022, with Grand Gulf Nuclear Station in Mode 1 and at 100 percent power, the reactor was manually scrammed due to the loss of balance of plant (BOP) transformer 23. All control rods fully inserted into the core and all systems responded appropriately. Reactor level is being maintained with condensate and feedwater. Reactor pressure is being maintained with turbine bypass valves. The cause of the loss of BOP transformer 23 is under investigation at this time. Standby Service Water 'A' and 'B' were manually initiated to supply cooling to Control Room A/C, ESF switchgear room coolers, and plant auxiliary loads. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B) as an event or condition that resulted in actuation of the Reactor Protection System and 10 CFR 50.72(b)(3)(iv)(A) due to the actuation of Standby Service Water. The NRC Senior Resident Inspector was notified. | Feedwater Service water Reactor Protection System Control Rod | |
| ENS 55030 | 11 December 2020 18:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Main Turbine / Generator Trip | On December 11, 2020 at 1204 CST, Grand Gulf Nuclear Station (GGNS) experienced an Automatic Reactor Scram from 100 percent Reactor Power after a Main Turbine and Generator Trip. All Control Rods fully inserted and there were no complications. All systems responded as designed. Reactor pressure is being maintained with Main Turbine Bypass Valves. Reactor water level is being maintained in normal band with the condensate system. No radiological releases have occurred due to this event from the unit. The NRC Branch Chief has been notified. | Main Turbine Control Rod | |
| ENS 54986 | 6 November 2020 08:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Turbine/Generator Trip | On November 6, 2020, at 0239 CST, Grand Gulf Nuclear Station (GGNS) experienced an Automatic Reactor Scram from 84 percent Reactor Power after a Main Turbine and Generator Trip. All control rods fully inserted and there were no complications. All systems responded as designed. Reactor pressure is being maintained with Main Turbine Bypass Valves. Reactor water level is being maintained in normal band with the condensate system. No radiological releases have occurred due to this event from the unit. The NRC Resident has been notified. | Main Turbine Control Rod | |
| ENS 54855 | 25 August 2020 04:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram After Loss of Feedwater - Grand Gulf | On August 24, 2020 at 2305 CT at Grand Gulf Nuclear Station (GGNS) an Automatic Reactor Scram occurred after a trip of the Reactor Feed Pump B and subsequent lowering of reactor water level to 11.4 inches Narrow Range. The scram occurred with Reactor Power at 14% and the main generator offline. All control rods fully inserted and there were no complications. All systems responded as designed. Main Steam Isolation Valves were manually closed to control reactor cooldown, Currently GGNS reactor pressure is being maintained at 450-600psig. Reactor water level is being maintained with condensate through startup level control. No radiological releases have occurred due to this event from the unit. The NRC Resident has been notified. Decay heat is being removed via the main condenser. Notified R4DO. | Feedwater Main Steam Isolation Valve Main Condenser Control Rod | |
| ENS 54824 | 8 August 2020 06:27:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram Due to a Turbine High Pressure Control Valve Malfunction | On August 8, 2020, at 0127 CDT, Grand Gulf Nuclear Station was manually shut down due to a turbine high pressure control valve malfunction. Reactor pressure is being controlled with bypass control valves to the main condenser. Reactor level is being maintained with condensate and feedwater through startup level control. The plant is stable in MODE 3 and proceeding to cold shutdown. The cause of the 'D' high pressure control valve malfunction is under investigation at this time. All rods fully inserted and there were no complications. All 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. Additionally, at 0159 CDT, with all rods fully inserted and after the 0127 CDT manual reactor Scram, an automatic valid RPS actuation signal was received. This event is also being reported under 10 CFR 50.72(b)(3)(iv)(A), as an event or condition that results in a valid actuation of any of the systems listed in 10 CFR 50.72(b)(3)(iv)(B). | Feedwater Reactor Protection System Main Condenser | |
| ENS 54725 | 25 May 2020 09:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Main Turbine Trip | An (automatic) reactor SCRAM occurred at 0433 CDT, on 05/25/2020, from 66 percent core thermal power. The cause of the SCRAM was due to a Main Turbine Trip. The cause of the Turbine Trip is under investigation. All systems responded as designed. No loss of offsite power or (Emergency Safety Feature) (ESF) power occurred. No (Emergency Core Cooling System) (ECCS) or Emergency Diesel Generator initiations occurred. Main Steam Isolation valves remained open and no radioactive release occurred due to this event. The plant is stable in mode 3. 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. The NRC Resident Inspector has been notified. Decay heat removal is through the Feedwater and Condensate System. | Feedwater Reactor Protection System Emergency Diesel Generator Main Steam Isolation Valve Decay Heat Removal | |
| ENS 54062 | 12 May 2019 15:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Reactor Scram Due to Partial Loss of Service Water | At 1039 CDT the reactor was manually (scrammed) due to a partial loss of plant service water. The loss of plant service water was caused by a loss of (balance of plant) BOP transformer 23. Reactor power was reduced in an attempt to restore pressure to plant service water. Reactor level is being maintained with condensate and feedwater. Reactor pressure is being maintained with bypass control valves. Standby Service Water A and B were manually initiated to supply cooling to Control Room A/C and (Engineered Safety Feature) ESF switchgear room coolers. The cause is under investigation. The NRC Resident Inspector has been notified. 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 also reported under 10 CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of RPS and Standby Service Water. The plant is currently in a normal electrical lineup.
This is an update to the original notification. The Drywell and Containment exceeded the technical specification (TS) temperature limits of 135 degrees F (TS Limiting Condition of Operation (LCO) 3.6.5.5) and 95 degrees F (TS LCO 3.6.1.5), respectively. An 8-hour notification is being added for any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material per 10 CFR 50.72(b)(3)(v)(C). Notified R4DO (Alexander). | Feedwater Service water Reactor Protection System | |
| ENS 53894 | 23 February 2019 20:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram After Turbine Control Valve Fast Closure | Actuation of RPS (Reactor Protection System) with the reactor critical. Reactor scram occurred at 1458 (CST) on 2/23/2019 from 100% power. The cause of the scram was due to Turbine Control Valve Fast Closure. All control rods are fully inserted. Currently reactor water level is being maintained by the Condensate Feedwater System in normal band and reactor pressure is being controlled via Main Turbine Bypass valves to the main condenser. No ECCS (Emergency Core Cooling System) initiation signals were reached and no ECCS or Diesel Generator initiation occurred. The Low-Low Set function of the Safety Relief Valves actuated upon turbine trip. This was reset when pressure was established on main turbine bypass valves. The cause of the turbine trip is still under investigation. There were no complications with scram response. The licensee notified the NRC Resident Inspector. There was no maintenance occurring on the main turbine at the time of the scram. | Feedwater Main Turbine Safety Relief Valve Main Condenser Control Rod | |
| ENS 53788 | 12 December 2018 06:00:00 | 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 | En Revision Imported Date 12/17/2018 | EN Revision Text: MANUAL REACTOR SCRAM DUE TO FAILED OPEN TURBINE BYPASS VALVE At 1351 CST, the reactor was manually shutdown due to 'A' Turbine Bypass Valve opening. The Main Steam Line Isolation Valves were manually closed to facilitate reactor pressure control. Reactor level is being maintained through the use of Reactor Core Isolation Cooling System, Control Rod Drive System, and High Pressure Core Spray System. High Pressure Core Spray System was manually started to initially support reactor water level control. Reactor Pressure is being controlled through the use of the Safety Relief Valves and the Reactor Core Isolation Cooling System. The plant is stable in MODE 3. The cause of the 'A' Turbine Bypass Valve opening is under investigation at this time. The NRC Resident Inspector has been notified.
This is an update to EN # 53788 to correct an error on the event classification block of the form. The original notification did not have the block for 8 hour notification for Specified System Actuation checked. The actuation of Reactor Core Isolation Cooling System was discussed in original notification. The licensee notified the NRC Resident Inspector. Notified R4DO (Taylor). | Reactor Core Isolation Cooling High Pressure Core Spray Main Steam Line Safety Relief Valve Control Rod | |
| ENS 53608 | 14 September 2018 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram | At 1644 (CDT) a manual reactor scram was inserted by placing the Reactor Mode Switch to Shutdown. At 1643 (CDT) the Condensate Booster Pump A tripped on low suction pressure. At 1644 (CDT) the Reactor Feed Pump A tripped on low suction pressure. A Recirculation Flow Control Valve runback occurred as designed. Reactor Water level was approaching the Automatic Low Water Level 3 (11.4 inches) scram set point and manual actions were taken by placing the Mode Switch to Shutdown before the low level set point was reached. All systems responded as expected following the manual scram. The plant is stable in mode 3. This event is being reported under 10CFR 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. The NRC Senior Resident Inspector has been notified. All control rods fully inserted, and decay heat is being removed through the turbine bypass valves to the main condenser. The licensee is investigating the cause of the event. | Reactor Protection System Main Condenser Control Rod | |
| ENS 53188 | 31 January 2018 00:22:00 | 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 Main Turbine Load Oscillations | On 1/30/2018 at 1750 (CST), the Reactor Pressure Control Malfunctions ONEP (Off Normal Event Procedure) was entered due to main turbine load oscillations of approximately 30 MWe peak to peak. At 1822 (CST), a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown due to continued main turbine load oscillations. Reactor SCRAM ONEP, Turbine Trip ONEP, and EP-2 were entered. Reactor water level was stabilized at 36 inches narrow range on startup level and reactor pressure stabilized at 933 psig using main turbine bypass valves. Reactor Water Level 3 (11.4 inches) was reached which is the setpoint for Group 2 (RHR to Radwaste Isolation) and Group 3 (Shutdown Cooling Isolation). No valve isolated in these systems due to all isolation valves in these groups being in their normally closed position. The lowest Reactor Water level reached was -36 inches wide range. No other safety system actuations occurred and all systems performed as designed. That event is being reported under 10CFR 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 also reported under 10CFR 50.72(b)(3)(iv)(A), as any event or condition that results in actuation of RPS. The MSIVs are open with decay heat being removed via steam to the main condenser using the bypass valves. Off site power is stable, and the plant is in a normal shutdown electrical lineup. RCIC (Reactor Core Isolation Cooling) was out of service for maintenance, and the reactor water level did not reach the system activation level. The cause of the main turbine load oscillations being investigated. The licensee notified the NRC Resident Inspector. | Reactor Protection System Reactor Core Isolation Cooling Main Turbine Shutdown Cooling Main Condenser | |
| ENS 53090 | 25 November 2017 08:38:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram During Startup | At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. At 0149 (CST), with reactor power just above the point of adding heat, IRM (Intermediate Range Monitor) channels A, C, and D received a spurious upscale trip signal which immediately cleared. Upon investigation, operability of RPS (Reactor Protection System) scram function for Intermediate Range Detectors was placed in question. 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. The licensee notified the NRC Resident Inspector.
At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. At 0149 (CST), with reactor power just above the point of adding heat, Intermediate Range Monitor neutron flux detector (IRM) channels A, C, and D received a spurious Upscale Trip signal which immediately cleared. Upon investigation, IRM channels A, C, and D were declared Inoperable. IRM G was already Inoperable for another reason. RPS scram function from IRM channels B, E, F, and H was always Operable and available. That event is being reported under 10CFR 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. This Revised Statement to Event Notification # 53090 is being made to make it clear that only four IRM channels (A, C, D, G) were Inoperable and that the IRM RPS SCRAM function was still available from the four remaining Operable IRM channels (B, E, F, and H). The licensee notified the NRC Resident Inspector. Notified R4DO (O'Keefe)
On 11/25/17, at 0149 (CST), with reactor power just above the point of adding heat, Intermediate Range Monitor neutron flux detector (IRM) channels A, C, and D received a spurious Upscale Trip signal which immediately cleared. Upon investigation, IRM channels A, C, and D were declared Inoperable. IRM G was already Inoperable for another reason. At 0238 (CST) a manual reactor scram was inserted by placing the Reactor Mode Switch in Shutdown. RPS scram function from IRM channels B, E, F, and H was always Operable and available. That event was initially 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. After the trip alarms were received, the Operators spent approximately twenty minutes investigating possible causes and implications, and consulted with Reactor Engineering and the Shift Technical Advisor. The investigation showed that the plant was stable and the upscale IRM alarms were spurious. A review of plant technical specifications by the operators determined that a plant shutdown was not required. After further discussions, Operations concluded that a shutdown to allow further investigation of the issue was the prudent course of action. Prior to shutting down, Operations spent approximately twenty minutes reviewing procedures, notifying personnel to exit containment, and conducting a brief. The shutdown was then conducted by inserting a manual reactor scram by placing the reactor mode switch in SHUTDOWN. This was initially reported under 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the RPS. Based on the sequence of events, and Operator actions in conducting the shutdown, the event is considered 'part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.73(a)(2)(iv)(A). In accordance with NUREG-1022, Section 3.2.6, the event is not reportable as an actuation of RPS. The licensee notified the NRC Resident Inspector. Notified R4DO (Taylor). | Reactor Protection System Intermediate Range Monitor | |
| ENS 52663 | 4 April 2017 05:10:00 | 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 Condensate Leak | At 0010 (CDT), 04/04/2017, the reactor was manually scrammed from approximately 75 (percent) core thermal power due Condensate Storage tank level lowering to 24 feet. All control rods fully inserted and all systems actuated and operated as designed. No safety relief valves actuated. Reactor level and pressure are currently being controlled within normal bands. RCIC (reactor core isolation cooling) was manually initiated for level control. This event is reportable under 10CFR50.72(b)(2)(iv)(B) for the reactor trip and 50.72(b)(3)(iv)(A) for the manual start of the reactor core isolation cooling system. The cause of lowering level was a condensate pipe leak. Decay heat is being removed via steam dumps to the condenser. The electrical grid is stable and supplying plant loads. The licensee has notified the NRC Resident Inspector. | Reactor Core Isolation Cooling Safety Relief Valve Control Rod | |
| ENS 52044 | 25 June 2016 19:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Turbine Trip | At 1407 (CDT), during power ascension to 100 percent, turbine control valves closed unexpectedly causing reactor protection trip signals that in turn caused a reactor scram. Reactor scram, turbine trip ONEPs (Off Normal Event Procedure), and EP2 (Emergency Procedure for Level Control) were entered. Reactor water level was stabilized at 36 inches narrow range on startup level and reactor pressure stabilized at 935 psig using bypass valves. No other safety system actuations occurred and all systems performed as designed. All control rods inserted. Reactor level is maintained by feedwater. Normal electrical shutdown configuration is through offsite electrical power sources. The Safety Relief Valves lifted, then closed. The plant is stable at normal level and pressure and remains in Mode 3. The event is under licensee investigation. The licensee notified the NRC Resident Inspector. | Feedwater Safety Relief Valve Control Rod | |
| ENS 52012 | 17 June 2016 07:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram During Testing | During planned stop and control valve testing, two main turbine high pressure stop valves closed instead of the expected one (stop valve 'B'). This caused the main turbine control valves, power, reactor pressure to swing and a division 2 half SCRAM. Control rods were inserted to reduce power and the power swings. At 0257 (CDT) the reactor automatically SCRAMMED. Reactor SCRAM, Turbine Trip (procedures) ONEPs and EP-2 were entered. Reactor water level was stabilized at 34 inches narrow range on startup level control and reactor pressure stabilized at 884 psig using main turbine bypass valves. No other safety related systems actuated and all systems performed as expected. The plant is in its normal shutdown electrical lineup using normal feedwater and turbine bypass valves for decay heat removal. Reactor pressure is slowly trending down. The licensee is investigating the cause of the second stop valve shutting. The licensee notified the NRC Resident Inspector. | Feedwater Main Turbine Decay Heat Removal Control Rod | |
| ENS 51827 | 29 March 2016 16:23:00 | 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 Turbine Trip | A Reactor Scram occurred at 1123 CDT on 03/29/2016 from 35% CTP (core thermal power). The cause of the Scram appears to be a Turbine Generator trip. The station's procedures, '05-S-01-EP-2 RPV (Reactor Pressure Vessel) Control, 05-1-02-I-1 Reactor Scram ONEP (Off Normal Event Procedure) and 05-1-02-l-2 Turbine Generator Trip ONEP,' were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF (engineered safety feature) power occurred. No ECCS (emergency core cooling systems) initiation signals were reached and no ESF or Diesel Generator initiations occurred. All control rods fully inserted. MSIVs (main steam isolation valves) remained open, no SRVs (safety relief valves) lifted, and no containment isolation signals were generated. Currently, reactor water level is being maintained by the Condensate and Feedwater System in normal band and reactor pressure and temperature are being maintained by the Reactor Water Cleanup System. The main condenser is available. There are no challenges to Primary or Secondary Containment at this time. The licensee has notified the NRC Resident Inspector. | Feedwater Secondary containment Reactor Water Cleanup Main Condenser Control Rod | |
| ENS 50795 | 8 February 2015 00:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram | A reactor SCRAM occurred at 1856 CST on 2/7/15 from 100 percent core thermal power. The cause of the SCRAM appears to be a Generator/Turbine trip, but it is still under investigation. Appropriate off-normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached, and no ECCS or Emergency Diesel Generator initiations occurred. Main Steam Isolation Valves remained open and Safety Relief Valves lifted and reseated as designed. Currently, reactor water level is being maintained by the Condensate and Feedwater system in normal band, and reactor pressure is being controlled via turbine bypass valves to the main condenser. Following the reactor SCRAM, all rods fully inserted and all systems functioned as expected. The plant is in a normal electrical lineup. The licensee has notified the NRC Resident Inspector. | Feedwater Emergency Diesel Generator Main Steam Isolation Valve Safety Relief Valve Main Condenser | |
| ENS 49972 | 29 March 2014 15:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to a Turbine Trip | Actuation of RPS with reactor critical. Reactor Scram occurred at 1008 (CDT) on 03/29/2014 from 87% CTP (core thermal power). The cause of the Scram appears to be a Turbine Generator Trip. 05-S-01-EP-2 RPV Control, 05-1-02-I-1 Reactor Scram ONEP and 05-1-02-l-2 Turbine Generator Trip ONEP were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF (engineered safety feature) power occurred. No ECCS (emergency core cooling systems) initiation signals were reached and no ESF or Diesel Generator initiations occurred. All control rods are fully inserted. MSIVs (main steam isolation valve) remained open and no SRVs (safety relief valves) lifted. Currently, reactor water level is being maintained by the Condensate and Feedwater system in normal band and reactor pressure is being controlled via Main Turbine Bypass Valves to the main condenser. There are no challenges to Primary or Secondary Containment at this time. The licensee has notified the NRC Resident Inspector. | Feedwater Secondary containment Main Condenser Control Rod | |
| ENS 49920 | 17 March 2014 10:14:00 | 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 Steam Leak in Low Pressure Turbine Line | On 3/17/2014 at 0514 (CDT) the reactor was manually scrammed from approximately 41% core thermal power due to a steam leak in the turbine building. All control rods fully inserted and all systems actuated and operated as designed. All Main Steam Isolation Valves were manually shut. The Reactor Core Isolation Cooling System was manually initiated to assist in level control and pressure control. No safety relief valves actuated automatically. Manual cycling of safety relief valves and Reactor Core Isolation Cooling are being used to maintain reactor water level and pressure within normal bands. Group 2 and 3 RHR isolation signals were received; however no valve movement occurred since the affected valves are normally closed. This event is reportable under 10CFR50.72(b)(2)(iv)(B) for the reactor trip and 50.72(b)(3)(iv)(A) for the manual start of the reactor core isolation cooling system. The licensee informed the NRC Resident Inspector. | Main Steam Isolation Valve Reactor Core Isolation Cooling Safety Relief Valve Control Rod | |
| ENS 49225 | 30 July 2013 19:32:00 | 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 Turbine Generator Trip | Actuation of Reactor Protection System with reactor critical. Reactor Scram occurred at 1432 CDT on 7/30/2013 from 100% Power. The cause of the scram appears to be a Turbine Generator trip. 05-S-01-EP-2, 'Reactor Pressure Vessel Control,' 05-1-02-I-1, 'Reactor Scram Off Normal Event Procedure,' and 05-1-02-I-2, 'Turbine and Generator Trip Off Normal Event Procedure,' were entered to mitigate the transient. No Loss of Off-site Power occurred. No Emergency Core Cooling System or Diesel Generator initiation occurred. Reactor Core Isolation Cooling initiated and injected. The lowest reactor water level reached was -36 inches wide range (RCIC initiation set point is -41.6 inches wide range). Main Steam Isolation Valves remained open and no Safety Relief Valves actuated. Currently, Main Turbine Bypass valves are controlling reactor pressure to the Main Condenser and Condensate and Feedwater is controlling reactor water level in the normal band and removing decay heat. There are no challenges to Primary or Secondary Containment. The NRC Senior Resident Inspector was notified. * * * UPDATE FROM CHRIS ROBINSON TO PETE SNYDER AT 1841 EDT ON 7/30/13 * * * The first out recorder indicated that RPS actuation signal was due to high reactor pressure as a result of the turbine control valves going shut. Notified R4DO (Farnholtz). | Feedwater Secondary containment Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Reactor Pressure Vessel Emergency Core Cooling System Safety Relief Valve Main Condenser | |
| ENS 48673 | 15 January 2013 00:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram from 100% Power Due to a Turbine/Generator Trip | Actuation of RPS (Reactor Protection System) with reactor critical. The Reactor Scram occurred at 1805 (CST) 01/14/13 from 100% CTP (Core Thermal Power). The cause of scram appears to be a Turbine Generator Trip. 05-S-01-EP-2 RPV Control, Reactor Scram ONEP (Off Normal Event Procedure) 05-1-02-I-1, and Turbine and Generator Trips ONEP 05-1-02-1-2 were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached and no ECCS or Diesel Generator initiation occurred. All control rods are fully inserted. MSIVs remained open and SRVs lifted and reseated as designed. Currently, reactor water level is being maintained by the Condensate and Feedwater system in the normal band and reactor pressure is being controlled via Main Turbine Bypass valves to the main condenser. There are no challenges to Primary or Secondary Containment at this time. The licensee notified the NRC Resident Inspector. | Feedwater Secondary containment Main Condenser Control Rod | |
| ENS 48652 | 5 January 2013 05:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to a Turbine/Generator Trip | Actuation of RPS with reactor critical. Reactor Scram occurred at 2337 CST 01/04/13 from 94% CTP (Core Thermal Power). The cause of the scram appears to be a Generator/Turbine trip. Appropriate off normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached and no ECCS or Diesel Generator initiation occurred. MSIVs remained open and SRVs lifted and reseated as designed. Currently, reactor water level is being maintained by the Condensate and Feedwater system in normal band and reactor pressure is being controlled via bypass valves to the condenser. The NRC Resident Inspector has been informed. See similar EN #48637. | Feedwater | |
| ENS 48637 | 29 December 2012 06:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to a Turbine/Generator Trip | Actuation of RPS (Reactor Protection System) with reactor critical. Reactor Scram occurred at 0018 (CST), 12/29/12, from 100% CTP (Core Thermal Power). The cause of scram appears to be a Generator/Turbine trip. Appropriate off normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF (engineered safety feature) power occurred. No ECCS initiation signals were reached and no ECCS or Diesel Generator initiation occurred. MSIVs (Main Steam Isolation Valves) remained open and SRVs (Safety Relief Valves) lifted. Currently, reactor water level is being maintained by the condensate and feedwater system in normal band and reactor pressure is being controlled to limit cooldown. All control rods inserted. The plant is in hot shutdown with decay heat removal to the condenser and the electrical line-up is in a normal configuration. The cause of the turbine/generator trip is under investigation. The licensee notified the NRC Resident Inspector. | Feedwater Decay Heat Removal Control Rod | |
| ENS 47679 | 20 February 2012 01:04:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram During Shutdown Due to Lowering Reactor Vessel Water Level (Rvwl) | On 2/19/2012 at 1904 hrs (CST) the reactor was manually scrammed from approximately 23% core thermal power due to lowering reactor water level. All control rods fully inserted and all systems actuated and operated as designed. The Reactor Core Isolation Cooling System was manually initiated to assist in level control. No safety relief valves actuated. Reactor level and pressure are currently being controlled within normal bands. Group 2 and 3 RHR Isolation signals were received, however no valve movement occurred since the affected valves are normally closed. This event is reportable under 10CFR50.72(b)(2)(iv)(B) for the reactor trip and 50.72(b)(iv)(A) for the manual start of the core isolation cooling system. The lowest Reactor Vessel Water Level (RVWL) observed was -38 inches WR (Wide Range). RVWL was restored by placing the "A" Reactor Feed Pump (RFP) in service. The "B" RFP which had been operating was secured for troubleshooting. The Unit was in the process of shutting down for its scheduled Refueling Outage #18. The NRC Resident Inspector was in the control room at the time of the transient. | Reactor Core Isolation Cooling Safety Relief Valve Control Rod | |
| ENS 45753 | 8 March 2010 22:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram on Low Reactor Water Level | An 'A' reactor feed pump trip occurred and initiated a recirculation pump flow control valve runback. During the flow control valve runback, the 'A' recirculation hydraulic power unit tripped causing an incomplete runback. Reactor water level 3 (11.4 inches) was reached which is an RPS scram setpoint and also a setpoint for Group 2 (RHR to Radwaste) and Group 3 (Shutdown Cooling Isolation). No valves isolated in these systems due to (the valves) being in their normally closed position. The lowest reactor water level reached was -28 inches wide range. Appropriate off normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached, and no ECCS or Diesel Generator initiation occurred. All safety systems performed as expected. MSIVs remained open and no SRVs lifted. Main condenser and pressure control systems remained in service. Currently, reactor water level is being maintained by the condensate and feedwater system in the normal level band and reactor pressure is being controlled to limit cooldown. During the scram, all rods inserted into the core. The plant is in its normal shutdown electrical lineup. Decay heat removal is via the steam bypass valves to the condenser. The licensee has notified the NRC Resident Inspector. | Feedwater Shutdown Cooling Decay Heat Removal Main Condenser | |
| ENS 44601 | 26 October 2008 16:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram and Isolation Due to Turbine Trip | Actuation of RPS with reactor critical. Reactor Scram occurred at 11:25, 10/26/08 from approximately 50% CTP (core thermal power). A turbine/generator trip and automatic RPS reactor scram on TCV (turbine control valve) fast closure occurred. Exact cause of turbine/generator trip is not known at this time. All withdrawn control rods fully inserted to position 00. Reactor Water Level 3 (11.4") was reached which is an RPS Scram Setpoint and also a setpoint for Group 2 (RHR to Radwaste) and Group 3 (Shutdown Cooling Isolation). No valves isolated in these systems due to their being in their normally closed position. Lowest reactor water level reached was -3" Wide Range. Appropriate off normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached, and no ECCS or DG initiation occurred. All safety systems performed as expected. MSIVs remained open and no SRV's lifted. Main Condenser and pressure control system remained in service. Currently, reactor water level is being maintained by the condensate system in normal level band and reactor pressure is being controlled to limit cooldown. The licensee notified the NRC Resident Inspector. | Shutdown Cooling Main Condenser Control Rod | |
| ENS 44595 | 23 October 2008 12:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram with Manual Rcic Initiation | 'A' Reactor Feed Pump speed decreased to zero with no trip signal evident in the Control Room for the 'A' Reactor Feed Pump. The reactor scrammed due to loss of feedwater flow on a Level 3 (11.4") RPS scram signal as designed. Operators implemented appropriate off normal event procedures to mitigate the transient with all systems responding as designed. Lowest reactor water level observed was -35" wide range. All withdrawn control rods fully inserted to position '00'. The Reactor Core Isolation Cooling System was manually initiated and was used to restore water level to within the normal band. No ECCS initiations were received and all systems responded as designed. Additionally, no SRVs lifted as a result of this event. Level 3 is also a setpoint for Group 2 (RHR to Radwaste) and Group 3 (Shutdown Cooling Isolation) automatic isolation. No valves isolated in these systems due to them being in their normally CLOSED position prior to the event. Currently, reactor water level is being maintained by the condensate system in normal level band and reactor pressure is being controlled to limit cool down. The licensee has notified the NRC Resident Inspector. | Feedwater Reactor Core Isolation Cooling Shutdown Cooling Control Rod | |
| ENS 44086 | 21 March 2008 20:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram from a Main Turbine Trip | Actuation of an RPS signal while the reactor was critical. At 1525 a generator trip signal on the main transformers initiated a reactor scram due to Turbine Stop and Control Valve fast closure. Safety relief valves operated initially to lower reactor pressure. Turbine bypass valves operated initially to lower reactor pressure. Turbine bypass valves are maintaining pressure control currently. Reactor recirc pumps A and B transferred to slow speed operation as expected. No ECCS initiations were received. Reactor level is being controlled with normal systems condensate and feedwater. Lowest level indicated was approximately -6" wide range. Level 3 initiations occurred to group 3 isolation valves. No valves operated, they are normally closed. The site is investigating the cause of the trip. A probable cause is the "C" phase differential trip which was received from the main transformer. All control rods fully inserted during the scram. The plant is on its normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector. | Feedwater Main Transformer Safety Relief Valve Control Rod | |
| ENS 43899 | 12 January 2008 22:26:00 | 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 Degraded Cooling to Main Transformers | On 1/12/08 at 1626 hours a manual scram was inserted due to degraded cooling on the main transformers. All control rods fully inserted and all systems actuated and/or operated as designed. No safety relief valves opened during the event and reactor pressure is currently being controlled within normal bands with bypass valves. Reactor water level is also being controlled within normal bands with the condensate and feedwater system. Group 2 and 3 isolation signals were received however, no valve movement occurred since these valves are normally closed." The licensee is investigating a potential electrical fault which may have caused the event, but it appears that there was an electrical fault in the transformer cooling system. The grid is normal with no transmission system warnings in effect. All feeds are available. Reactor vessel level is 32.6 inches stable with normal condensate system feeding the vessel. Primary plant pressure is 815 psig. All safety and BOP electrical buses are energized normally. No bus power was lost. Decay heat path is via bypass valves to the main condenser. The licensee will notify the NRC Resident Inspector. | Feedwater Main Transformer Safety Relief Valve Main Condenser Control Rod | |
| ENS 43584 | 21 August 2007 19:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scrammed Due to Reactor Water Level 3 Rps Actuation | The 'A' Reactor Feed Pump controller failed downscale causing reactor water level to lower. Feed pump was manually tripped as reactor water level passed through 20 inches (Narrow Range) in an attempt to stabilize level. Reactor scrammed on Level 3 (11.4") as designed. Mode switch placed in Shutdown. Lowest level observed was -25". All control rods fully inserted. No ECCS initiations were received and all systems responded as designed. Level 3 is also a Group 3 Shutdown Cooling Isolation setpoint. No valves isolated due to them being in their SOI (system operating instructions) position 'closed' prior to the event. The IC Tech group was working on the Feedwater Pump Controller at the time of the trip and could have possibly caused this situation. However, this has yet to be determined. The licensee notified the NRC Resident Inspector. | Feedwater Shutdown Cooling Control Rod | |
| ENS 43376 | 19 May 2007 16:27:00 | 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 Turbine Trip on Loss of Condenser Vacuum | Grand Gulf scrammed at 11:27 on 5/19/07 due to Turbine trip caused by loss of vacuum (believed to be from condenser neck seal). All safety systems responded as required. All control rods inserted. No ECCS systems were required to initiate. Reactor vessel level 3 was reached and both divisions of isolation logic initiated. All associated valves already closed (normal lineup). Present condition is Mode 3 and reactor pressure is 250 psig. Cooldown is in progress. No relief valves lifted as a result of the scram. Decay heat was initially removed via bypass valves to condenser. The plant then shifted to decay heat removal via drains and SRV's. Suppression pool temperature is 80 degrees and available if required. Offsite power is available and all emergency systems are available if needed. The licensee will notify the NRC Resident Inspector. | Decay Heat Removal Control Rod | |
| ENS 41405 | 12 February 2005 01:59:00 | 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 | Rps Actuation | The following text was obtained from the licensee via facsimile: At approximately 1959 hours Central Standard Time, Grand Gulf Nuclear Station experienced a loss of Electrical Bus 11R. This caused a loss of additional electrical buses, and a subsequent reactor scram on low reactor water (level). The feedwater system was lost, and reactor low level 2 was reached. The Reactor Core Isolation Cooling System and High Pressure Core Spray System injected into the reactor and restored reactor water level. A primary containment, secondary containment, and drywell isolation occurred as expected. The Division 1 Diesel started and picked up the Division 1 bus. The Division 3 Diesel Generator started on the reactor low water level 2. The condenser is removing decay heat. The plant status currently is stable, with normal reactor water level and feedwater restored. All control rods inserted. The cause for the loss of Bus 11R is under investigation. Safety systems appeared to function as designed. The licensee notified the NRC Resident Inspector. | Feedwater Secondary containment Reactor Core Isolation Cooling Primary containment High Pressure Core Spray Control Rod |