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| Discovered date | Reporting criterion | Title | Event description | |
|---|---|---|---|---|
| ENS 57032 | 16 March 2024 19:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Main Feedwater and Main Steam Isolations | The following information was provided by the licensee via phone and email: At 1449 CDT, Waterford 3 Steam Electric Station was operating at 100 percent power when a manual reactor trip was initiated due to main feed isolation valve (FW-184B) and main steam isolation valve (MS-124B) going closed unexpectedly. Emergency feedwater (EFW) was automatically actuated. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip. The unit is currently stable in Mode 3. All control rods fully inserted as expected. This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour nonemergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Decay heat is being removed through the turbine bypass valves and the atmospheric dump valve on loop '2'. There is no primary to secondary system leakage. The cause of the isolations is still being investigated. |
| ENS 57014 | 7 March 2024 19:42:00 | Non | Technical Specification Violation | The following information was provided by the licensee via phone and email: On March 7, 2024, at 1142 PST, an operator trainee operating the reactor under the direction of a licensed operator initiated a planned manual scram. Following the planned manual scram, the licensed operator did not switch the console switch to 'off' or remove the key from the console. The reactor did not meet the definition of 'reactor secured' and thus the staffing requirements of technical specification 6.1.3 were still required to be met. The licensed operator then left the control room, securing the door on their way out. At 1200 PST, a licensed senior reactor operator (SRO) entered the control room and found the key in the console with the switch in the 'operate' position. This SRO placed the switch in the 'off' position, secured the key, logged the action, and notified the Director. Throughout the duration of the event, all control rods were fully inserted. Project Manager (Wertz) will be contacted. Oregon Department of Energy and the Oregon Radiation Protection Services will be notified. |
| ENS 57006 | 5 March 2024 06:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 0132 EST, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main feedwater isolation signal which resulted in steam generator lo-level reactor trip. The reactor trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the main feedwater isolation is being investigated. |
| ENS 57004 | 4 March 2024 00:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Main Turbine Trip on LOW Condenser Vacuum | The following information was provided by the licensee via email: On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram. Additionally, following the scram a low RPV (reactor pressure vessel) level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR (residual heat removal) shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system. This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration. |
| ENS 56968 | 15 February 2024 08:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Turbine Trip/Reactor Trip | The following information was provided by the licensee via email: At 0247 CST on 2/15/2024, Callaway Plant was in mode 1 at approximately 100 percent power when a turbine trip and reactor trip occurred. All safety systems responded as expected with the exception of an indication issue on the feedwater isolation valves, which were confirmed closed. A valid feedwater isolation signal and auxiliary feedwater actuation signal were also received as a result of the reactor trip. The plant is being maintained stable in mode 3. All control rods fully inserted from the reactor trip signal and decay heat is being removed via the auxiliary feedwater system and steam dumps. The NRC Resident Inspector was notified. |
| ENS 56936 | 29 January 2024 17:02:00 | 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 email: At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram caused by a main turbine trip. Investigation is still ongoing. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods were fully inserted. The licensee indicated that the turbine trip may have been caused by a power load imbalance, however the cause of the incident is under investigation. The scram was not complex. Decay heat is currently being removed thru bypass valves dumping to the main condenser. Initially unit 2 lost the use of the bypass valves due to lack of condenser vacuum. Unit 2 used the high pressure coolant injection (HPCI) system in the condenser storage tank (CST) to CST mode to remove decay heat. Residual heat removal was used to keep the torus cool. Condenser vacuum was regained and unit 2 is back to removing decay heat with the turbine bypass valves. There was no impact to unit 3. The licensee confirmed there was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: Licensee adds 8-hour non-emergency 10 CFR 50.72(b)(3)(iv)(A) specified system actuation report to original 4-hour non-emergency 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation report. At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram by a main turbine trip. All control rods inserted. Reactor core isolation cooling system (RCIC) was manually initiated for level control. HPCI was manually initiated for pressure control. Primary containment isolation system (PCIS) Group II and III isolations occurred (specified system actuation). Investigation is ongoing. The NRC Resident Inspector has been notified. |
| ENS 56935 | 28 January 2024 02:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 2141 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. The licensee notified the NRC Resident Inspector. |
| ENS 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. |
| ENS 56889 | 15 December 2023 00:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram | The following information was provided by the licensee via phone call and email: On December 14, 2023, at 1939 EST, Hope Creek reactor scrammed following closure of turbine control valve number 4. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The outage control center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. |
| 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. |
| ENS 56803 | 19 October 2023 16:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: On 10/19/2023, at approximately 1110 (CST), with Unit 1 in mode 1 at 100 percent power, the reactor automatically tripped. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The cause of the trip is being investigated. Operations responded and stabilized the plant. Auxiliary feedwater actuated as expected. Decay heat is being removed by the steam generator through the steam generator power operated relief valve. The trip was complex as non-safety related power was lost to both Unit 1 and Unit 2. Unit 1 is currently in mode 3 and on natural recirculation as both reactor coolant pumps are without power. Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was lost for approximately 70 minutes. No impacts to the SFP temperature were observed. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Due to the actuation of the auxiliary feedwater system following the reactor trip, this event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The second paragraph of the original report is amended as follows to correct information regarding the spent fuel pool for Unit 2: Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was maintained at all times with one train of SFP cooling. The second train lost power and was restarted approximately 70 minutes (after power was lost). No impacts to the SFP temperature were observed. Notified R3DO (Orth) and IR MOC (Crouch) and NRR EO (Felts) via email |
| ENS 56790 | 13 October 2023 01:27:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip | The following information was provided by the licensee via email: On 10/12/23 at 2127 EDT, with the Unit 1 in Mode 1 at 100% Power, operators identified degrading condenser vacuum and manually tripped the reactor. All control rods inserted as expected. The trip was not complex, and all systems responded normally post-trip. The cause of the degraded condenser vacuum was an unexpected closure of the condenser air ejector regulator. The cause of the air ejector regulator going closed is not fully understood and is being investigated. Following the SCRAM, Operators responded and stabilized the plant. Decay heat is being removed by the Main Steam System through the Atmospheric Relief Valves (ARVs) and Auxiliary Feed Water (AFW) systems. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56731 | 9 September 2023 15:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: On 9/9/23 at 1143 EDT, with the Unit 1 in Mode 1 at 100 percent power, all 4 turbine control valves closed resulting in a reactor protection system (RPS) automatic reactor trip on over temperature differential temperature. All control rods inserted as expected. The trip was not complex and all systems responded normally post-trip. The cause of the control valve closure has not been determined. Following the SCRAM, operators responded and stabilized the plant. Decay heat is being removed by the main steam system through the atmospheric relief valves and auxiliary feed water systems. Due to the RPS actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56710 | 2 September 2023 10:32:00 | 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 | Automatic Reactor Scram Due to Feedwater Transient | The following information was provided by the licensee via email: On 9/2/2023 at 0632 EDT, a feedwater transient occurred resulting in an reactor protection system (RPS) automatic reactor scram on low level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 recirculation sample system isolation, Group 3 traveling in-core probe (TIP) isolation valve isolation, Group 6 and 7 reactor water cleanup isolation, and Group 9 containment purge isolations. All control rods inserted as expected. High pressure core spray and reactor core isolation cooling initiated and injected as expected. ECCS systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable and in Mode 3. These 4 hour and 8 hour non-emergency reports are being made in accordance with 10 CFR 50.72(b)(2) (iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident was informed. There was no impact on Unit 1. |
| ENS 56660 | 4 August 2023 21:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via phone and email: At 1746 EDT on 08/04/2023, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to number 2 steam generator low low level. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the number 2 steam generator low low level is being investigated. |
| ENS 56593 | 27 June 2023 20:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via phone and email: At 1626 EDT, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a main turbine trip. The (reactor) trip was not complex with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed using the auxiliary feedwater and steam dump systems. Unit 1 is not affected. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The expected actuation of the auxiliary feedwater system (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All control rods are fully inserted. The cause of the turbine trip is being investigated. |
| ENS 56531 | 20 May 2023 08:15:00 | 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: On 5/20/2023 at 0315 CDT, Browns Ferry Unit 1 was at 80 percent reactor power performing, 'Turbine control valve fast closure turbine trip and RPT (recirculation pump trip) initiate logic testing'. During performance of this test, Unit 1 received a full reactor scram. An investigation is in progress to determine the cause of the scram. All systems responded as expected, and Unit 1 is stable at zero percent power in mode 3. All control rods fully inserted into the core. Main steam isolation valves remained open with main turbine bypass valves controlling pressure. Reactor feedwater pumps remained in service to control reactor water level. Primary containment isolation signals groups 2, 3, 6, and 8 were received with expected system actuations. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. The event is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. The NRC Resident has been notified. |
| ENS 56389 | 4 March 2023 15:10:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Reactor Protection System (RPS) | The following information was provided by the licensee via email: At 0910 (CST), with Unit 2 in Mode 4 at 0 percent power, an actuation of a reactor scram on low charging water header pressure occurred during restoration from hydrostatic test conditions. All control rods were already fully inserted prior to the receipt of the scram signal. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Unit 2 RPS system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| 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. |
| ENS 56278 | 17 December 2022 05:51:00 | 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 email: On December 16, 2022 at 2351 CST, with the Unit in Mode 1 at 13 percent power, a manual scram was inserted due to lowering Reactor Pressure Vessel (RPV) pressure, which occurred following an unexpected opening of Main Turbine Bypass Valve 1. All control rods fully inserted. Following actuation of the manual scram, RPV pressure lowered, resulting in an automatic Primary Containment lsolation (PCIS) Group 1 isolation (expected response). The main steam isolation valves and steam line drain valves all closed. The Group 1 (isolation) has been reset allowing RPV pressure control with steam line drains to the main condenser. All systems responded as designed. The plant is stable in Mode 3. Investigation of the bypass valve opening is ongoing. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation and 50.72(b)(3)(iv)(A) Specified System Actuation. There was no impact on health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. |
| ENS 56220 | 13 November 2022 04:19:00 | 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 email: On November 12, 2022, at 2319 CST, an actuation of the reactor protection system (RPS) initiated a full scram. The plant was in Mode 2, reactor pressure was 149 pounds. The high pressure coolant injection (HPCI) injection valve, HPCI-MOV-MO19, opened and injected cold water into the reactor vessel while HPCI system testing was in progress. The cause is still under investigation. All control rods inserted. Plant is currently in Mode 3 and stable. All systems operated as designed with no Primary Containment Isolation System group isolations. This event is being reported under two event classifications: 50. 72(b)(2)(iv)(B) -- "Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. 50. 72(b)(3)(iv)(A) -- "Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The NRC Resident has been informed. |
| ENS 56216 | 10 November 2022 12:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip and Auxiliary Feedwater System Actuation | The following information was provided by the licensee via email: At 0744 EST on November 10, 2022, DC Cook Unit 2 tripped automatically on high-high level of number 23 steam generator (SG). The reason for the high-high level in SG 23 is under investigation. This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The DC Cook NRC Resident Inspector has been notified. Unit 2 is being supplied by offsite power. All control rods fully inserted. All Auxiliary Feedwater Pumps started properly. Decay heat is being removed via the Steam Dump System. Preliminary evaluation indicates all plant systems functioned normally following the reactor trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the post trip review. No radioactive release is in progress as a result of this event. |
| ENS 56153 | 7 October 2022 06:19:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Control Rod Drive Mechanism (CRDM) Penetration Degraded | The following information was provided by the licensee via fax: Control Rod Drive Mechanism (CRDM) penetration 69 degraded. At 0119 (CDT) on October 7, 2022, it was determined that the CRDM penetration 69 was degraded because examination identified unacceptable indications in accordance with ASME Code Case N-729-6. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The notification is being corrected to state: At 0119 (CDT) on October 7, 2022, it was determined that the Control Rod Drive Mechanism (CRDM) penetration 69 was degraded because liquid penetrant testing, performed on the seal weld, identified unacceptable indications in accordance with ASME Section III and NRC approved licensee relief request for a previously performed embedded flaw repair. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. Notified R3DO (Ruiz). |
| ENS 56119 | 26 September 2022 07:06:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | The following information was provided by the licensee via email: At 0001 EDT on September 26, 2022, James A. FitzPatrick (JAF) removed the generator from service as part of a planned shutdown for refueling. At 0306 EDT, with the mode switch in Startup/Hot Standby and inserting rods, JAF experienced a spurious Scram and closure of seven out of eight main steam isolation valves (MSIV's). The reactor protection system (RPS) actuated during the event, resulting in all control rods being fully inserted. The cause of the closure of MSIV's and the Scram is being investigated. This condition is being reported as a four-hour NRC report per 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation, and as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the safety system actuation based on the multiple main steam isolation valves closing on an isolation signal. There was no impact to the health and safety of the public. The NRC Resident Inspector has been notified.
The following update was provided by the licensee via email: This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A) for Reactor Protection System (RPS) actuation along with Main Steam Isolation Valves (MSIV) system actuation. An analysis of reactor criticality was performed for the period of time prior to the RPS actuation event. Operators were inserting control rods per the shutdown Reactivity Management Plan. The Intermediate Radiation Monitoring (IRM) readings preceding the scram signal demonstrate a negative reactivity direction without control rod movement. The analysis concluded that the reactor was subcritical when RPS was actuated. The NRC Resident Inspector has been notified. Notified R1DO (Young). |
| ENS 56089 | 4 September 2022 20:39:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Activation | The following information was provided by the licensee via fax: On September 4, 2022 with Unit 2 in Mode 3, an (Reactor Protection System) RPS actuation and Containment Isolation occurred on (Reactor Pressure Vessel) RPV Low Level (Level 3) of 159.3 inches due to issues with the normal feedwater level control system during plant cooldown. The RPS actuation occurred with control rods already inserted and a containment isolation on Level 3. The containment isolation signal impacted (Residual Heat Removal) RHR Shutdown Cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. Operators took manual control of RPV level and restored level to the normal operating band shortly after the low level was received. This is being reported under 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72(b)(3)(iv)(B). The NRC Resident Inspector was notified. |
| ENS 56076 | 28 August 2022 17:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Reactor Coolant Pump Trip | The following information was provided by the licensee via fax or email: On August 28, 2022 at 1348 EDT, DC Cook Unit 1 reactor automatically tripped due to a trip of the #13 Reactor Coolant Pump. This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The DC Cook Resident NRC Inspector has been notified. Unit 1 is being supplied by offsite power. All control rods fully inserted. All Auxiliary Feedwater Pumps started properly. Decay heat is being removed via the Steam Dump System. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 1 remains stable in Mode 3 while conducting the Post Trip Review. No radioactive release is in progress as a result of this event. Unit 2 remains stable at 100% power / Mode 1. |
| 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. |
| ENS 55964 | 25 June 2022 03:38:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Main Turbine Trip | The following information was provided by the licensee via email: At 2338 EDT, on June 24, 2022, with the unit in Mode 1 at 100 percent power, the reactor automatically scrammed due to an RPS actuation following a Main Turbine Trip. The cause of the turbine trip is not known at this time. The scram was not complex, with systems responding normally post-scram. Operations responded and stabilized the plant. Reactor water level has been recovered and maintained at the normal level. Decay Heat is being removed by the Main Steam system to the main condenser using the Turbine Bypass Valves. All Control Rods inserted into the core. The transient occurred with no surveillances or activities in progress. Investigation into the cause of the Turbine Trip is in progress. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The low reactor water level caused an isolation of Primary Containment (Groups 4/13/15) as expected. The Primary Containment Isolation Event is being reported under 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified. |
| ENS 55963 | 25 June 2022 01:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Main Steam Isolation Valve Closure | The following information was provided by the licensee via email: At 2012 CDT, Waterford 3 Steam Electric Station, Unit 3 was operating at 100 percent power when an automatic reactor trip occurred due to Main Steam Isolation Valve MS-124B going closed unexpectedly. Subsequently, both main feedwater isolation valves shut. Emergency Feedwater (EFW) was automatically actuated. Preliminary evaluation indicates that all plant systems functioned normally after the reactor trip. The unit is currently stable in Mode 3. All control rods fully inserted as expected and all other plant equipment functioned as expected. This was an uncomplicated scram. This event is being reported as a 4-hour non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B) as an actuation of the reactor protection system (RPS) when the reactor is critical and as an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as valid actuation of the EFW system. The NRC Resident Inspector has been notified. |
| ENS 55943 | 15 June 2022 11:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Automatic Auxiliary Feedwater Actuation | The following information was provided by the licensee via email: At 0724 EDT on 6/15/2022, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to lowering Steam Generator levels due to a secondary plant perturbation in the Heater Drain System. All control rods fully inserted into the core and the Auxiliary Feedwater System automatically started as designed in response to the full power reactor trip. The trip was not complex, with all systems responding normally post-trip. There was no equipment inoperable prior to the event that contributed to the reactor trip or adversely impacted plant response. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the condenser steam dump valves. Unit 2 is not affected and remains at 100 percent power and stable. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Auxiliary Feedwater System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55910 | 24 May 2022 08:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Following Manual Turbine Trip from High Vibrations on Main Turbine | The following information was provided by the licensee via email: On May 24, 2022, at 0414 EDT, while rolling the Unit 1 main turbine during the Unit 1 Cycle 31 refueling outage, the Unit 1 main turbine experienced high vibrations and the main turbine was manually tripped with reactor power at 12 percent. Main turbine vibrations persisted and the reactor was manually tripped, Main Steam Stop Valves were closed, and main condenser vacuum was broken. This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The DC Cook Resident NRC Inspector has been notified. Unit 1 is being supplied by offsite power. All control rods fully inserted. Both Motor Driven Auxiliary Feedwater Pumps started properly. Decay heat is being removed via Steam Generator Power Operated Relief Valves. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 1 remains stable in Mode 3 while conducting the Post Trip Review. No radioactive release is in progress as a result of this event. |
| ENS 55909 | 23 May 2022 21:16:00 | 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 | Automatic Reactor Scram | The following information was provided by the licensee via email: At 1716 hours EDT on May 23, 2022, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed. Unit 1 reactor was being operated at approximately 100 percent (Rated Thermal Power) RTP. The Control Room received indication that both divisions of (Reactor Protection System) RPS actuated from (Reactor Pressure Vessel) RPV high pressure signals and all control rods fully inserted. The Main Turbine bypass valves opened automatically to control reactor pressure. Reactor water level lowered to -42 inches causing Level 3 and Level 2 isolations. (High Pressure Coolant Injection) HPCI (Emergency Core Cooling System) ECCS actuation occurred as designed at -38 inches and injected to the Reactor Vessel. No other ECCS system actuations occurred. (Reactor Core Isolation Cooling) RCIC automatically initiated as designed at -30 inches. The Operations crew subsequently maintained reactor water level at the normal operating band using Feedwater pumps. The reactor is currently stable in Mode 3. An investigation is in progress into the cause of the Automatic SCRAM. The NRC Senior Resident Inspector was notified. A voluntary notification to (Pennsylvania Emergency Management Agency) PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10 CFR 50.72(b)(2)(iv)(A) & 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A). |
| ENS 55899 | 16 May 2022 19:52:00 | 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 Electrical Transients | The following information was provided by the licensee via fax: Unit 2 experienced multiple electrical transients resulting in a Group I Primary Containment Isolation Signal (PCIS) isolation and subsequent unit reactor scram. Low reactor water level during the automatic scram caused PCIS Group II and III isolation signals. Following the PCIS Group I isolation, all main steam lines isolated. All control rods inserted and all systems operated as designed. The following additional information was obtained from the licensee via phone in accordance with Headquarters Operations Officers Report Guidance: Peach Bottom Unit 2 automatically scrammed from 100 percent power due to an electrical transient and subsequent PCIS Group I isolation (Main Steam Isolation Valve closure). Unit 2 lost main feedwater due to the PCIS Group I isolation, however, all other systems responded as expected following the scram. High Pressure Coolant Injection is maintaining pressure control while Condensate Pumps are maintaining inventory. The unit is currently stable and in Mode 3. Peach Bottom Unit 3's Adjustable Speed Drives were impacted by the electrical transients and the unit reduced power to 98 percent power. The NRC Resident Inspector was notified. |
| ENS 55856 | 23 April 2022 06:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip During Control Rod Testing | The following information was provided by the licensee via fax or email: On April 23, 2022, at 0224 (EDT) hours, with Unit 2 in Mode 1 at 100 percent power, two control rods dropped during control rod testing resulting in misalignment, which required a manual reactor trip in accordance with plant procedure. All safety systems functioned as expected. The Auxiliary Feedwater system actuated as designed to provide makeup flow to the steam generators. Operations responded and stabilized the plant. Decay heat is being removed by the steam generator power operated relief valves. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The automatic start of the Auxiliary Feedwater system is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). The cause of the dropped rods is being investigated. Unit 1 is not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified." The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Electrical power is in normal off-site arrangement. All emergency electrical supplies are available. |
| ENS 55843 | 14 April 2022 13:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip During Control Rod Testing | The following information was provided by the licensee via email: On April 14, 2022, at 0928 (EDT) hours, Unit 1 automatically tripped from 100 percent power during the control rod operability periodic test. The reactor trip occurred during the manipulation of the rod control mode selector switch as part of the rod operability testing. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. The reactor trip was uncomplicated, and all control rods fully inserted into the core. This reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater pumps actuated as designed because of the reactor trip and provide makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater system is reportable per 10 CFR 50.72(b)(3)(iv) (A) for a valid actuation of an ESF (Engineered Safety Features) system. The Auxiliary Feedwater pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 1 is in a normal shutdown electrical lineup. An investigation into the cause of the reactor trip is underway. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: There was no affect to Unit 2. Unit 2 is operating at 100 percent power. |
| ENS 55821 | 5 April 2022 06:23:00 | 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 | Scram on LOW Level | The following information was provided by the licensee via telephone and email: On 4/5/2022, at time 0223, during maintenance on Feedwater Level Control Valve 2FWS-LV10B, a Feedwater transient occurred resulting in an RPS Automatic Reactor Scram on Low Level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 Recirculation Sample System Isolation, Group 3 TIP ((Traversing Incore Probe)) Isolation Valve Isolation, Group 6 and 7 Reactor Water Cleanup Isolation and Group 9 Containment Purge Isolations. All control rods inserted as expected. High Pressure Core Spray and Reactor Core Isolation Cooling initiated and injected as expected. ECCS Systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable in Mode 3. These 4 hour and 8-hour non-emergency ENS ((Emergency Notification System)) reports are being made in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident was informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: There was no impact on Unit 1. |
| ENS 55748 | 18 February 2022 09:59:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and System Actuation | The following information was provided by the licensee via telephone and email: On 2/18/2022, McGuire Nuclear Station Unit 2 experienced a turbine runback to 55 percent power. Based on concerns with unit stability, the reactor was manually tripped at 0459 (EST). All Auxiliary Feedwater pumps started on low steam generator level as required. The reactor trip was uncomplicated with all systems responding normally post trip. A feedwater isolation occurred as designed. Unit 1 was not affected. Due to the Reactor Protection System actuation while critical, actuation of the Turbine Driven Auxiliary Feedwater Pump and Motor Driven Auxiliary Feedwater pumps along with the Feedwater Isolation, this event is being reported as a four hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour nonemergency notification per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods fully inserted. Decay heat is being removed via the condenser and normal feedwater. Unit 2 is in a normal shutdown electrical lineup. |
| ENS 55732 | 4 February 2022 22:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | The following information was provided by the licensee via email: At 1700 EST, on February 4, 2022 with the unit in Mode 1 at 58 percent power, the reactor automatically scrammed due to low Reactor water level due to a transient on the Feedwater System while preparing to shutdown for a refueling outage. The scram was not complex, with systems responding normally post-scram. Operations responded and stabilized the plant. Reactor water level has been recovered and maintained at normal level. Decay Heat is being removed by the Main Steam system to the main condenser using the Turbine Bypass Valves. All Control Rods inserted into the core. The transient occurred while in the process of removing the South Reactor Feed Pump from service. While reducing speed on the South, the North Reactor Feed Pump increased in speed and tripped on low suction. The plant was preparing to shut down for a refueling outage when the trip occurred. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, in preparation of plant shutdown, Primary Containment De-Inerting was in progress. The low Reactor water level caused an isolation of Primary Containment (Groups 4/13/15). The Primary Containment Isolation Event is being reported under 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified. |
| ENS 55698 | 7 January 2022 18:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Turbine Trip / Reactor Trip | The following information was provided by the licensee via email: At 1223 CST on January 7, 2022, Callaway Plant was in Mode 1 at approximately 100 percent power when a turbine trip / reactor trip occurred. All safety systems responded as expected with the exception of an indication issue with the 'B' Feedwater Isolation Valve, which was confirmed closed. A valid Feedwater Isolation Signal and Auxiliary Feedwater Actuation Signal were also received as a result of the reactor trip. The plant is being maintained stable in Mode 3. All control rods fully inserted from the reactor trip signal, and decay heat is being removed via the Auxiliary Feedwater and Steam Dump Systems. The NRC Senior Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The plant is in a normal shutdown electrical lineup. |
| ENS 55632 | 7 December 2021 17:48:00 | Non 10 CFR 50.72(b)(1), 50.54(x) TS Deviation | Research Reactor Tech Spec Deviation Resulting in a Reactor Scram | On 12/7/2021 at 1248 EST, the Breazeale reactor (50-005) (a TRIGA reactor) was operating for a NUCE 451 lab. The reactor operator was trying to perform a $0.75 (measure of reactivity) square wave with a setpoint of 500 kW. When the operator entered the setpoint, he did not hit ENTER so the 500 kW was not stored in the console, which defaulted to the current power level (100 W) as the setpoint. Once the square wave was executed, the control rods immediately began to move in to counter the $0.75 of reactivity from the pulse rod and maintain the power level at 100 W. The operator noticed that the setpoint was incorrect, and after 9 seconds, changed the power setpoint to 500 kW. At this time, the rod bank began to move out, adding $1.20 over the course of 4 seconds. (The maximum total reactivity beyond critical is estimated to be ~$1.10). The reactor scrammed based on high log range (fission chamber) power and high wide range (GIC) power. The last two points of data from the data historian indicate that the period was +0.25 seconds. Based on the rod insertion speed and differential rod worth at the position from which the scram was initiated, it was estimated that the maximum power following the scram (setpoint = 1.08 MW) was approximately 1.29 MW. The highest data point recorded by the historian was 1.38 MW (log fission chamber data), which is corroborated by the estimate calculated based on rod speed, and 1.38 MW represents the best estimate of the maximum reactor power. The reactor technical specifications (TS) dictate that: "The maximum power level SHALL be no greater than 1.1 MW (thermal)." (TS 3.1.1.b). This condition applies to non-pulse operation. According to the TS definitions, the reactor is neither "secured" nor "shut down" during a scram, and therefore must be considered to be operating while the rods are in motion after the scram is initiated. Therefore, this event resulted in the violation of TS 3.1.1.b by allowing power to reach 1.38 MW, higher than the 1.1 MW scram setpoint. It is worth noting that TRIGA reactors like the Breazeale reactor are designed to be pulsed to several gigawatts of power, and the 1.1 MW limit is based on steady state power analysis, not power transient analysis. The fuel reached a maximum temperature of 42 C, far below the safety limit of 1150 C (TS 2.1). This event is reportable due to: exceeding an LCO in the technical specifications (1.1 MW power limit) and an unanticipated change in reactivity greater than $1 when the rod bank drove out following the change in power setpoint. The root cause of this event was operator error. The operator failed to follow best practices by checking that the setpoint was entered correctly, and then acted outside of procedure to attempt to correct the setpoint. The reactor was immediately secured and tagged out pending corrective actions identified in the event evaluation document, AP4 2021-03. The immediate corrective actions, completed on 12/8/21, were to: 1 - add a pen-and-ink revision to SOP-1 instructing the operator to verify the power setpoint; 2 - hold a reactor staff training on the event, its causes, and the importance of following procedure and checking values entered into the console; 3 - implement an administrative prohibition on square waves until the console software can be changed to add a feature to prevent recurrence of this event. Following these corrective actions reactor operation was approved by the ADO (Level 2). A detailed written report will be sent to the Reactor Safeguards Committee and NRC by December 21st. The licensee will notify the Non-Power Production or Utilization Facility (NPUF) Licensing Branch Project Manager. |
| ENS 55616 | 30 November 2021 17:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Automatic Scram | At 1254 EST on November 30, 2021, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed during Turbine Valve Cycling surveillance activities. Unit 1 reactor was being operated at approximately 80 percent rated thermal power with turbine valve cycling surveillance activities in progress. The Control Room received indication that both divisions of RPS (reactor protection system) actuated from turbine valve closure signals and all control rods fully inserted. The Main Turbine was manually tripped, and turbine bypass valves opened automatically to control reactor pressure. Reactor water level lowered to -35 inches causing Level 3 and Level 2 isolations. No ECCS (emergency core cooling systems) actuations occurred. RCIC (reactor core isolation cooling) automatically initiated as designed at -30 inches. The Operations crew subsequently maintained reactor water level at the normal operating band using Feedwater pumps and RCIC was placed in a standby lineup. The reactor is currently stable in Mode 3. An investigation is in progress into the cause of the turbine valve closure signals. The NRC Senior Resident Inspector was notified. A voluntary notification to PEMA (Pennsylvania Emergency Management Agency) will be made. This event requires a 4-hour ENS notification in accordance with 10CFR50.72(b)(2)(iv)(B) and an 8-hour ENS notification in accordance with 10CFR50.72(b)(3)(iv)(A). Unit 2 was not affected and remains at 100 percent power, Mode 1. |
| ENS 55575 | 14 November 2021 10:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Trip Due to Lowering Main Condenser Vacuum | At 0525 EST, November 14, 2021, "Unit 2 was manually scammed by operations due to lowering main condenser vacuum. This resulted in PCIS (primary containment Isolation system) Group II/III isolation signals. All control rods inserted, and all systems operated as designed. Unit 3 is unaffected and remains at 100 percent power in Mode 1. The Resident Inspector was notified. |
| ENS 55514 | 11 October 2021 17:21:00 | 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 | Automatic Reactor Scram After Main Turbine Trip | At 1321 EDT on October 11, 2021, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed due to a trip of the Main Turbine. Unit 2 reactor was being operated at approximately 95 percent RTP (rated thermal power) with no evolutions in progress. The Control Room received indication of a Main Turbine trip with both divisions of RPS (Reactor Protection System) actuated and all control rods inserted. Turbine bypass valves opened automatically to control reactor pressure and subsequently failed open causing the reactor to depressurize. When reactor pressure reached approximately 560 psig, the operations crew manually closed the Main Steam Isolation Valves (MISVs) to stop the depressurization. Reactor water level lowered to -31 inches causing Level 3 (+13 inches) isolations. No (automatic) ECCS (Emergency Core Cooling System) actuations occurred. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) were manually initiated to control reactor water level. The Operations crew subsequently maintained reactor water level at the normal operating band using RCIC and reactor pressure was controlled with HPCI in pressure control mode and main steam line drains. The Reactor Recirculation Pumps tripped as designed on EOC-RPT (end of cycle recirculation pump trip). The reactor is currently stable in Mode 3. An investigation into the cause of the turbine trip is underway. The NRC Resident Inspector was notified. A voluntary notification to PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A). |
| ENS 55416 | 18 August 2021 15:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 1036 CDT on 8/18/2021, Wolf Creek experienced a reactor trip due to low level in B Steam Generator. Auxiliary feedwater system actuated as designed. All systems actuated as expected. Decay heat is currently being removed by the auxiliary feedwater system. The NRC Senior Resident Inspector has been informed. All control rods fully inserted, and offsite power remained available. |
| ENS 55390 | 31 July 2021 21:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip of Unit 1 | At 1646 (CDT) on 7/31/21, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to control board indications of a Unit 1 'B' Main Feed Pump trip. After the reactor trip, one of the Condenser Steam Dump valves cycled to intermediate and remained stuck. The Condenser Steam Dump Valve was isolated locally using manual isolation valves. The 'B' Feed Regulating Bypass Valve did not control in automatic and was taken to manual to control the level in 'B' Steam Generator. The Auxiliary Feedwater System automatically actuated as designed when the valid actuation signal was received. Operations stabilized the plant in Mode 3. Decay heat is being removed by atmospheric dump valves due to condenser unavailability. Unit 2 is unaffected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. During the transient, all control rods inserted into the core. There is no known primary to secondary leakage. During the transient, no relief valves or safeties lifted. The plant is currently maintaining normal operating temperature and pressure with all safety equipment available. The plant is in its normal shutdown electrical lineup. |
| ENS 55380 | 26 July 2021 22:31:00 | Non | Reactor Shutdown Due to Control Rod Drive Failure | On July 26, 2021 at 1731 CDT, while the reactor was subcritical during a reactor startup, the University of Missouri-Columbia Research Reactor (MURR) was manually shut down due to the failure of the control rod drive mechanism for shim control blade B. MURR was not in compliance with one (1) Limiting Conditions of Operations (LCO). TS 3.2.a states, 'All control blades, including the regulating blade, shall be operable during reactor operation.' A spare control rod drive mechanism was installed for control blade B, post-installation operability testing was conducted satisfactorily, and permission from the Reactor Facility Director was obtained prior to the reactor returning to operation later on July 26, 2021. Currently, MURR is operating at 10 MW (full power). A detailed event report will follow within 14 days. |
| ENS 55370 | 21 July 2021 22:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor SCRAM | At 1826 EDT on July 21, 2021, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a trip of the Main Turbine. Unit 1 reactor was operating at 100 percent reactor power with no evolutions in progress. The Control Room received indication of a Main Turbine trip with both divisions of RPS (Reactor Protection System) actuated and all control rods inserted. The Reactor Recirculation Pumps tripped on EOC-RPT (end of cycle recirculation pump trip). Reactor water level lowered to +8 inches causing Level 3 (+13 inches) isolations. No ECCS (Emergency Core Cooling Systems) or RCIC (Reactor Core Isolation Cooling system) actuations occurred. The Operations crew subsequently maintained reactor water level at the normal operating band using Reactor Feed Water. The reactor is currently stable in Mode 3 with main condenser available. Investigation into the trip of the Main Turbine is in progress. The NRC Resident Inspector was notified. A voluntary notification to PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10CFR50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72(b)(3)(iv)(B). |
| ENS 55322 | 23 June 2021 03:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Rx Trip Due to Steam Leak in Moisture Separator Re-Heater Crossover Pipe | On June 22, 2021, at 2331 EDT, DC Cook Unit 2 Reactor was manually tripped due to a large steam leak in a crossover pipe of the Moisture Separator Re-heater (MSR) to the low pressure turbine. This notification is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation as a four (4) hour report, and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Auxiliary Feedwater System, as an eight (8) hour report. The DC Cook Resident NRC Inspector has been notified. Unit 2 is being supplied by offsite power. All control rods fully inserted. All Auxiliary Feedwater Pumps started properly. Decay heat is being removed via the Steam Dump System. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the Post Trip Review. No radioactive release is in progress as a result of this event. Unit 1 was not affected. |
| ENS 55239 | 6 May 2021 16:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip | On May 6, 2021 at 1223 (EDT), Unit 1 was manually tripped from 60 percent power due to degrading main condenser vacuum. Unit 1 was in the process of decreasing power due to increased secondary sodium levels identified earlier in the day. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. This reactor protection actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). Auxiliary Feedwater pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater pumps is reportable per 10 CFR 50,72(b)(3)(iv)(A) for a valid actuation of an ESF (Engineered Safeguards Features) system. Decay heat is being removed by the condenser steam dump system. The electrical system is in normal lineup for shutdown conditions. There was no effect on Unit 2 operation. The NRC resident inspector has been notified. |
| ENS 55172 | 6 April 2021 01:49:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Protection System (RPS) Actuation at Zero Percent Power | At 2149 EDT on April 5, 2021, with the power plant in Mode 2 at zero percent power, an actuation of the RPS system occurred following the decision to abort plant start-up. The reason for the RPS actuation was to align the plant to Mode 3, from Mode 2, following manually inserting all control rods using the Rod Control System. The RPS system initiated as designed when the mode switch was taken from 'Start-up' to 'Shutdown' to align the plant to Mode 3 from Mode 2. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
This is a retraction of an event notification made on 4/6/2021 at 0432 EST (EN#55172). This event was initially reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS System. This event was later determined to be pre-planned, in accordance with Technical Specifications, and not the result of a significant event, therefore not meeting the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). On the evening of April 4, 2021, while commencing reactor start up, it was determined that control rod withdrawal to add positive reactivity for the start-up would not overcome the negative reactivity of plant heat up. The control room team determined that the proper course of action would be to insert all control rods . The control room briefed and notified the Outage Control Center about its decision, then proceeded to insert all control rods. The control room manually inserted all control rods using the control rod hydraulic system. Following insertion of all control rods, the mode switch was taken to the shutdown position to meet the prerequisites of the procedure for maintaining hot shutdown. This action establishes Mode 3 in accordance with Technical Specifications and aligns the plant to perform the necessary work prior to a plant restart. By placing the mode switch in the shutdown position, a scram signal is generated for 10 seconds. NUREG-1022 offers guidance that states 'Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event.' The actions the operating crew took that night are accurately described by this statement in NUREG-1022 'shifting alignment of makeup pumps or closing a containment isolation valve for normal operational purposes would not be reportable.' In this situation, the Mode switch was taken to shutdown to align the plant to mode 3 for normal operational purposes, and not to mitigate a significant event. When the mode switch was taken to shut-down, RPS initiated as designed, there was no mis-operation or unnecessary actuation. This actuation was determined to be pre-planned, in accordance with Tech Specs, and not the result of a significant event, therefore not meeting the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident has been notified. Notified R3DO (McGraw). |