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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| 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. | Reactor Protection System Control Rod | |
| ENS 53903 | 1 March 2019 04:17:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Trip of Safety Related Bus | On February 28, 2019, at 2217 CST, LaSalle Unit 2 experienced a trip of the 241Y Safety Related Bus during surveillance testing resulting in a valid undervoltage actuation signal to the Common Emergency Diesel Generator ('O' EDG), causing it to start and load to Bus 241Y. The purpose of the surveillance testing was to demonstrate the operability of the breakers necessary to provide the second off site source to Unit 2. This event is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A), as an event that results in a valid actuation of the emergency AC electrical power system. In addition to the 241Y bus trip and 'O' EDG actuation signal, the following plant responses occurred as designed due to the momentary loss of this AC Bus: "A" RPS de-energized due to the loss of the 2A Reactor Protection System Motor-Generator Set, and the running Unit 2 Fuel Pool Cooling pump tripped. The Non-Safety Related Bus 241X de-energized resulting in a trip of the Unit 2 Station Air Compressor. All systems have been restored and troubleshooting is currently in progress. Unit 1 remained in MODE 1 during this event. The NRC Senior Resident Inspector has been notified. | Reactor Protection System Emergency Diesel Generator | |
| ENS 50346 | 5 August 2014 22:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | This notification is being provided pursuant with SAF 1.6 10CFR50.72(b)(2)(iv)(B) and SAF 1.7 10CFR50.72(b)(3)(iv)(A). At 1734 CDT on August 5, 2014, LaSalle Unit 2 automatically scrammed due to an RPS actuation. The MSIVs isolated on a Group 1 signal, the cause is under investigation. The reactor water cleanup system isolated during the transient. The plant is stable with Reactor Pressure Control being maintained by the Reactor Core Isolation Cooling System and SRVs and level being controlled by the Low Pressure Core Spray System. The plant is planned to remain in hot shutdown pending investigation of the trip." The Unit 2 electric plant is in a normal shutdown lineup. All control rods inserted fully on the scram. Unit 1 was not affected by the Unit 2 transient. The licensee notified the NRC Resident Inspector.
The initial notification to the NRC stated that the reactor water cleanup system had isolated during the transient. The actual status is being corrected to state that the reactor water cleanup pump tripped during the transient. The licensee has notified the NRC Resident Inspector. Notified R3DO (Stone). | Reactor Core Isolation Cooling Core Spray Reactor Water Cleanup Control Rod | 05000374/LER-2014-001 |
| ENS 48939 | 17 April 2013 20:11:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Notification of Unusual Event Declared Due to Loss of Offsite Power from a Lightning Strike | LaSalle Unit 1 and LaSalle Unit 2 have both experienced an automatic reactor scram, in conjunction with a loss of offsite power. This was caused by an apparent lightning strike in the main 345kV/138kV switchyard during a thunderstorm. 138kV line 0112 has been inspected in the field, and heavy damage has been noted on the insulators on two of the three phases on a line lightning arrestor line side. The plant systems have all responded as expected. All five diesel generators started, and have loaded on to their respective buses as designed. All rods went full in on both units during the respective scrams. HPCS (High Pressure Core Spray) system was started on each unit and automatically aligned for injection for initial level control. The MSIVs (Main Steam Isolation Valves) are shut on both units with decay heat being removed via the safety relief valves. Suppression pool cooling is in progress. The licensee will notify the NRC Resident Inspector and has notified the State. Notified DHS, FEMA, USDA, HHS, DOE, NICC, EPA, and Nuclear SSA via email.
In addition to information (previously provided), LaSalle Unit 2 received a high drywell pressure signal (1.77 psig) due to loss of containment cooling from the loss of power. At the time of this high drywell pressure signal, high pressure core spray pump and 2B residual heat removal (RHR) pump was already in operation, the low pressure core spray system and 2A residual heat removal system was secured and (placed) in pull to lock. When the signal was satisfied the ECCS (Emergency Core Cooling Systems) signal was processed but only the 2C RHR pump would have started. In this case, the 2C RHR pump tripped when the signal was received. There is no evidence of reactor coolant leakage. There was no additional ECCS systems discharging into the RCS (Reactor Coolant System). As (initially stated), level was controlled using High Pressure Core Spray and level control is now being maintained using the Reactor Core Isolation Cooling (RCIC) systems. The 2C RHR pump trip is under investigation. Due to the initial loss of offsite power for both Unit 1 and Unit 2 reported at 1511 (CDT), multiple containment isolation valves isolated and closed as expected. Once initial containment isolations were verified, two Unit 2 primary containment vent and purge valves were opened to vent the Unit 2 containment. Once Unit Two containment pressure reached 1.77 (psig), these two vent valves isolated as expected. Due to the loss of offsite power, the Station Vent Stack Wide Range Gas Monitor (WRGM) and the Standby Gas Treatment Wide Range Gas Monitor (VGWRGM) also lost power. Manual sampling has been implemented and power is restored to the VGWRGM, however the VGWRGM has not been declared operable yet. Normal radiation levels have been reported from the manual sampling. (This is being reported in accordance with 10CFR50.72(b)(3)(xiii).) The licensee notified the NRC Resident Inspector and the State of Illinois. Notified the R3 IRC, NRR EO(Skeen), IRD MOC (Grant).
After the Unit 2 primary containment vent and purge system isolated on the Unit 2 containment High Pressure signal, Venting of the Unit 1 primary containment was commenced. At 2005 CDT, Unit 1 primary containment pressure reached the Group 2 primary containment isolation system setpoint (1.77 PSIG) causing the primary containment vent and purge valves being used to vent the Unit 1 containment to isolate. Unit 1 primary containment venting was being performed through the Standby Gas Treatment system which is a filtered system. In addition to the primary containment isolation signal on high drywell pressure, an ECCS initiation on high drywell pressure also occurred. The ECCS signal resulted in an auto start of the 1C RHR system. The 1B RHR system was already running in suppression pool cooling mode. 1A RHR and LPCS had been secured to prevent overloading the common diesel generator for division 1. The common diesel generator supplies both Unit 1 and Unit 2 division 1 ESF busses. The licensee informed the NRC Resident Inspector. Notified NRR EO (Skeen), IRD MOC (Grant) and R3IRC (Louden).
LaSalle has terminated the unusual event which was initiated at 1511 on 4/17/13 and reported under EN 48939. This unusual event has been terminated based on meeting the following established criteria. This report is being made in accordance with 10CFR50.72.(c)(1)(iii). 1) Off-site power has been restored to all ESF busses 2) Fuel Pool Cooling has been restored on both units 3) Primary Containment Chillers have been restored on both units 4) Drywell pressure is less than ECCS initiation setpoint 5) ECCS signals cleared to allow diesels to be placed in stand by Recovery of remaining plant systems will be managed through the Outage Control Center (OCC)." The licensee informed the NRC Resident Inspector. Notified R3DO (Orth), NRR EO (Chernoff), IRD (Grant), DHS, FEMA, USDA, HHS, DOE, NICC, EPA, and Nuclear SSA via email.
In addition to the 10 CFR 50.72 Sections initially identified, the Loss of Offsite Power was also reportable under 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of systems needed mitigate the consequences of an accident. This event is considered a safety system functional failure for both Units 1 and 2. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Orth). | Primary Containment Isolation System Reactor Core Isolation Cooling Primary containment High Pressure Core Spray Core Spray Residual Heat Removal Standby Gas Treatment System | 05000373/LER-2013-002 |
| ENS 42348 | 20 February 2006 06:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Site Area Emergency - Reactor Scram | While shutting down Unit 1 IAW LGP-2-1 with the main turbine off line the plant experienced a turbine control system (EHC) malfunction. This resulted in opening all main turbine bypass valves and subsequent reactor low pressure condition. The low pressure condition resulted in a closure of all main steam isolation valves (MSIV's) and automatic reactor scram. Three rods failed to indicate fully inserted on the scram. Plant emergency operating procedures were entered. All ECCS and plant systems operated as expected. All control rods are fully inserted and the ATWS emergency operating procedure has been exited. The main turbine had been taken off line prior to this event and shutdown was being conducted with heat removal on the bypass valves. Pressure control is currently by using the steam line drains with the Safety Relief Valves in manual. Level is being maintained using normal feedwater. The electrical grid is stable though the plant is not yet on backfeed. Diesel generators are operable. The licensee will notify the NRC Resident Inspector. The NRC entered Monitoring Mode at 0243. Notified R4DO (Graves), DHS (SWO), FEMA (J Kanupp), DOE (S. Morrone), EPA(NRC) (Nowak), USDA (Amanda), HHS (Kleiman).
In addition to the initial notification, La Salle is reporting a group 1 isolation in accordance with 10CFR50.72(b)(3)(iv)(a).
The site exited the Site Area Emergency and entered the recovery phase.
At the time of the scram @ 0023 hours rod 38-43 showed position 24 and rods 26-15 and 34-47 showed unknown. Based on more than one rod out condition, it is unanalyzed until shutdown margin can be verified.
NRC exited Monitoring Mode. Notified IRD (Wilson), NRR EO (J. Lyons), R3DO (Lara), R4DO (Graves), NRR (Dyer), DHS (SWO), FEMA (Eerwin), DOE (Joe Stambaugh), NRC (does not take updates), USDA (Jim Brzostek), HHS (SOC) (Lt Hrynyshen).
Post trip evaluations have confirmed that all control rods were fully inserted within four minutes of the reactor scram. A review of the post-trip data suggests that there were only control rod indication problems on the three subject rods and all control rods were fully inserted immediately at the time of the reactor scram. Follow-up evaluations also demonstrated that even if the three subject control rods remained fully withdrawn in a cold shutdown condition, the reactor would have remained adequately shutdown. Additional confirmatory evaluations are continuing. The licensee issued a press release on this event. The licensee notified the NRC Resident Inspector. R3DO(Hills) has been notified. | Feedwater Main Steam Isolation Valve Main Turbine Safety Relief Valve Control Rod | 05000373/LER-2006-001 |
| ENS 41787 | 22 June 2005 04:40:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Multiple Specified Systems Actuations Due to Loss of Ac Safety Related Buses | At 23:40 CDST, LaSalle Unit 2 experienced a trip of the 4160 Volt AC feed breaker to 480 Volt AC Safety Related buses 235X & 235Y. As a result of this feed breaker trip, containment isolation valves closed in multiple systems. The systems which isolated included Containment Monitoring System, Drywell Floor Drains and Drywell Equipment Drains, Reactor Recirculation Flow Control Hydraulic System, Drywell Instrument Nitrogen System, Reactor Water Cleanup System, and Reactor Recirculation Sample System. This event is reportable under 10 CFR50.72(b)(3)(iv)(A). In addition to the- above mentioned system isolations, the following plant responses occurred due to the loss of these 480 Volt AC buses: Reactor Building Ventilation was lost due to the closure of the secondary containment isolation dampers; Multiple Division 1 containment isolation valves lost their AC power source, a 1/2-SCRAM occurred due to the loss of the 2A Reactor Protection System Motor-Generator set, the DC charger feed to the Division 1 DC Battery/Bus was lost, and the battery is currently maintaining availability of DC power to the bus. Troubleshooting has determined that the cause of the breaker trip was due to a defective neutral ground current protective relay. This relay has been replaced and the buses have been re-energized. System restoration is currently in progress. The licensee informed the NRC Resident Inspector.
The following is an update to EN# 41787. Restoration of all required safety systems with the exception of the Reactor Core Isolation Cooling (RCIC) system has been completed. The associated ECCS systems were filled, vented and restored to an available status at 0630 CDST and full operability established at 0803 CDST. The Division 1 DC battery was fully recharged and returned to an operable status at 0803 CDST. Restoration of the RCIC system and investigation into the cause of the relay failure is currently in progress. The licensee has notified the NRC Resident Inspector. Notified R3DO (Bruce Burgess). | Secondary containment Reactor Protection System Reactor Core Isolation Cooling Reactor Building Ventilation Reactor Water Cleanup | 05000374/LER-2005-003 |
| ENS 40495 | 2 February 2004 04:34:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | One Half Scram Signal Due to the Loss of the "B" Reactor Protection System Bus. | This report is being made pursuant to 10 CFR50.72(b)(3)(iv)(A), system actuation not including Reactor Protection System; due to a trip of the "D" Electric Power Monitoring Assembly on the "B" Reactor Protection System Motor Generator set resulting in a trip of the "B" Reactor Protection System bus. The loss of "B" Reactor Protection System bus caused the actuation of multiple Primary Containment Isolation Systems. Station abnormal procedures have been entered and the plant stabilized. The system actuations that occurred have been verified and restorations of these systems are in progress. Troubleshooting plans are being developed to determine cause of the trip and to correct the deficient condition. The NRC Resident Inspector was notified of this event by the Licensee. | Reactor Protection System Primary Containment Isolation System | 05000374/LER-2004-001 |