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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 54197 | 3 August 2019 07:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | En Revision Imported Date 8/7/2019 | EN Revision Text: AUTOMATIC REACTOR SCRAM ON LOW REACTOR WATER LEVEL At 0226 (CDT), an automatic scram on low reactor water level occurred due to a trip of the 'B' Reactor Feed pump. All control rods fully inserted. Reactor water level 2 was reached and the High Pressure Core Spray system, Reactor Core Isolation Cooling system, Division 3 diesel generator, Standby Gas Treatment Systems 'A' and 'B' and all shutdown safety related service water pumps started as expected. Reactor Core Isolation Cooling and High Pressure Core Spray injected as expected. All level 2 containment isolation signals occurred as expected and all level 2 containment valves closed as expected. Reactor water level is currently being controlled in band by condensate. Reactor pressure is being maintained by main turbine Bypass Valves. This event is being reported under 10 CFR 50.72(b)(2)(iv)(A), for ECCS discharge to RCS; 10 CFR 50.72(b)(2)(iv)(B), for RPS actuation, and 10 CFR 50.72(b)(3)(iv)(A), for specified system actuation. The NRC Senior Resident Inspector has been notified. No safety relief valves lifted during the transient. The plant is in a normal shutdown electrical lineup with all safety equipment available. The licensee notified the Illinois Emergency Management Agency per their communications protocol.
Following automatic initiation of the High Pressure Core Spray (HPCS) System as described above, the HPCS System was manually secured following station procedures after verification that additional RPV (reactor pressure vessel) injection was no longer required. Securing HPCS injection in this manner prevents automatic restart of the system in the event of a subsequent low RPV level condition, rendering it inoperable. As the HPCS system is considered a single train safety system, this meets the reportability requirements of 10 CFR 50.72(b)(3)(v)(D). This reportable condition was identified following review of post-scram actions. The HPCS system has been restored to a Standby lineup. The licensee will be notifying the NRC Resident Inspector. Notified R3DO (Pelke).
Following the scram, the Primary Containment to Secondary Containment and the Drywell to Primary Containment differential pressure limits were exceeded. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.4, Primary Containment Pressure, and 3.6.5.4, Drywell Pressure, Actions A.1, B.1, and B.2 were entered. Primary Containment to Secondary Containment differential pressure and Drywell to Primary Containment differential pressure were restored to within the LCO limits at 1505 on 8/3/19 and the associated TS Actions were exited. This event is reportable under 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that could have prevented the fulfillment of the primary containment function due to being outside the initial conditions to ensure that drywell and containment pressures remain within design values during a loss of coolant accident. This event is also reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of the drywell and primary containment functions to control the release of radioactive material for the same reason. The licensee notified the NRC Resident Inspector. Notified R3DO (Pelke). | Secondary containment Service water Reactor Core Isolation Cooling Primary containment High Pressure Core Spray Standby Gas Treatment System Safety Relief Valve Control Rod | |
| ENS 53623 | 26 September 2018 05:00:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Undervoltage Condition Initiates an Automatic Start of the Diesel Generator | At 0946 CDT on 9/26/2018, a disruption in power to the offsite 138 kV line and the subsequent trip of the Emergency Reserve Auxiliary Transformer (ERAT) Static VAR Compensator (SVC) resulted in a degraded voltage signal on the Division 1- 4.16 kV safety bus. The degraded voltage signal resulted in a trip of the ERAT feed to the bus, blocking closure of the 345 kV Reserve Auxiliary Transformer (RAT) feed to the bus and auto start of the Division 1 Emergency Diesel Generator (EDG). The Division 1 EDG successfully started and re-energized the Division 1- 4.16 kV bus as designed. The unit is stable with the Division 1 EDG carrying the Division 1- 4.16 kV bus. The Ameren Transmission System Operator in St. Louis, MO informed the station that they had received a report that a 138 kV to 13.8 kV transformer at Clinton Route 54 substation was on fire and the South feed to the Tabor substation cycled as a result of this fault. The NRC Resident Inspector and Illinois Emergency Management Agency Resident Inspector have been notified. | Emergency Diesel Generator | |
| ENS 47533 | 18 December 2011 15:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Rhr Cooling Due to Incorrect Reactor Water Level Indication | On 12/18/11 at approximately 0930 CST, with the plant in Mode 4, while Operations was lowering reactor water level as part of restoration activities following the Reactor Pressure Vessel hydrostatic test, an automatic reactor scram and Residual Heat Removal (RHR) 'A' pump trip occurred due to a valid Reactor Water Level Low (Level 3) signal. The cause of the low level appears to be due a significant disparity between the Upset and Shutdown Range level instrumentation and the Level 3 RPS (Reactor Protection System) instrumentation. No rod motion occurred during the reactor scram as all control rods were fully inserted at the time of the event. Operations immediately entered the station off-normal procedures for Loss of Shutdown Cooling and Reactor Scram. The 'A' RHR pump was restored at 0956 CST. Reactor coolant temperature increased from approximately 128.7 deg F to 131.7 deg F during the event. At 1112 CST, the scram was reset and at 1216 CST, the off-normal procedures were exited. Following the event as part of trouble shooting, maintenance personnel performed a fill and vent of reactor water level transmitter, 1B21N027. At the completion of this fill and vent, indicated water level changed from 195" to 86" on Shutdown Range and from off-scale high (>180") to 103"on Upset Range. No change in other reactor water level indication was observed. This event is reportable under 10 CFR 50.72 (b)(3)(v)(B), as an event or condition that could have prevented the fulfillment of a safety function needed to remove residual heat, and 10 CFR 50.72 (b)(3)(iv)(A), event or condition that results in a valid actuation of the reactor protection system (RPS). The licensee has notified the NRC Resident Inspector. | Reactor Protection System Shutdown Cooling Reactor Pressure Vessel Residual Heat Removal Control Rod | 05000461/LER-2011-008 |
| ENS 40869 | 14 July 2004 05:45:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps and Containment Isolation Actuation Due to Low Rpv Level | On July 14, 2004, at 0045 CDT, the Clinton Power Station experienced a drop in RPV (Reactor Pressure Vessel) level from nominally 30 inches to the Level 3 (the RPS and Isolation setpoint). At this time control rods were fully inserted (see Event Notification - #40868), however, the containment isolation valves associated with this setpoint were open, and these valves automatically closed successfully. Level was restored to normal by automatic makeup from the condensate system. There were no other malfunctions. The cause for the level drop is being investigated. The licensee will notify the NRC Resident Inspector. | Control Rod | |
| ENS 40368 | 2 December 2003 22:58:00 | 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 Low Reactor Water Feed Pump Suction Pressure | On December 2, 2003, at 1658, a manual scram was inserted from 88% power, which is the maximum power due to the plant being in coastdown. The manual scram was inserted due to feedwater suction pressure being below the trip set point for the operation of the Turbine Driven Reactor Feed Pumps and Reactor Pressure Vessel water level trending down. The initiating event was the loss of the 480V Unit Sub 1I. All plant systems operated normally on the scram with the exception of those systems that lost power due to the tripping of 480V Unit Sub 1I. The plant is shutdown at 0% power in Mode 3, maintaining pressure between 800 and 1065 psig, and reactor water level between level 3 and 8. Troubleshooting is in progress on the cause of the loss of the 480V Unit Sub 1I." Plant is identifying the loads lost during the loss of the 480V Unit Sub 1I and the cause of the low feedwater suction pressure. No additional specified system actuations occurred. No SRVs lifted and the electrical plant lineup is stable and in a normal lineup for plant conditions with the exception of the lost 480V bus. The licensee has notified the NRC Resident Inspector.
As a consequence of the manual reactor scram inserted at 1658 CST, reactor water lowered to Level 3. This level is a valid RPS actuation and containment isolation signal (Groups 2, 3 and 20). These actuations are being reported consistent with 10CFR50.72(b)(3)(iv)(A). The Level 3 RPS actuation is expected during a high power reactor scram. Additionally, on December 2, 2003, at 1935 CST, a Level 3 RPS actuation reoccurred while transferring reactor coolant makeup from the Motor Driven Reactor Feed (MDRFP) to a Condensate/Condensate Booster Pump pair. Reactor pressure was being lowered to the discharge pressure of the Condensate/Condensate Booster pumps. This Level 3 is a valid RPS actuation and containment isolation signal. These actuations are being reported consistent with 10CFR50.72(b)(3)(iv)(A). The plant is shutdown at 0% power in Mode 3, maintaining pressure between 550 and 750 psig, and reactor water level between level 3 and 8. Troubleshooting is in progress on the cause of the loss of the 480V Unit Sub 1I. Both of these events will be reported as a single Licensee Event Report (LER). The plant continues to remove decay heat via the turbine bypass valves to the main condenser. The licensee will notify the NRC Resident Inspector. Notified R3DO Bruce Burgess. | Feedwater Reactor Pressure Vessel Main Condenser |