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| Report date | Site | Event description | |
|---|---|---|---|
| 05000373/LER-2017-007 | 18 August 2017 | Lasalle LaSalle | On June 22, 2017, the Unit 1 Low Pressure Core Spray (LPCS) system was declared inoperable due to loss of corner room area cooling and loss of motor cooling. The common diesel generator cooling water pump received an automatic trip signal while being secured. The LPCS pump remained in standby during the event. Troubleshooting identified the most likely cause for the trip of the emergency cooling water pump breaker was a malfunction of the cooling water pump control switch or the cooling water supply fan control relay. Both suspected components were replaced. The causal investigation did not identify a specific cause; however, there is a high level of confidence that the failure modes were eliminated by the corrective actions taken during troubleshooting. This component inoperability is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. This condition could have prevented the LPCS system, a single train safety system, from performing its design function. There was minimal safety consequences associated with the condition since other required emergency safety systems remained operable, there were no actual demands for Unit 1 LPCS, and safety margins were maintained. |
| 05000374/LER-2017-003 | 9 August 2017 | Lasalle LaSalle | On February 11, 2017, Unit 2 was in Mode 5 for a planned refueling outage. While attempting to fill and vent the Unit 2 High Pressure Core Spray (HPCS) system, no flow was observed from the drywell vent valves or downstream of the HPCS injection valve. The HPCS system was already inoperable to support scheduled surveillances performed on February 8, 2017 in which the HPCS injection isolation valve had been cycled five times satisfactorily. Troubleshooting determined the cause of the valve malfunction was due to stem-disc separation. The valve internal components were replaced prior to restart of the unit from the refueling outage. The root cause of the valve failure was insufficient capacity of the shrink-fit stem collar, combined with multiple high-load cycles, which resulted in loosening and eventual shear failure of the wedge pin and threads. This component failure is reported in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. This condition could have prevented the HPCS system, a single train safety system, from performing its design function if the valve failure occurred during an actual demand. This component failure is also reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specifications (TS) 3.5.1 "ECCS - Operating," since the HPCS system could have been 1 inoperable for greater than the TS 3.5.1, Required Action B.2, Completion Time of 14 days to restore HPCS system to operable status. There were minimal safety consequences associated with the condition since HPCS was not required to be operable at the time of the failure, and other required emergency safety systems remained operable. There were no actual demands for Unit 2 LHPCS, other ECCS systems, or the reactor core isolation cooling (RCIC) system during this period. - --- ------- - NRC FORM 366 (04-2017) - 01 003 2017 |
| 05000373/LER-2017-006 | 17 July 2017 | Lasalle LaSalle | On May 17, 2017 at 0908 CDT, during Unit 1 full-power operations, operators received an unexpected alarm for the Low Pressure Core Spray (LPCS) pump injection high flow and automatic closure of the LPCS minimum flow valve (1E21-F011). Inspections indicated the flow switch that actively controls the LPCS minimum flow valve had a faulty diaphragm which allowed for water intrusion into the device. There were no impacts on plant operations. The required actions of Technical Specifications 3.5.1, "ECCS - Operating" and TS 3.3.5.1, "Emergency Core Cooling System (ECCS) Instrumentation" were entered. The switch was replaced and LPCS system tested, which allowed full restoration of the system on May 17, 2017 at 18:45 CDT. This condition could have prevented the LPCS system from performing its design function. This condition is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. There was minimal safety consequences associated with the condition since other emergency safety systems remained operable. |
| 05000374/LER-2017-004 | 14 July 2017 | LaSalle | During the February 2017 Unit 2 refueling outage, two main steam safety relief valves (SRV) did not pass Technical Specification (TS) Surveillance Requirement 3.4.4.1 and Inservice Testing (1ST) Program lift pressure requirements. Both SRVs (2B21-F013C and 2B21-F013L) lifted below their expected lift pressures. On February 16, 2017, SRV 2B21-F013C was required to lift within plus or minus three percent of 1175 psi (i.e., 1175 psi plus or minus 35.2 psi), but actually lifted at 1131 psi. On February 17, 2017 SRV 2B21-F013L was required to lift within plus or minus three percent of 1195 psi (i.e., 1195 psi plus or minus 35.8 psi), but actually lifted at 1130 psi. Multiple test failures are reportable under 10 CFR 50.73(a)(2)(i)(B) as an operation or condition prohibited by the plants Technical Specifications. Both SRVs lifted prior to their expected lift pressures, which is conservative in regards to maintaining reactor pressure vessel over-pressure limits. Both SRVs were replaced during the outage. A failure analysis was conducted; however, it did not identify a cause for the valves lifting below their expected lift set-points. |
| 05000373/LER-2017-005 | 18 April 2017 | Lasalle LaSalle | On February 17, 2017, at 2353 CST, during Unit 1 power ascension from a previous forced shutdown, operators inserted a manual scram as a result of a high reactor water level condition caused by a rapid change in feedwater flow. The high reactor water level trip occurred due to a failure of the feedwater regulating valve (FRV) 1FW005 positioner arm, which caused the regulating valve to be driven to the full open position. This resulted in a rapid increase of reactor water level that required operators to perform a manual reactor scram. The plant was placed in a stable condition, with no complications in achieving shutdown. This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of the reactor protection system (RPS). The safety significance of this condition was minimal, as plant equipment responded as expected for the event. The feedback drive arm and new positioner were installed on the Unit 1 FRV, which allowed the unit to restart following repairs. The FRV positioner feedback assembly had been installed in 2002, during a modification to upgrade of the air-operated valve assembly. This design has proven not to be robust enough for the FRV application. |
| 05000373/LER-2017-004 | 17 April 2017 | LaSalle | On February 16, 2017, Unit 1 was in Mode 2 for startup at five percent power, and Unit 2 was defueled for a planned refueling outage with movement of irradiated fuel (MIF) in progress. At 0835 CST, both air-lock doors of the Unit 1 reactor building to the chemistry corridor were opened simultaneously for approximately five seconds during personnel ingress. The employee immediately secured both doors in the interlock and notified the Main Control Room. While both interlock doors were open, Technical Specifications (TS) Surveillance Requirement (SR) 3.6.4.1.2 to verify one secondary containment access door in each access opening is closed was not met. Secondary containment was declared inoperable for the period of time that both interlock doors were open. TS 3.6.4.1 Required Action (RA) A.1 to restore secondary containment to operable status within four hours was entered and exited. TS 3.6.4.1 RA C.1 to immediately suspend MIF in secondary containment was entered and exited. This event is reportable in accordance with 10 CFR 50.73(a)(2)(v)(C) and 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material or mitigate the consequences of an accident. The most probable cause of the interlock failure was the intermittent failure of a circuit board which is designed to prevent more than one door to be open at a time. Previous events and prior causal investigations have indicated degradations in the relays for these circuit cards, and the corrective actions to replace interlock door circuit cards was ongoing at the time this event occurred. The door interlock was satisfactorily tested by maintenance technicians, and the door interlock was returned to service on February 18, 2017. |
| 05000373/LER-2017-003 | 14 April 2017 | Lasalle LaSalle | On February 13, 2017, LaSalle County Station Unit 1 was in Mode 1 at 100 percent power and Unit 2 shut down for a planned refueling outage. At 2309 CST, a reactor scram signal was received on Unit 1 due to turbine control valves fast closure while the station was aligning back-feed operation to the Unit 2 main power transformer (MPT). The Unit 1 turbine trip was due to the main generator trip on differential current. The plant was placed in a stable condition with reactor pressure maintained by the turbine bypass valves and reactor water level controlled using feedwater. Unit 2 was unaffected by the event. The root cause of the Unit 1 trip on differential current was a marginal generator differential relay design that was prone to responding to faults outside its zone of protection. Both units' 345 kV ring buses were connected together through the cross-tie bus, which allowed the Unit 1 generator to supply some of the electrical current that resulted in its differential circuit to create an unbalanced current that actuated the differential relay. This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of the reactor protection system (RPS), including reactor scram. There were no safety consequences associated with the event since RPS and other emergency safety systems functioned as designed. |
| 05000374/LER-2017-002 | 31 March 2017 | Lasalle LaSalle | On January 30, 2017, during routine surveillance testing of the Unit 2 Division 3 Diesel Generator Cooling Water (DGCW) system, the cooling water strainer backwash valve was unable to open. The Division 3 DGCW system was declared inoperable. Upon investigation, operators determined the cause of the valve malfunction was due to stem-disc separation. Division 3 DGCW is a support system for the Division 3 Emergency Diesel Generator and the High Pressure Core Spray (HPCS) system. The required actions of Technical Specifications (TS) 3.7.2 and 3.5.1 were entered on January 30, 2017 when the DGCW and HPCS system, respectively, were determined to be inoperable. TS 3.7.2 Required Action (RA) A.1 requires the supported system to be immediately declared inoperable. TS 3.5.1 RA B.2 requires restoration of the HPCS system to operable within 14 days. TS 3.8.1 was not applicable since a note provides that Division 3 AC electrical power sources are not required to be operable when HPCS is inoperable. The valve was replaced, and the HPCS system was returned to operable on February 2, 2017. This condition could have prevented the HPCS system, a single train safety system, from performing its design function. This event is reportable in accordance with 10 CFR 50.73(a)(2)(v)(D) as an event or condition that could have prevented fulfillment of the safety function of structures or system that are needed to mitigate the consequences of an accident. There were minimal safety consequences associated with the event since the other emergency safety systems remained operable, and the Division 3 DGCW system remained functional as it retained the ability to provide the required flow through the system. The apparent cause of the stem-disc separation was erosion due to the carbon-steel valve internals in a raw water system environment. |
| 05000374/LER-2017-001 | 24 March 2017 | Lasalle LaSalle | On January 23, 2017, operators initiated a manual scram of the LaSalle County Station Unit 2 reactor as a result of observing a generator run-back due to a generator stator winding cooling (GC) system malfunction. Initial troubleshooting identified the most likely cause was plugging in the 'A' GC heat exchanger, based on inspection of the GC system flow-path components. The GC system was realigned to the 'B' heat exchanger until inspections could be performed in the upcoming refueling outage, and the unit was re-started on January 24, 2017. Further inspections of the GC components were performed while the unit was shut down for a planned refueling outage. These inspections determined the cause of the GC system failure was stem-disc separation in the 'A' GC heat exchanger inlet valve. The valve was repaired during the refueling outage. This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event or condition that resulted in manual or automatic actuation of the Reactor Protection System (RPS). There were no safety consequences associated with the event since there was no loss of safety function, and the RPS functioned as designed. |
| 05000373/LER-2017-001 | 8 February 2017 | Lasalle LaSalle | On October 18, 2016, the Unit 1 Reactor Core Isolation Cooling (RCIC) system tripped on low suction pressure during a normal system start following completion of scheduled maintenance activities. The system was restored to operable on October 20, 2016. A second event involving a Unit 1 RCIC system trip on low pressure suction pressure occurred on November 17, 2016, during the system's quarterly operability surveillance. The system was restored to operable on November 20, 2016. The component failure analysis completed on December 16, 2016, determined the cause of both Unit 1 RCIC system trips was a failure of the electronic governor-remote (EG-R) hydraulic actuator. The Unit 1 RCIC inoperable period was from the first system trip on October 18, 2016, to when full restoration was completed on November 20, 2016. This time was greater than allowed by Technical Specifications (TS) 3.5.3, "RCIC System," Condition A Completion Time of 14 days. This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by the plant's TS. The root cause for the low suction pressure trips was inadequate management of the EG-R preventative maintenance (PM) strategy. Corrective actions included replacement of the EG-R and a plan to implement an appropriate PM strategy for the RCIC EG-R. The safety consequences were minimal since the RCIC system is not credited in the safety analysis, and the credited High Pressure Core Spray (HPCS) system remained available to provide its safety function. |
| 05000374/LER-2016-001 | 18 April 2016 | Lasalle LaSalle | On February 17, 2016, Unit 2 was in Mode 1 at 100 percent power and Unit 1 was in Mode 5 for refueling outage L1R16 with no fuel movements or operations with the potential to drain the reactor vessel (OPDRV) in progress. At 1035 hours CST, both air-lock doors of the Unit 2 Chemistry Lab corridor / Unit 2 Reactor Building interlock were open at the same time for approximately five seconds. While both interlock doors were open, Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.2 to verify one secondary containment access door in each access opening is closed was not met for Unit 2. Secondary containment was declared inoperable for the period of time that both interlock doors were open. TS 3.6.4.1 Required Action (RA) A.1 to restore secondary containment to OPERABLE status within four hours was entered and exited. The cause was failure of the relays on the door controller circuit card. The controller circuit card was replaced, which restored the interlock functionality. Corrective actions include determining the cause of vendor quality issues with the controller circuit card relays and procurement of a more reliable controller circuit card following the identification of the cause of the relay failures from vendor analysis. |
| 05000373/LER-2016-001 | 11 April 2016 | Lasalle LaSalle | On February 10, 2016, Unit 1 was in Mode 1 at 91 percent power and Unit 2 was in Mode 1 at 100 percent power. At 2207 hours CST, it was reported that the Unit 1 reactor building ventilation exhaust damper 1VRO5YA failed and began to show dual indication. As a result, the Unit 1 reactor building ventilation exhaust fans tripped off, causing a positive reactor building differential pressure on both units. The damper and secondary containment were declared inoperable, and Technical Specification 3.6.4.1 Required Action A.1 was entered on both units to restore secondary containment to operable status within four hours. In addition, Technical Specification 3.6.4.2 Required Action A.1 was entered to isolate the penetration with one closed or deactivated automatic valve within eight hours. The cause was an intermittent failure of a solenoid on one of the two half damper blades on the 1VRO5YA exhaust isolation damper. This led to the exhaust damper blade half intermittently changing its position, which resulted in secondary containment pressure going positive. The solenoid valves on both halves of the 1VRO5YA exhaust damper were replaced, and the failed solenoid was sent to a vendor for failure analysis. |