Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 53409 | 17 May 2018 23:02:00 | On May 17, 2018, with the Unit in Mode 4, Clinton Power Station experienced the concurrent inoperability of two Emergency Diesel Generators (DG). This event is being reported as an 8-hour non-emergency notification per 10 CFR 50.72(b)(3)(v) as, 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.' On May 17, 2018, at 15:03 CDT, it was identified that the Division 2 DG air start receiver isolation valves 1DG160 and 1DG161 were shut. With these valves shut, the Division 2 DG is inoperable. At the time, Division 1 DG was also inoperable for a 4160V 1A1 partial bus outage. The Technical Specification (TS) Actions for TS 3.8.2, AC Sources Shutdown, and TS 3.5.2, RPV Water Inventory Control, were entered as a result of the inoperability of the onsite AC power sources to isolation valves being credited to limit RPV DRAIN TIME. Division 2 DG air receivers were realigned at 15:45 CDT and the Division 2 DG auto start function was restored. Offsite power was available throughout this event and there was no impact to the health and safety of the public or plant personnel. Investigation is ongoing. It has been determined that the air start receiver isolation valves remained closed following system restoration on May 11, 2018. However, the Division 2 DG was not required to be Operable by TS until 00:45 CDT on May 14, 2018 when the Division 1 DG was made inoperable as a result of scheduled plant maintenance. The NRC Resident Inspector has been notified. |
| ENS 53110 | 9 December 2017 18:42:00 | At approximately 1347 (CST) on 12/09/17, the Main Control Room received annunciators that indicated a trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1 breaker. Numerous Division 1 components lost power (powered from unit subs 1A and A1). The Division 1 containment Instrument Air isolation valves had failed closed by design due to the loss of power. Due to the loss of containment instrument air, several control rods began to drift into the core as expected and, by procedure, the reactor mode switch was placed in the shutdown position at 1353 (CST). All control rods fully inserted. Also due to the loss of power, the Fuel Building ventilation dampers failed closed by design. With the normal ventilation system secured, secondary containment differential pressure rose to slightly greater than 0 inches water gauge which exceeded the Technical Specification requirement of greater than 0.25 inches vacuum water gauge at 1348 (CST). The Control Room entered EOP-8, Secondary Containment Control. Secondary Containment differential pressure was restored within Technical Specification requirements at 1351 (CST) by starting the Division 2 Standby Gas Treatment system. This event is being reported as a manual actuation of the Reactor Protection System (RPS) and as a Condition that Could Have Prevented Fulfillment of a Safety Function. The cause is currently under investigation. The NRC Resident has been notified. The licensee informed the NRC Resident Inspector.
During a review of plant logs it was identified that the primary to secondary containment differential pressure was identified to be outside of Technical Specification 3.6.1.4 limits of 0 plus or minus 0.25 psid at 2009 on 12/9/17 due to the primary containment ventilation system dampers closing as a result of the loss of power. This parameter is an initial safety analysis assumption to ensure that primary containment pressures remain within the design values during a Loss of Coolant Accident (LOCA). As a result, this condition is reportable as an unanalyzed condition that significantly degrades plant safety. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone).
During the post transient review of the trip of the 4160 V 1A1 breaker 1AP07EJ, 480V XFMR 1A and A1, it was identified that the unplanned INOPERABILITY of the Low Pressure Core Spray (LPCS) system due to the loss of power to the injection valve constitutes an event or condition that could have prevented fulfillment of a safety function and is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. The High Pressure Core Spray (HPCS) remained available to perform the core spray function, if necessary, during a design basis Loss of Coolant Accident (LOCA), however HPCS and LPCS are each considered single train safety systems. The NRC Senior Resident Inspector has been notified. Notified the R3DO (Stone). |
| ENS 51939 | 17 May 2016 18:43:00 | On May 17, 2016 with the plant in Mode 4 (Cold Shutdown) during a refueling outage, personnel entered the drywell to perform a walkdown. At 0945 CDT, water was identified leaking from flexible hoses located at the inner elbow of MSL (Main Steam Line) B and MSL C. It was concluded that the leakage was from an elbow tap welded to the flexible hoses associated with flow instrumentation on MSL C and MSL B. Due to the refueling outage, the plant subsequently entered Mode 5 at 0955 and is currently in Mode 5 (Refueling) and 0 percent rated thermal power. The degraded component on MSL B was previously replaced in 2008 and on MSL C in 2007. The station has determined that this event is reportable under the provisions of 10 CFR 50.72 (b)(3)(ii)(A) as an event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers being seriously degraded, as an 8-hour notification. The NRC Resident Inspector has been notified. |
| ENS 47544 | 21 December 2011 19:56:00 | At 1641 CST, during preparations to synchronize the generator, hydrogen pressure began to lower. Personnel were dispatched to identify the source of the leak. At 1752 CST reports from personnel in the turbine building identified high hydrogen levels and the turbine building was evacuated. An Unusual Event (HU7) was declared at 1757 CST and state notification was completed at 1809 CST. Turbine building roof vents were opened to ventilate the turbine building and all elevations were verified to be clear of hydrogen. Access to the turbine building was restored. The Unusual Event was terminated at 1838 CST. The licensee has notified the NRC Resident Inspector. |
| ENS 47533 | 18 December 2011 18:08:00 | 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. |
| ENS 47587 | 12 January 2012 11:36:00 | The following information is provided as a sixty (60) day telephone notification to the NRC in accordance with 50.73 (a)(1) in lieu of the submittal of a written LER to report a condition that resulted in an invalid actuation of general containment isolation signals affecting containment isolation valves in more than one system and emergency service water systems that normally do not run and that serve as ultimate heat sinks. NUREG-1022, Revision 2, identifies the information that needs to be reported as discussed below. On December 4, 2011, Clinton Power Station was in Mode 5 (Refueling) conducting refueling outage activities, including the installation of a self test system power supply modification. During post-maintenance activities associated with the affected power supplies, an unexpected voltage reading was identified. While investigating the anomaly, one of the two nuclear system protection system essential logic power supplies was de-energized with the redundant power supply not yet restored. At 0609 (CST), the power supply was re-energized, resulting in an inadvertent actuation of the logic. Containment isolation valves on the following systems automatically closed: Instrument Air, Service Air, and Containment Floor Drain. Additionally, the Division 1 Shutdown Service Water (SX) Pump started. These actuations were not valid since the actuations were associated with low reactor water level conditions (Level 1 and Level 2) which were non-existent at the time of the event. The associated actuations were successful had the actuations been valid. This event is reportable in accordance with 10CFR50.73 (a)(2)(iv)(A) and has been entered into the Corrective Action Program. The licensee notified the NRC Resident Inspector. |
| ENS 43964 | 6 February 2008 12:56:00 | On February 6, 2008 at 1119 hours, Clinton Lake level reached 694 feet 0 inches (mean sea level). Normal lake level is 690 feet 0 inches. Per Clinton Power Station (CPS) procedure 4303.02, 'Abnormal Lake Level,' the CPS Main Dam - Emergency Action Plan requires that a 'Flood Watch' be declared. This resulted in a notification to the DeWitt County Sheriff's Department that occurred at 1133 hours. Lake level continues to slowly rise due to the melting snow and heavy rains received in the area for the past several days. Expected peak is approximately 694 feet 6 inches. Mitigating measures are currently being instituted as a precautionary measure in the screen house, per CPS procedures. The NRC Resident has been notified. High lake level only affects the plant intake structure.
On February 8, 2008 at 1655 hours, Clinton Lake level reached 693 ft-0 inches (mean sea level). Clinton Power Station (CPS) Procedure 4303.02, 'Abnormal Lake Level', and the CPS Main Dam - Emergency Action Plan allow termination of the 'Flood Watch.' This resulted in a notification to the DeWitt County Sheriff's Department that occurred on February 8, 2008 at 2043 hours when the Offnormal procedure was exited. Maximum lake level observed was 694 ft-4 inches (mean sea level). Lake level is currently 692 ft-6 inches (mean sea level) and lowering. Inspections have identified no damage due to the heavy rainfall or associated elevated lake level. The licensee informed the NRC Resident Inspector. Notified R3DO (Tom Kozak). |
| ENS 43802 | 21 November 2007 18:55:00 | On 11/21/2007 at 1408 hours, the Clinton Power Station lost a portion of the Safety Parameter Display System in support of Process Computer replacement. These portions are not expected to be restored to service until Process Computer replacement activities are completed. The Plant Process Computer is expected to be fully operational by February 1, 2008. This event is reportable due to the major loss of emergency assessment capability per 10CFR50.72 (b)(3)(xiii) (when the SPDS has been unavailable for 8 hours or greater). The NRC Resident Inspector has been notified. The alarm for secondary containment sump level has been lost to the SPDS. Compensatory measures are in effect.
The system upgrade has been completed along with full confidence run testing. The SPDS is now considered operational and emergency assessment capability restored. The licensee notified the NRC Resident Inspector and State authorities. R3DO (Ring) was notified. |
| ENS 41130 | 18 October 2004 17:35:00 | In 1991, four unirradiated nuclear instrumentation (NI) detectors (includes detector, cable, and connector) were identified as being damaged. These four NI detectors were dispositioned by cutting the detectors from the cable and placing these detectors in a small Special Nuclear Material (SNM) container which then had tamper seals applied. The remaining cables and connectors were disposed of as trash. On October 6, 2004, the small SNM container was moved from its normal storage location to another plant location in preparation for disposal and was opened to verify the contents on October 7, 2004. The actual contents of the container were three NI connectors (not detectors) and one NI detector. Therefore, three NI detectors containing SNM were identified as unaccounted for. The contents of this SNM container have been periodically inventoried since 1991 verifying that the tamper seals are intact in accordance with the site SNM inventory procedure, and in March 2004, a periodic SNM inventory was performed in accordance with Exelon SNM inventory procedure. As allowed by the Exelon SNM inventory procedure, however, the contents of the small SNM container were not inspected since the tamper seals were intact. A search of the normal NI detector storage location was performed and it was verified that the unaccounted for detectors were not in another container. The other NI detectors in the storage area were accounted for by performing a serial number inventory. Only the three NI detectors missing from the small SNM container are unaccounted for. The total mass of U235 in the unaccounted for SNM was much less than one gram. The actual total mass was determined to be approximately 6 milligrams. It has been determined that the U235 activity of the unaccounted for SNM was greater than 10 times the activity listed in 10CFR20, Appendix C. The actual U235 activity was determined to be 0.012 microcuries, which exceeds 10 times the Part 20 Appendix C activity (i.e., 0.010 microcuries). Therefore, this loss of SNM is reportable in accordance with the requirements of 10 CFR 20.2201(a)(1)(ii). Both the NRC Resident and IEMA Resident were notified. CPS records are being reviewed to determine the actual disposition of the NI cables that were disposed of in 1991. |