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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56328 | 1 February 2023 05:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Scram Due to Turbine Trip | The following information was provided by the licensee via fax and telephone: Generator trip due to power load unbalance which caused a turbine trip and subsequent reactor scram. Experienced a trip on circulating water pump A. NRC Resident Inspector notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Off-site power available and unaffected. Decay heat removal via main steam line and drains to condenser. Plant is stable in mode 3. | Decay Heat Removal Main Steam Line | |
| 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 53698 | 28 October 2018 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram During a Reactor Shutdown | At 0445 (CDT), with reactor power less than 1% rated thermal power on Instrument Range Monitor (IRM) ranges 6 and 7, Clinton Power Station received an automatic Reactor Protection System (RPS) actuation. The Reactor Scram Off Normal procedure was entered and all control rods were verified to be fully inserted. The apparent cause of the scram is cold water injection causing an upscale trip of the IRMs due to Motor Driven Reactor Feedwater Pump (MDRFP) Feedwater Regulating valve 1FW004 valve coming off the full shut seat momentarily. All systems responded appropriately following the scram and the plant is currently stable. Clinton Power Station will be proceeding to Mode 4 to support the planned Maintenance Outage. The NRC Senior Resident Inspector has been notified. | Feedwater Reactor Protection System Control Rod | |
| ENS 53110 | 9 December 2017 19:48:00 | 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)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Manual Reactor Scram Due to Loss of Division 1 Ac Power to Numerous Components | 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). | Secondary containment Reactor Protection System Primary containment High Pressure Core Spray Core Spray Standby Gas Treatment System Control Rod | |
| ENS 52800 | 11 June 2017 03:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Loss of Feedwater Heating | At 2256 CDT on 6/10/17, Clinton operators manually scrammed the reactor from 99 percent power due to a loss of feedwater heating. The scram was uncomplicated and the plant is stable and in mode 3. All rods inserted and decay heat is being removed by the condenser. All offsite power is available. The cause of the loss of feedwater heating is under investigation. The NRC Resident Inspector and the State of Illinois have been notified. | Feedwater | 05000461/LER-2017-007 |
| ENS 52777 | 31 May 2017 01:38:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram | At 2038 (CDT), Clinton Power Station received an automatic RPS (Reactor Protection System)actuation. EOP-1 (Emergency Operating Procedure) was entered on RPV (Reactor Pressure Vessel) Level 3. The cause of the scram is unknown at this time. All systems responded appropriately following the scram and the plant is currently stable. Reactor level is being maintained by normal feedwater and decay heat is being removed to the main condenser via the steam dump bypass valves. The plant is in a normal shutdown electrical lineup. The plant main generator was synchronized to the electrical grid and the plant was conducting control rod scram time testing at the time of the reactor trip. The licensee notified the NRC Resident Inspector. | Feedwater Main Condenser Control Rod | |
| ENS 49958 | 26 March 2014 00:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Loss of Condenser Vacuum | The plant was stable at 85 percent power when off-gas flow started lowering. The operating crew entered the loss of vacuum off-normal procedure and commenced an emergency power reduction to attempt to slow/halt the loss of vacuum while attempting various prescribed actions in the off-normal procedure. When vacuum reached 24 in. Hg, a rapid plant shutdown was performed as prescribed in the off-normal procedure. Power was at 46 percent when a rapid plant shutdown was commenced with the mode switch placed in shutdown at 1942 CDT. All control rods fully inserted on the scram, no emergency core cooling system injected or was required, no safety/relief valve(s) lifted and all systems responded as expected on the scram. The plant will remain in mode 3 with normal makeup from the feedwater system and pressure control on the turbine bypass valves with the main condenser available as a heat sink. All vital and non-vital electrical busses are powered from reserve off-site sources and no emergency diesel generators started on the scram. The cause of the loss of vacuum is under investigation. The licensee notified the NRC Resident Inspector. | Feedwater Emergency Diesel Generator Emergency Core Cooling System Main Condenser Control Rod | |
| ENS 49632 | 13 December 2013 23:57:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip During Feed Pump Swap | During a feed pump shift from the motor driven reactor feed pump to the 'A' turbine driven reactor feed pump during power ascension, reactor pressure vessel (RPV) water level was approaching the Reactor Protective System (RPS) automatic scram set point when a manual reactor scram was inserted. All control rods fully inserted. RPV water level is being maintained by the normal condensate booster / feedwater systems. RPV pressure is being maintained via normal main steam system. There were no actuations of any Emergency Core Cooling System and no Safety Relief Valves actuations. All systems responded as expected. The licensee notified the NRC Resident Inspector. | Feedwater Reactor Pressure Vessel Emergency Core Cooling System Safety Relief Valve Control Rod Main Steam | 05000461/LER-2013-009 |
| ENS 49617 | 9 December 2013 02:27:00 | 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)(v)(C), Loss of Safety Function - Release of Radioactive Material | Manual Scram Due to Loss of Division 1 480 Vac Power Causing Loss of Instrument Air to Containment and Scram Air Header | While operating at rated electrical power, the station experienced a transformer fault which resulted in a loss of Division 1 480 VAC power. This resulted in the operators inserting a Manual Scram due to loss of Instrument Air to Containment and the scram air header. On the scram, all control rods fully inserted and no safety relief valves lifted. Reactor vessel level is being maintained by normal feedwater and decay heat is being removed via steam to the main condenser through the steam bypass valves. The plant is currently in Mode 3 and proceeding to Mode 4 to comply with Technical Specification requirements. The plant is in a normal shutdown electrical lineup with the exception of the loss of Division 1 480 VAC power. Reporting in accordance with 10CFR50.72(b)(3)(v)(C) due to loss of normal ventilation to secondary containment which resulted in a positive secondary containment pressure for approximately 15 minutes. Secondary Containment required pressure was restored at 2043 CST. Reporting in accordance with 10CFR50.72(b)(3)(v)(D) due to loss of Division 1 480 VAC power resulting in loss of a single train of Low Pressure Core Spray. The licensee has notified the NRC Resident Inspector. | Feedwater Secondary containment Core Spray Safety Relief Valve Main Condenser Control Rod | 05000461/LER-2013-008 |
| ENS 48974 | 26 April 2013 13:55:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Rapidly Decreasing Level in the Ehc Oil Reservoir | On 4/26/13 at about 0855 CDT while operating at rated electrical power, operators initiated a manual reactor scram due to rapidly decreasing level in the main Electro Hydraulic Control (EHC) oil reservoir. All systems responded as expected with no complications. The cause of the main EHC decrease in level is under investigation. The plant is stable in mode 3. The NRC Resident Inspector has been notified. The licensee reports that bypass valves remain available via a separate EHC system and decay heat is being routed to the condenser. | 05000461/LER-2013-003 | |
| ENS 48812 | 7 March 2013 13:56:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram on Generator Trip/Turbine Trip | While operating at rated electrical power, a main generator trip and subsequent turbine trip resulted in a reactor scram. The cause of the generator trip is under investigation. All systems operated as expected with no complications. The plant is stable in Mode 3. The licensee notified the NRC Resident Inspector and the State of Illinois. | 05000461/LER-2013-002 | |
| ENS 47489 | 29 November 2011 23:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Unit Experienced an Automatic Reactor Scram During a Planned Shutdown | During a planned shutdown in preparation for refueling outage C1R13, shortly after tripping the main turbine, an automatic reactor scram occurred due to high reactor steam dome pressure. Preliminarily, per the plant process computer, the reactor pressure was observed to reach approximately 1074 psig approximately 26 seconds after main steam bypass valves unexpectedly closed. An investigation will be conducted to determine the cause of the bypass valve closure and reactor scram. No safety relief valves lifted as a result of the pressure increase. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B) as an event or condition that results in actuation of the reactor protection system when the reactor was critical. The licensee is manually controlling the steam bypass valves to remove decay heat via the main condenser. All control rods fully inserted. The Unit is in a normal shutdown electrical lineup. The licensee informed the NRC Resident Inspector. | Reactor Protection System Main Turbine Safety Relief Valve Main Condenser Control Rod Main Steam | 05000461/LER-2011-004 |
| ENS 45433 | 15 October 2009 10:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Reactor Recirculation Pump Trip | At 0537 hours on October 15, 2009, the 'B' Reactor Recirculation pump tripped. The reactor mode switch was placed in shutdown due to rising reactor water level (approx. 49 inches) prior to the Level 8 automatic scram setpoint (52 inches). All controls rods inserted as a result of the manual scram. All systems performed as expected. Reactor water level is being controlled by the motor driven feedwater pump. Main steam isolation valves were manually closed and decay heat was initially controlled through the main steam line drains to the main condenser via the main turbine bypass valves. Reactor Core Isolation Cooling (RCIC) was manually placed into service (tank-to-tank) to assist in RPV pressure control. RCIC is currently being used for decay heat removal. Investigation is underway to determine the cause of the Reactor Recirculation pump trip. The licensee will be contacting the state, and issuing a press release. The NRC Senior Resident Inspector has been notified. | Feedwater Main Steam Isolation Valve Reactor Core Isolation Cooling Reactor Recirculation Pump Decay Heat Removal Main Steam Line Main Condenser | 05000461/LER-2009-005 |
| ENS 43976 | 11 February 2008 04:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Increasing Reactor Vessel Level | On February 10, 2008, at approximately 2207 hours CST, the 'B' Reactor Recirculation pump tripped from fast speed to off. The resultant reactor pressure vessel level swell resulted in reactor pressure vessel level being greater than 48 inches and rising. Per Clinton Power Station operating procedures, the 'A' Reactor Operator placed the mode switch to the 'shutdown' position initiating a manual reactor scram. All control rods inserted (fully). CPS is currently stable in Mode 3. Decay heat is being removed to the main condenser via the turbine bypass valves. Reactor level is being maintained using the motor driven reactor feedpump. No SRVs lifted during the transient. Licensee is in progress of restoring the electrical plant to a normal shutdown lineup. The licensee notified the NRC Resident Inspector. * * * UPDATE AT 2226 ON 2/13/08 FROM M. EVANS TO P. SNYDER * * * After further analysis of the scram, it was determined that an automatic level 8, high reactor water level, scram occurred approximately one second prior to the reactor operator placing the mode switch to the 'shutdown' position. The licensee notified the NRC Resident Inspector. | Reactor Recirculation Pump Reactor Pressure Vessel Main Condenser Control Rod | |
| ENS 42807 | 27 August 2006 22:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Trip Due to Reactor High Water Level | At 1705 CDT on August 27, 2006, a high water level trip occurred resulting in a reactor scram. All control rods fully inserted on the scram signal. Reactor water level is being controlled in the normal operating band and reactor pressure is being controlled in a normal band. The apparent cause of the high level trip was a High Pressure Core Spray (HPCS) system initiation. There is no indication that the HPCS initiation was caused by an actual parameter reaching a trip setpoint. Division four nuclear system protection system (NSPS) is the current focus of troubleshooting activities. The Reactor Core Isolation Cooling (RCIC) system isolated after the scram. Troubleshooting is in progress to determine the cause. Both offsite power sources are operable and emergency diesel generators are operable and available if required. All safety related systems are available if required. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator Reactor Core Isolation Cooling High Pressure Core Spray Control Rod | |
| ENS 42430 | 20 March 2006 10:53:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Main Turbine Trip Caused Reactor Scram | At 0453 CST on March 20, 2006, a main turbine trip occurred resulting in a reactor scram. All control rods fully inserted on the scram signal. All systems performed as expected. Reactor water level is being controlled in the normal operating band and reactor pressure is being controlled at 600 psig. The cause of the turbine trip is under investigation. There is no indication that the turbine trip was caused by an actual parameter reaching a trip setpoint. A turbine Low Vacuum trip alarm is indicated in the Main Control Room and is the current focus of troubleshooting activities. Actual main condenser vacuum is at its normal value. The site is on off-site power, and the EDG's are available if required. All safety related systems are available if required. The licensee notified the NRC Resident Inspector. | Main Condenser Control Rod | |
| ENS 40868 | 13 July 2004 21:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Possible Adverse Weather | On July 13, 2004 at 1610 CDT, the Clinton Power Station 345 KV unit output breakers GCB 4506 and 4510 opened resulting in a turbine control valve fast closure. This caused a reactor protection system actuation and reactor scram. Clinton was in Mode 1, operating at 95% rated thermal power when the event occurred. A tornado warning was in effect at the time of the event and weather conditions were degrading in the area surrounding the plant. The exact cause of the trip of the unit breakers is not yet known. All systems responded as expected following the scram with the exception of Reactor Recirculation pump A which tripped off instead of downshifting to slow speed on a reactor water level 3 signal. The unit is currently in Mode 3 with reactor pressure vessel level and pressure in their normal bands. Recirculation Pump "A" remains secured until the licensee completes their investigation. All control rods fully inserted following the scram. The Main Condenser is in service removing decay heat. All ECCS equipment including the EDGs are available, if needed. The licensee informed the NRC Resident Inspector. | Reactor Protection System Reactor Recirculation Pump Reactor Pressure Vessel Main Condenser Control Rod | |
| ENS 40604 | 23 March 2004 01:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram from 93% Power Due to Generator Trip | The plant had a reactor SCRAM, cause is under investigation with initial indications of a Main Generator over voltage alarm and Generator lockout #2 were indicated on the Generator first hit panel. All plant systems responded as expected to the generator trip and Reactor SCRAM. All rods fully inserted and no ECCS actuations or relief valves lifted. The plant is stable in Mode 3 with the recovery plan being implemented. The NRC Resident Inspector was notified. | ||
| 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 |