Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56997 | 28 February 2024 18:50:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Automatic Start of Emergency Diesel Generators | The following information was provided by the licensee via phone and email: At 1350 EST on 2/28/2024, with Calvert Cliffs Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 65 percent power, an actuation of the '1A' and '2A' emergency diesel generators' auto-start occurred due to an undervoltage condition on the number 11 and number 21 4kV buses which are fed from the number 11 13kV bus. The '1A' and '2A' emergency diesel generators automatically started as designed when the 4kV buses' undervoltage signals were received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the '1A' and '2A' emergency diesel generators. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The undervoltage condition was caused by the feeder breaker to the number 11 13 kV bus opening during electrical maintenance. | Emergency Diesel Generator | |
| ENS 56991 | 24 February 2024 20:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Steam Generator 22 Feed Pump Trip | The following information was provided by the licensee via email: At 1546 EST, with unit 2 at 100 percent power, the reactor was manually tripped due to the '22' steam generator feed pump tripping. The trip was uncomplicated with all systems responding normally post-trip. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Operations responded using emergency operation procedure EOP-0, Post Trip Immediate Actions and EOP-1, Uncomplicated Reactor Trip and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. ESFAS (engineered safety features actuation systems) actuation (auxiliary feedwater manual actuation) is reportable under 10 CFR 50.72(b)(3)(iv)(A) 8-hour report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 56856 | 16 November 2023 07:27:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: At 0227 EST on 11/16/23, Calvert Cliffs Unit 2 experienced an automatic trip from the reactor protection system (RPS) based on reactor trip bus undervoltage (UV). At that time, a loss of U-4000-22 (13 kV to 4 kV transformer) caused a loss of 22, 23, and 24 4 kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV. The loss of 22 and 23 4 kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4 kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4 hour report ESFAS (engineering safety features actuation system) actuation (2B DG start on UV) is reportable under 10 CFR 50.72(b)(3)(iv)(A) - 8 hour report AFW operation is reportable under 10 CFR 50.73(a)(2)(iv)(A) - 60 day report The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. There was no impact on Unit 1 operations. Unit 2 is stable in mode 3.
ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8 hour report Notified R1DO (Defrancisco). | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 56841 | 8 November 2023 11:45:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Actuation of Auxiliary Feedwater System (AFW) | The following information was provided by the licensee via phone and email: At 0645 EST, on November 8, 2023, with Unit 2 in Mode 3 at zero percent power, a manual actuation of the auxiliary feedwater system (AFW) occurred during a planned plant cooldown. The reason for the AFW manual-start was a trip of the 22 steam generator feed pump due to a high casing level. The 23 AFW motor driven pump was manually started in accordance with implementation of AOP-3G, Malfunction of Main Feedwater System to restore steam generator levels. There was no impact to Unit 1. 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 AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No other systems were affected. No other compensatory or mitigation strategies implemented. Plant cooldown was the only significant evolution in progress. No impact to other technical specifications or limiting conditions for operation. All systems functioned as required. The electric plant is being supplied by offsite power with all diesel generators available. No significant increase in plant risk. There was nothing unusual or not understood. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 56839 | 7 November 2023 21:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to NON-SAFETY Related Bus Under Voltage | The following information was provided by the licensee via email: At 1617 on 11/7/2023, Calvert Cliffs Unit 2 experienced an automatic trip from a Reactor Protection System (RPS) based on reactor trip bus under voltage (UV). At that time a loss of U-4000-22 caused a loss of 22, 23, and 24 4kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV condition. The loss of 22 and 23 4kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4-hour report. ESFAS actuation (2B DG start on UV) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. Site Senior NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 was unaffected. Estimation of duration of shutdown is 24 hours. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Main Condenser | |
| ENS 56108 | 14 September 2022 17:48:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Safety Injection Actuation Signal While Shutdown | The following information was provided by the licensee via email: With Unit 2 in Mode 5 and cooling down for a scheduled maintenance outage, Operations was bypassing safety injection and safety injection bypass sensor modules per procedure. Due to a human performance issue, a safety injection actuation signal was generated. All equipment operated as expected. Operations has reset the safety injection actuation signal and restored the equipment to its required condition. Unit 1 was at 100 percent power and unaffected by this issue. The NRC Resident Inspector will be notified. | ||
| ENS 55684 | 3 January 2022 17:23:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Due to Loss of Electrical Load | The following information was provided by the licensee via email: At 1223 (EST) on 01/03/2022, Calvert Cliffs Unit 2 automatically tripped from 100 percent power due to loss of electrical load. The cause is under investigation. The site Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods inserted and decay heat is being removed via the condenser. The plant is in a normal shutdown electrical lineup. There was no impact on Unit 1. | ||
| ENS 55597 | 21 November 2021 15:46:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip and Automatic Auxiliary Feedwater Actuation | At 1046 EST on November 21, 2021, with Calvert Cliffs Nuclear Power Plant Unit 2 in Mode 1 at 100 percent power, the reactor was manually tripped due to lowering levels in both steam generators following a loss of the 21 and 22 steam generator feed pumps. An Auxiliary Feedwater System actuation occurred to restore steam generator water levels. The trip was not complicated, with all systems responding normally. Decay heat is being removed by the Auxiliary Feedwater System. Calvert Cliffs Nuclear Power Plant Unit 1 is unaffected and remains in Mode 1 at 100 percent power. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification. RPS actuation, per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the automatic actuation of the Auxiliary Feedwater System is being reported as an eight-hour, non-emergency notification, Specific System Actuation, per 10 CFR 50.72(b)(3)(vi)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 55147 | 22 March 2021 02:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Lowering Steam Generator Level | At 2216 EDT on 3/21/2021, Calvert Cliffs Unit 2 was manually tripped from 37 percent power due to lowering level in the 21 Steam Generator. All systems responded per design. Main Feedwater was secured and Auxiliary Feedwater was manually initiated. The Site Senior Resident has been notified. The cause of the lowering level in the 21 Steam Generator is under investigation. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 55020 | 3 December 2020 14:23:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of an Emergency Diesel Generator | At 0923 EST on December 3, 2020, with Unit 1 in Mode 1 at 100 percent power, an actuation of the Emergency AC Electrical Power System (Emergency Diesel Generator 1A) occurred during normal plant operations. The reason for Emergency Diesel Generator 1A auto start was due to Class 1E 4KV Bus 11 feeder breaker opening. The Emergency Diesel Generator 1A automatically started as designed when the loss of voltage signal on 4KV Bus 11 was received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the Emergency AC Electrical Power System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The cause of the 4KV Bus 11 Feeder Breaker opening is unknown at the present time and is under investigation. | Emergency Diesel Generator | |
| ENS 52406 | 4 December 2016 03:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip Caused by Main Turbine Trip | On 12/3/16 at 2224 EST, Calvert Cliffs Unit-2 experienced an automatic reactor trip from full power due to a leak in the Unit-2 Main Turbine Electro-Hydraulic Control (EHC) system. The EHC leak caused the Unit-2 Main Turbine governor valves to close, resulting in a turbine trip and automatic reactor trip. The site Outage Control Center is manned, and investigation into the cause of the leak is underway. Unit-2 remains stable in Mode 3 with normal heat removal. Unit-1 remains at full power and was not affected by the trip. The plant is in a normal shutdown electrical lineup. All Control rods fully inserted and no primary or secondary safety relief valves lifted during the trip. The licensee has notified the NRC Resident Inspector. The licensee will be notifying Calvert County. | Main Turbine Safety Relief Valve Control Rod | |
| ENS 51967 | 31 May 2016 20:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Calvert Cliffs Loss of Load Turbine/Reactor Trip from 100 Percent Power | On 05/31/2016 at 1626 (EDT), a Reactor Trip occurred on Loss of Load RPS (Reactor Protection System) actuation. The Turbine Trip was apparently caused by a failed ESFAS (Engineered Safety Feature Actuation System) logic module. The failed NSR (Non Safety Related) logic module was associated with 11 Steam Generator High Level Turbine Trip. All systems responded as designed. Normal decay heat removal is to the condenser. All offsite power sources remained in service. The NRC Resident Inspector was notified. | Steam Generator Decay Heat Removal | |
| ENS 51683 | 25 January 2016 08:15:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Elevated Condenser Sodium Levels | At 0315 EST on 1/25/16, Calvert Cliffs Unit 1 was manually tripped from 10 percent power due to elevated condenser sodium levels. All systems responded per design. Main Feed was secured and auxiliary feed water was initiated. The elevated sodium levels are believed to be due to a condenser tube leak. The reactor is currently shutdown and stable in Mode 3 and will remain in Mode 3 until repairs are effected. Unit 2 was not affected and remains at full power. The Licensee has notified the NRC Resident Inspector. | Auxiliary Feedwater | |
| ENS 51577 | 1 December 2015 23:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip After Steam Generator Feed Pump Trip | On 12/01/2015 at 1820 EST, the Main Control Room received a 22 Steam Generator Feed Pump trip. The 22 Steam Generator Feed Pump was not able to be reset and the Main Control Room manually tripped the Unit 2 Reactor. The licensee entered Emergency Operating Procedure (EOP)-0, 'Post Trip Immediate Actions' and all safety functions were met. At 1833, Unit 2 transitioned into EOP-1, 'Uncomplicated Reactor Trip.' At 1841, Unit 2 transitioned into Operating Procedure #4 , 'Plant Shutdown from Power to Hot Stand-by.' The plant is stable in Mode 3. All control rods inserted fully on the reactor trip. No primary or secondary safety relief valves lifted. The steam generators are being fed by the 21 steam generator feed pump and decay heat is being dumped to the condenser via the steam dumps. The electric plant is in a normal shutdown electrical lineup and there was no impact on Unit 1. Unit 1 continues to operate at 100 percent power. The cause of the 22 steam generator feed pump trip is still under investigation. The licensee notified the NRC Resident Inspector. | Steam Generator Safety Relief Valve Control Rod | |
| ENS 50961 | 7 April 2015 16:45:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Dual Unit Automatic Reactor Trips Due to a Voltage Transient Resulting in Generator Trips | A loss of Main Generator Load which caused a Reactor Trip on Units 1 & 2. A switchyard voltage transient from a highline occurred, which caused an undervoltage condition on both units' safety related 4KV buses. Unit 1 is on normal heat removal to the condenser. Unit 2 is on auxiliary feedwater and normal condenser bypass valves for temperature control. An Auxiliary Feedwater Actuation System (AFAS) actuation occurred on Unit 2. The (Unit 2) 2B emergency diesel generator did not start and load on its respective 24-4 KV bus. The 24-4KV Bus was repowered from the alternate feeder breaker. Cause of the emergency diesel failure to start is under investigation. All safety functions are met for both units. All control rods fully inserted. The site is in a normal shutdown electrical configuration powered from offsite. The site plans to stay in Mode 3 pending restart. The licensee notified the NRC Resident Inspector, State and local authorities. A press release is planned.
During post trip review, it was determined that the 21 saltwater pump had to be manually started. With the failure of 2B emergency diesel generator, there were no saltwater pumps running for approximately 12 minutes. Additional troubleshooting determined the 2A emergency diesel generator sequencer did not automatically start 21 saltwater pump. The 2B emergency diesel generator was returned to service on 4/8/2015 at 1730 (EDT). The loss of saltwater (pump) and emergency diesel generator is reportable as an event that could have prevented fulfillment of a safety function and is also an unanalyzed condition. The licensee has notified the NRC Resident Inspector. Notified R1DO (Ferdas), IRD MOC (Grant), NRR EO (Morris). | Emergency Diesel Generator Auxiliary Feedwater Control Rod | 05000317/LER-2015-002 |
| ENS 50078 | 1 May 2014 14:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | At 1016 (EDT) CCNPP (Calvert Cliffs Nuclear Power Plant) Unit 1 automatically tripped due to an RPS actuation. Cause is under investigation. All safety functions are met with normal heat removal. Electric plant is in a normal lineup. No ESFAS (Engineered Safety Feature Actuation System) actuations have occurred. Steam Generator atmospheric dump valves momentarily opened and then closed. There is no known steam generator tube leakage. All control rods fully inserted on the trip. There was no impact on Unit 2 from this event. The licensee notified the NRC Resident Inspector and provided a courtesy notification to the Calvert County Control Center. | Steam Generator Control Rod | |
| ENS 49754 | 22 January 2014 02:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Dual Unit Trip on the Loss of the "21" 13Kv Bus | Dual Unit Trip due to loss of '21' 13 KV bus . All safety functions are met for both units. Unit 1 remained with normal heat removal. Unit 2 lost power to its normal heat sink and is stable on Auxiliary Feed water and Atmospheric Dump Valves for temperature control. Both trips were automatic trips. Due to loss of power a Under Voltage actuation occurred on both units ('14' and '24' 4Kv bus). Due to loss of main feed on Unit 2 a Auxiliary Feed water Actuation System (AFW) actuation occurred on Unit 2. Cause is under investigation. All control rods fully inserted on the loss of power to the Control Rod Drive Mechanisms (CRDMs). Both Units Reactor Coolant Pumps (RCPs) remained running during the transient. The normal Unit 2 heat sink was unavailable due to the loss of the operating circulating water pumps resulting in a loss of condenser vacuum. The Unit 2 AFW actuation included one of two steam-driven pumps and the motor-driven pump. Both Units Emergency Diesel Generators started and loaded and have since been secured. Both Units are stable and will remain in mode 3 (Hot Standby) pending the results of the investigation. The licensee will inform the NRC Resident Inspector. | Emergency Diesel Generator Auxiliary Feedwater Control Rod | 05000317/LER-2014-001 05000318/LER-2014-001 |
| ENS 49054 | 21 May 2013 09:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip After a Feed Pump Trip | Reactor trip. All safety functions met with normal heat removal. 22 SGFP (Steam Generator Feedpump) exhibited high vibrations and signs of coupling damage. Further investigation will be performed. All control rods fully inserted on the trip. Steam Generator level is being maintained with the remaining feedpump. Decay heat is being dumped to the main condenser. Electrical power is in the normal shutdown lineup. No relief or safety valves lifted during the trip. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector. | Steam Generator Main Condenser Control Rod | 05000318/LER-2013-004 |
| ENS 49012 | 9 May 2013 01:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The reactor automatically tripped at 2147 EDT. All control rods fully inserted on the trip and all systems responded as expected. Decay heat removal is to the main condenser. The plant is in its normal shutdown electrical lineup. The licensee is investigating the cause of the reactor trip. The licensee notified the NRC Resident Inspector. | Decay Heat Removal Main Condenser Control Rod | 05000318/LER-2013-003 |
| ENS 47208 | 28 August 2011 03:02:00 | 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 | Unusual Event - Transformer Explosion Near Turbine Building | At 2248 on 8/27/2011, the Unit 1 Reactor experienced an automatic trip due to loss of load. This trip occurred due to a phase to phase short on the main generator output step-up transformer that resulted from a large section of turbine building siding breaking loose in high winds from Hurricane Irene and impacting the transformer. This impact resulted in an explosion (briefly until the trip removed power from the impact area) which met emergency action level declaration criteria A.U.6.2.2, 'Unanticipated explosion within Protected Area resulting in visible damage to permanent structures or equipment.' The Unusual Event was declared at 2302, 8/27/2011. Follow-up investigation determined no fire resulted from the explosion. Following the trip, Emergency Procedure, EOP-0, 'Post Trip Immediate Actions' was implemented. All safety functions were met during EOP-0 indicating an uncomplicated reactor trip response, allowing transition to EOP-1, 'Reactor Trip,' at 2300, 8/27/2011. During implementation of EOP-1, it was noted that #14 Containment Air Cooler had stopped running, as had #21 and #24 Containment Air Coolers on Unit 2. This was investigated and it was determined they had stopped running due to an instantaneous voltage drop that had occurred on the site distribution system during the phase to phase short event. This short duration voltage drop caused the Containment Air Coolers' controller to drop out and secure them. They were restarted without issue. At 2400, 8/27/2011, numerous alarms on the 1A DG started to be received. These were investigated and it was found that water was intruding down the DG exhaust piping resulting in a DC ground. Based on these indications the 1A DG was declared inoperable and appropriate technical specifications implemented. Besides the above issues plant response was as expected and EOP-1 was exited at 0130, 8/28/2011. (Procedure) OP-4, 'Shutdown from Power Operation to Hot Standby,' was implemented at that time. All control rods fully inserted on the reactor trip. The plant is in a normal post-trip electrical lineup. The licensee notified the NRC Resident Inspector. * * * UPDATE FROM GREGG BUCKMASTER TO PETE SNYDER AT 0811 EDT ON 8/28/11 * * * At 0755 EDT the licensee exited the Unusual Event condition based on the fact that they were able to inspect the area in the daylight and were satisfied that they knew the extent and nature of the damage. The licensee will notify the NRC Resident Inspector. Notified R1 IRC (Dentel), DHS (Hill), FEMA (Via). | Control Rod | 05000317/LER-2011-001 |
| ENS 45920 | 12 May 2010 17:51:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Actuation - Load Reject | At 1351 hrs. EDT on May 12, 2010, a Calvert Cliffs Unit 1 Reactor trip was generated by a complete load rejection when a main generator output breaker opened while its redundant breaker was opened for planned maintenance. The generator output breaker trip caused Reactor Coolant System (RCS) pressure to rise to the Reactor Protection System (RPS) High Pressure setpoint, which opened both Power Operated Relief Valves (PORVs). Both PORVs reclosed when RCS pressure was reduced to normal. All Control Rods fully inserted on the trip. Normal heat removal methods are currently in progress. Unit 1 is in a normal shutdown electrical lineup and there was no impact on Unit 2. The NRC Resident Inspector has been notified by the licensee. | Reactor Coolant System Reactor Protection System Control Rod | 05000317/LER-2010-003 |
| ENS 45709 | 18 February 2010 13:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Dual Unit Automatic Reactor Trips Due to Partial Loss of Offsite Power | Both U-1 and U-2 automatically tripped due to valid actuation of the Reactor Protection Systems (RPS). U-1 due to loss of 12B Reactor Coolant Pump (RCP) which resulted in a RCS Low Flow RPS Trip. Cause for loss of 12B RCP is currently not known but is suspected to be related to the electrical transient that occurred on U-2. U-2 tripped due to Loss of Load RPS Trip when the main turbine tripped due to an electrical malfunction and partial loss of offsite power. The electrical malfunction resulted in loss of power to all 4 Kv buses on U-2 with the exception of 21 4Kv bus (ZA train power) (Lost 22-26 4Kv buses) and the loss of 14 Kv bus (ZB train power) on U-1. All the buses lost are powered from the same in house service transformer, P-13000-2, which was lost due to the electrical transient. Cause of the electrical transient is being pursued but is unknown at this time. Loss of the 14 4Kv bus on U-1 and the 24 4Kv bus on U-2 resulted in an Engineered Safeguards Actuation System (ESFAS) valid actuations on both units due to Under Voltage (UV) conditions on those buses. The 1B (Diesel Generator) DG started automatically due to the 14 bus UV and is carrying that bus. The 2B DG received an automatic start signal due to the 24 bus UV but failed to start as expected. No other system actuations occurred. Both units are currently stable in Mode 3 at normal operating temperature (532 degrees) and pressure (2250 PSIA) with no other significant equipment malfunctions. Due to the loss of P-13000-2 (transformer), U-2 sustained a loss of normal heat removal due to loss of the main condenser cooling and the loss of secondary pumps. U-2 is removing heat with auxiliary feedwater pumps and (Steam Generator) SG atmospheric dump valves without issues. Current plans are to cool U-2 down to 445 degrees to maintain RCP seals cool until RCP's can be restarted. A decision on U-1 cool down has not yet been made. There is no primary to secondary leakage. U-2 is currently in a 12 hour LCO due to the partial loss of offsite power and 2B DG failing to start. The licensee has notified the NRC Resident Inspector. | Reactor Protection System Auxiliary Feedwater Main Turbine Main Condenser | 05000317/LER-2010-001 05000318/LER-2010-001 |
| ENS 43046 | 12 December 2006 14:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip During Turbine Control System Maintenance | At 0907 Calvert Cliffs Unit 1 experienced a manual reactor trip during turbine control maintenance. Maintenance was being performed on turbine control to replace circuit cards. This system is a two out of three logic system. Only one logic branch was being de-powered at a time. When the first logic branch was de-powered, the control room supervisor noticed a transient on the reactor coolant system and ordered a manual trip. Decay heat is being removed via normal methods through the Turbine Bypass Valves to the condenser. At the time of the trip, 11 Auxiliary Feedwater (AFW) Pump was being tested but it was not feeding the steam generators. This was an uncomplicated trip. The plant responded normally to the event. All control rods fully inserted. The plant is stable and operators are maintaining the unit in Hot Standby. The licensee notified the NRC Resident Inspector. The trip had no effect on Unit 2, The licensee was in no significant safety system LCO action statement at the time (other than for the AFW 11 surveillance that was in progress at the time of the event.) The steam generators remained on normal feed throughout the event. No Porv's or safeties lifted during the transient. All systems functioned as required. The licensee is investigating how de-powering only one control system logic circuits resulted in the transient. | Steam Generator Reactor Coolant System Auxiliary Feedwater Control Rod | 05000317/LER-2006-004 |
| ENS 42995 | 16 November 2006 05:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip Due to Turbine Trip and Discovery of After-The-Fact Emergency Condition of an Unsual Event | At 0018 Calvert Cliffs Unit 2 experienced an automatic reactor trip due to a turbine trip. "At the time a clearance order was being performed for upcoming maintenance on P-13000-2 (transformer). As a result of the turbine trip, RCS pressure rose to approximately 2420 psia causing the PORV's to open. Unexpectedly, a Pressurizer Safety Valve, RV-200 also lifted and reclosed when RCS pressure was lowered to approximately 1500 psia. As a result of the pressure decrease, a Safety Injection Actuation Signal (SIAS) occurred. Once the Pressurizer Safety Valve reclosed, RCS pressure began to rise to return to normal values. Decay heat is being removed via normal methods through the Turbine Bypass Valves to the Condenser. Normal Feedwater is being used. No Auxiliary Feedwater actuation occurred. Two Reactor Coolant Pumps were secured as a result of the SIAS. The plant responded normally to the event. The plant is currently stable and operators are conducting a plant cooldown to mode 5. All control rods fully inserted. The licensee notified the NRC Resident Inspector.
The automatic trip was due to high RCS pressure from the closure of the Turbine Intercept Valves. Both PORV's opened as designed. One remained open approximately 1.5 minutes causing RCS pressure to reduce to 1500 psia. The PORV should have closed at 2400 psia. Relief Valve 200 (RV-200) did not open as previously reported and was not the cause of the RCS pressure lowering. Acoustic monitoring indication were due to the close proximity of the PORV. Since the SIAS signal did not cause a reportable ECCS actuation, the reported 50.72(b)(2)(iv)(A) ECCS Actuation is retracted. The licensee notified the NRC Resident Inspector. Notified R1DO (P. Henderson).
Upon further review, the licensee has determined that this event met the criteria for Unusual Event Emergency Action Level (EAL A.U.2.2.1) for identified RCS leakage greater than 25 gpm. The licensee met this criteria for the duration that the PORV valve remained open (less than 2 minutes). The licensee did not recognize that it had met the criteria at the time of the event and is reporting this as an after-the-fact emergency condition of unusual event per the guidance in NUREG-1022. The licensee notified the NRC Resident Inspector. Notified R1DO (P. Henderson), NRR EO (Ross-Lee), and IRD Manager (Leach). | Feedwater Auxiliary Feedwater Control Rod | 05000318/LER-2006-001 |
| ENS 41452 | 1 March 2005 05:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram from 15% Power Due to High Turbine Vibration | The following details were provided by the licensee via fax as part of their telephonic notification: Manual reactor scram due to high turbine generator vibration during a planned Unit 1 power reduction to Mode 2 to perform maintenance to replace the vent line piping on 11 MSR drain tank. The manual reactor scram was initiated per AOP-7E, section V, 'High Turbine Vibration' trip criteria when bearing #5 exceeded 12 mils vibration. This event meets NUREG-1022 Rev. 2, Section 3.2.6 'System Actuation' Part 50.72(b)(2)(iv)(B) 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' All rods fully inserted and no ECCS or relief valve actuations occurred. Heat sink is the condenser and turbine bypass valves. The NRC resident inspector was notified. | Reactor Protection System | 05000317/LER-2005-002 |
| ENS 40601 | 20 March 2004 18:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Trip at Calvert Cliffs Unit 1 Due to a Loss of Feedwater | At 1340 on 3/20/2004, Calvert Cliffs Unit 1 Reactor automatically shutdown due to low Steam Generator Water Level. The low water level was caused by a loss of at least one Steam Generator Feed Pump. The loss was initially caused by a short or ground from Chart Recorder maintenance in Panel 1C29. 1C29 is a control panel in the control room. The chart recorder was a 500KV Bus Voltage Monitor. The post trip primary indications responded normally. The auto steam dump operation responded normally until the quick open signal was cleared at which time the Turbine Bypass Valves failed shut. It is unclear at this time why they failed shut. Steam dump continues through the use of the Atmospheric dump valves and feedwater is supplied via the Auxiliary Feedwater System with the use of 11 (Steam Driven Pump) & 13 AFW Pump (electric driven pump). Lowest Steam Generator Level was -210" in 11 S/G and -115" in 12 S/G level. Current Conditions are RCS pressure is 2250 PSIA and temperature is 532�F. All control rods properly inserted into the reactor core. The NRC Resident Inspector was notified. | Feedwater | 05000317/LER-2004-001 |
| ENS 40472 | 23 January 2004 20:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip with Safety Injection Actuation Signals | At 1529 ET the reactor was manually tripped due to low steam generator level as a result of the loss of the 22 SGFP (Steam Generator Feed Pump). When the reactor was tripped, the turbine bypass and atmospheric dump valves went full open and did not shut. This lead to a SIAS (Safety Injection Actuation Signal) and SGIS (Steam Generator Isolation Signal). At 1718 ET, a second SIAS actuation occurred while reestablishing pressurizer level. The cause of the SGFP trip is unknown. The cause of the over steaming on turbine bypass valves and atmospheric dump valves is unknown. All control rods fully inserted. Auxiliary Feedwater initiated normally. The licensee stated that there was no actual ECCS injection to the RCS. The plant electrical system responded normally and all emergency diesel generators remain operable. All ECCS systems remain operable. There are no primary to secondary leaks. Decay heat is currently being removed via the steam-driven auxiliary feedwater pump and the atmospheric steam dumps. As of 1830 ET, primary pressure is approximately 2103 psi and pressurizer level is at 242 inches. Plant conditions are being stabilized at normal hot standby values . The licensee notified the NRC Resident Inspector who responded to the Control Room.
The licensee reported that their post trip review determined that the reactor received an automatic trip about 2 seconds before the manual trip. The automatic trip was due to low steam generator water level. The licensee notified the NRC Resident Inspector. Notified NRR (Reis) and R1DO (Meyer)
During post event review of the 1/23/04 Rx Trip event (40472), it was discovered that Unit 2 met the entry criteria for EOP-4, Excess Steam Demand event. This condition required declaration of an Unusual Event. No declaration was made at the time because of the transitory nature of the condition. Transitory conditions such as this, however, still require reporting per NUREG-1022, Rev. 2. Immediate notification to the NRC via the Emergency Notification System is required per 10 CFR 50.72(a)(1)(i). The NRC Resident Inspector will be notified. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Control Rod | 05000318/LER-2004-001 |