Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56769 | 1 October 2023 03:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip | The following information was provided by the licensee via email: At 2014 (PDT) on 09/30/2023, with (Diablo Canyon) Unit 1 in Mode 1 at 11 percent reactor power in preparation for a pre-planned manual reactor trip into a scheduled refueling outage, the reactor was manually tripped due to a failed secondary system dump valve. Auxiliary feedwater was manually started in accordance with plant procedures. This event is being reported in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). There was no plant or public safety impact. 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: Diablo Canyon Unit 2 was unaffected. | Auxiliary Feedwater | |
| ENS 55526 | 16 October 2021 00:49:00 | 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 | On October 15, 2021, at 1749 PDT with Diablo Canyon Power Plant Unit 2 operating at approximately 90 percent power, the reactor was manually tripped in accordance with plant procedures due to increasing water level in feedwater heater 2-5B. The reactor trip was uncomplicated and the Auxiliary Feedwater system started as expected. The plant is stable in Mode 3. This notification is being made in accordance the requirements of 10 CFR 50.72(b)(2)(iv)(B) as a manual actuation of the reactor protection system and 10 CFR 50.72(b)(3)(iv)(A) as an automatic actuation of a specified safety system. The cause of the increased feedwater level is under investigation. There was no impact to the health and safety of the public or plant personnel. The Resident Inspector has been notified. | Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 53764 | 1 December 2018 08:00:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip of Unit 2 Due to a Load Rejection | At 1006 (PST), on December 1, 2018, with Unit 2 at 100 percent power, the reactor automatically tripped due to a load rejection from the 500 kV offsite electrical system. Operations responded and stabilized the plant. Decay heat is being removed by the Main Steam system to the main condenser using the steam dump valves. The cause of the load rejection is currently under investigation. 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). Additionally, due to the actuation of the Auxiliary Feedwater System, as expected, this event is being reported per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector was notified. A press release is planned for this event. All control rods fully inserted and the trip was uncomplicated. There was no effect on Unit 1. | Reactor Protection System Auxiliary Feedwater Main Condenser Control Rod Main Steam | |
| ENS 51965 | 30 May 2016 16:30:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Opening of Reactor Trip Breakers Due to Rod Position Error Greater than 12 Steps | On 5/30/2016 at 0930 (PDT), Unit 2 was in its 19th refueling outage in Mode 4 (hot shutdown, reactor subcritical). Routine post-maintenance testing of digital rod position indication (DRPI) was in progress in accordance with Surveillance Test Procedure (STP) R-1C. Rod Shutdown Bank A was being withdrawn. With Shutdown Bank A Group 1 demand indicating 14 steps, and Group 2 demand indicating 13 steps, Bank A DRPI indicated 12 steps but control rod B4 DRPI indication remained at Step 0. With the bank demand position exceeding rod B4's DRPI indication by greater than 12 steps, Operators manually opened the Reactor Trip Breakers, placing all the rods in a known position due to an inoperable DRPI system. All systems actuated as required and rods fully inserted. Manual initiation of a reactor trip where the actuation is not part of a pre-planned evolution is reportable under 10 CFR 50.72 (b)(3)(iv)(A). While the reactor trip was initiated in accordance with the STP's instructions, the manual actuation was not a preplanned outcome of the STP. Subsequently, it was determined that rod B4 had remained on the bottom at Step 0 as indicated by DRPI. A moveable gripper fuse was blown, preventing that rod from being withdrawn. The blown fuse was replaced and testing continued in accordance with the procedure. There was no impact to public, employee, or plant safety. The NRC Resident Inspector was notified. | Control Rod | |
| ENS 50586 | 1 November 2014 00:40:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Emergency Diesel Generator Start Due to Loss of Offsite 230Kv Startup Power Supply | At 1740 PDT, Diablo Canyon Power Plant experienced a loss of the offsite 230 kV startup power source. This was due to a 230 kV switchyard bus differential relay actuation during a rainstorm, resulting in valid anticipatory starts of each unit's three emergency diesel generators. All diesels successfully started but were not loaded. All systems operated as designed with no problems observed. The 230 kV startup power source is the offsite power system designed to be immediately available following an accident. However, the safety related onsite emergency diesel system would have provided power to mitigate the consequences of an accident while the 230 kV (supply) was unavailable. Unit 1 remains on line at 100 percent power. Unit 2 is currently in Mode 5 in the process of completing a planned refueling outage. Unit 2 power was supplied from the 500kV offsite power source at the time of this event and was therefore unaffected. The offsite 230 kV startup power source has been restored to service. The NRC Resident Inspector has been notified. The normal offsite power supply remained available and unaffected. | Emergency Diesel Generator | |
| ENS 49786 | 2 February 2014 19:29:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Automatic Reactor Trip | On February 2, 2014, at 11:29 PST, Diablo Canyon Power Plant Unit 2 500kV line differential relay actuated. This action tripped the turbine and opened the generator output breakers to isolate the generator. With the turbine tripped and Unit 2 operating above the P-9 50% power permissive, a reactor trip was initiated from the reactor protection system. All systems operated as designed with no problems observed. All three Unit 2 Auxiliary Feedwater pumps started, the Containment Fan Cooling units started and ran in slow speed, and the standby Auxiliary Saltwater train started, all as expected. Unit 2 is stable at normal operating temperature and pressure. All power transferred to the plant startup source without incident. Condenser vacuum was maintained. The preliminary cause of the differential relay actuation was a flashover of Phase B 500 kV to ground across the Phase B lightning arrestor during a rainstorm. Decay heat is being removed by steam dumps to the condenser. No relief valves lifted during the transient. The steam generators are being supplied by the auxiliary feedwater pumps. There were no injuries to personnel. Unit 1 was not affected. NRC Senior Resident Inspector and Region Branch Chief have been informed of this event. A press release is planned for local media. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 49287 | 16 August 2013 01:24:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Startup Power Results in Valid Starts of All Three Emergency Diesel Generators | At 1824 PDT on August 15, 2013, Unit 1 experienced a loss of startup power due to a failure of Startup Transformer 1-1 load tap changer. This loss caused a valid auto-start signal to all three emergency diesel generators and they all started successfully. At 1921, all EDGs were shutdown and returned to standby per plant procedures. As a result of the loss of startup power, power was also lost to site service buildings. ERDS was lost but compensatory measures are in place to transmit required data via the ENS line if required. The plant is in a 72-hr. shutdown LCO action statement under T.S. 3.8.1 for the loss of one of three qualified circuits. The two other qualified circuits (vital power via auxiliary transformers and the EDGs) remain operable. The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator | |
| ENS 49184 | 10 July 2013 16:50:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Automatic Reactor Trip | On July 10, 2013, at 0950 PDT, the Diablo Canyon Power Plant Unit 2 500kV line differential relay actuated. This action tripped the turbine and opened the generator output breakers to isolate the generator. With the turbine tripped and Unit 2 operating above the P-9 50% power permissive, a reactor trip was initiated from the reactor protection system. All plant equipment responded as designed. All three Unit 2 Auxiliary Feedwater pumps started, the Containment Fan Cooling units started and ran in slow speed, and the standby Auxiliary Saltwater train started, all as expected. Unit 2 is stable at normal operating temperature and pressure. All power transferred to the plant startup source without incident. Condenser vacuum was maintained. The preliminary cause of the differential relay actuation was a flashover of Phase A 500 kV to ground across the Phase A lightning arrestor during maintenance activities to wash the 500 kV insulators. NRC Resident Inspector has been informed of this event. A press release is planned for local media. Decay heat is being removed by steam dumps to the condenser. No relief valves lifted during the transient. The steam generators are being supplied by the auxiliary feedwater pumps. There were no injuries to personnel. Unit 1 was not affected. | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 49143 | 24 June 2013 04:20:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of 230 Kv Offsite Power Results in the Automatic Start of Emergency Diesel Generators | At 2120 PDT, Diablo Canyon Power Plant experienced a loss of the offsite 230 kV startup power source due to an offsite transmission system relay actuation, resulting in valid anticipatory starts of Units 1 and 2 three emergency diesel generators on each unit. All diesels successfully started but were not loaded. All systems operated as designed with no problems observed. The 230 kV startup power source is the only offsite power system designed to be immediately available following an accident. However, the safety related onsite emergency diesel system would have provided power to mitigate the consequences of an accident while the 230 kV was unavailable. Restoration of the 230 kV offsite power system is in progress." The licensee will notify the NRC Resident Inspector.
230 kV was restored to Operable on 06/24/13 at 0200 PDT. Normal 500 kV Offsite Power remained operable and was unaffected by this event. Following further review, Diablo Canyon Power Plant (DCPP) has determined that the loss of the 230 kV system was not a condition that could have prevented fulfillment of the safety function of a system credited to mitigate the consequences of an accident. This power source is not considered to be a safety-related system that is credited to mitigate any accident as described in the DCPP UFSAR, Chapters 6 and 15 accident analyses. PG&E concludes that the emergency diesel generators are the only power source needed to fulfill the accident mitigation function, and they did not become inoperable as a result of this event. The licensee will be notifying the NRC Resident Inspector. Notified R4DO (Werner). | Emergency Diesel Generator | |
| ENS 48796 | 1 March 2013 05:54:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Bus Inadvertently De-Energized with Unit Defueled | On February 28, 2013, at 2154 PST, Unit 2 4kV ESF Bus G deenergized while attempting a repair to the bus automatic transfer circuitry. The deenergization of 4kV ESF Bus G initiated a start signal to Diesel Generator 2-1, which supplies emergency power to 4kV ESF Bus G. Diesel Generator 2-1 did not start due to being placed in manual control to prevent starting automatically during the repair. However, a valid actuation signal was generated to start Diesel Generator 2-1. As the Diesel Generator was shut down and in manual control, no actuation occurred. This is reportable as a valid system actuation that was not part of a pre-planned sequence during testing. Unit 2 is currently defueled, with the core offloaded into the spent fuel pool. No loss of cooling occurred as spent fuel pool cooling equipment had been selected to unaffected buses. The NRC resident has been notified. | 05000323/LER-2013-001 | |
| ENS 48400 | 11 October 2012 19:08:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Experienced an Automatic Reactor Trip on a 500Kv Line Differential Relay Actuation | On October 11, 2012, at 1208 PDT, the Diablo Canyon Power Plant Unit 2 500kV line differential relay actuated, resulting in a unit trip. The unit trip actuated the turbine trip. With the turbine tripped and Unit 2 operating above the P-9 50% power permissive, a reactor trip was initiated from the reactor protection system. All plant equipment responded as designed. Unit 2 is stable at normal operating temperature and pressure. All power transferred to the plant startup source without incident. Condenser vacuum was maintained. The cause of the differential relay actuation is under investigation. Based on personnel observation it appears to have been initiated due to a flashover to ground across the phase 'A' main bank transformer capacitive coupled voltage transformer. PG&E will issue a press release regarding the Unit 2 trip. NRC Resident Inspectors have been informed of this event.
This update to clarify the status of the auxiliary feedwater system during the above noted transient. Both motor driven pumps and the steam driven pump automatically started as expected. In addition, on October 11, 2012, at 12:26 PDT, the Diablo Canyon Power Plant Unit 2 turbine driven auxiliary feedwater pump was secured by plant operators manually closing the steam supply valve to the pump turbine. This action was performed in accordance with plant operating procedures after the operators had verified that the indicated steam generator levels were greater than the procedural requirement of 16%. However, because the steam generator low level bistables associated with the auxiliary feedwater actuation circuits had not yet cleared, the emergency safeguards actuation signal drove the steam supply valve back open, restarting the turbine driven auxiliary feedwater pump. Operators increased steam generator levels to clear the bistables and successfully reclosed the steam supply valve. The NRC Resident Inspector has been informed of this event. Notified R4DO (Hagar). | Steam Generator Reactor Protection System Auxiliary Feedwater | |
| ENS 46900 | 27 May 2011 19:12:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Unit 1 Startup Power Results in Starting All Emergency Diesel Generators | On May 27, 2011, at 1212 PDT, Unit 1 startup power was lost due to actuation of startup transformer 1-1 overcurrent/differential relay 51-87UT11-1, which isolated the 12 kV feeder to startup bus. The loss of offsite startup power caused all Unit 1 emergency diesel generators (EDGs) to start in the standby mode. The EDGs started as designed with no issues observed. No vital loads were affected as a result of the 12 kV bus loss. On May 27, 2011, at 1241 PDT, all Unit 1 EDGs were shut down and returned to auto. Subsequently, startup transformers 1-1 and 1-2 were energized, followed by the 12 kV underground loop, and on May 27, 2011, at 1337 PDT, Unit 1 startup power was declared operable. The cause of the event was a human performance error while attempting to test the startup transformer 2-1 overcurrent/differential relay. The maintenance electrical worker inadvertently installed a jumper on the overcurrent/differential relay for startup transformer 1-1. Unit 2 startup power was cleared at this time and there was no effect on Unit 2. The NRC Resident Inspector was notified. | Emergency Diesel Generator | |
| ENS 46894 | 26 May 2011 09:26:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Loss of Startup Power Due to Fault in Switchyard Causes All Edgs to Start | At 0226 PDT on May 26, 2011, Unit 1 startup power was lost following actuation of the 230 KV line pilot wire differential relay 287X. Loss of power to the Unit 1 12 KV startup bus produced an undervoltage signal that caused all Unit 1 emergency diesel generators (EDGs) to start in a standby mode. The EDGs started as designed with no problems observed. No vital loads were affected as a result of the 12 KV bus loss and subsequent undervoltage. The cause of actuation of the differential relay 287X is under investigation at this time. At 0255 PDT on May 26, 2011, all Unit 1 EDGs were shutdown and returned to their normal at-power standby configuration. Unit 2 startup power was cleared at this time and there was no effect on Unit 2. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 46856 | 17 May 2011 16:54:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generators Start Due to Maintenance Activities | At 0954 (PDT) on May 17, 2011, Unit 1 startup power was lost following actuation of the 230 kV line pilot wire differential relay 287X. Loss of power to the Unit 1 12 kV startup bus produced a startup undervoltage signal that caused all Unit 1 emergency diesel generators (EDGs) to start in a standby mode. The EDGs started as designed with no problems observed. No vital loads were affected as a result of the 12 kV bus loss and subsequent undervoltage. Actuation of differential relay 287X was due to maintenance activities on the safeguard relay board where differential relay 287X is installed. At 1056 on May 17, 2011, the Unit 112 kV startup power system was made available to supply power to the 4 kV vital buses. All Unit 1 EDGs were shutdown and returned to their normal at-power standby configuration, and differential relay 287X (was) reset. Unit 2 was unaffected by this event. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 46701 | 26 March 2011 21:49:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip When a Loss of the 2-1 Main Feedwater Pump Occurred | This notification provides the 4-hour non-emergency event report for the manual reactor trip of Diablo Canyon Power Plant Unit 2 in accordance with 10 CFR 50.72(b)(2)(iv)(B) 'RPS Actuation (scram)'. Additionally, this notification provides the 8-hour non-emergency event report of the automatic actuation of the auxiliary feedwater system as a result of the reactor trip in accordance with 10 CFR 50.72(b)(3)(iv)(A) 'Specified System Actuation'. On March 26, 2011 at 1449 PDT operators at Diablo Canyon Power Plant Unit 2 manually initiated a reactor trip in response to loss of main feedwater pump 2-1. The main feedwater pump is believed to have tripped due to non-radioactive water spray on its control console. The water spray was caused by leakage from the flange of the relief valve on the feedwater heater 2-1A condenser dump valve line. Emergency plan activation was not required. The unit is stable in mode 3 (Hot Standby) with offsite power being supplied to all buses via the 230 kV startup circuit. Diesel generators 2-1 and 2-2 remain OPERABLE in standby. Diesel generator 2-3 remains unavailable due to pre-planned maintenance. All rods fully inserted on the reactor trip. The reactor is being cooled by the auxiliary feedwater system with the condenser in service. All systems performed as designed with no unexpected pressure or level transients. ECCS actuation was not required. Automatic main feedwater isolation, auxiliary feedwater actuation, and steam generator blowdown isolation occurred as expected. Unit 1 was unaffected by this event and remains at 100% power. The licensee has notified the NRC Resident Inspector, San Luis Obispo County, and the State of California. | Steam Generator Feedwater Auxiliary Feedwater | 05000323/LER-2011-001 |
| ENS 44588 | 22 October 2008 03:51:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Manual Reactor Trip Due to Jellyfish Intrusion | On October 21, 2008, with both units operating at 100% power, Operators manually actuated the Unit 2 reactor protection system (RPS/reactor trip) due to high differential pressure (DP) across the circulating water pumps' intake traveling screens. The high DP resulted from a rapid influx of jellyfish. All systems responded as designed. All control rods fully inserted. Auxiliary feedwater actuated as designed. The grid is stable with power being supplied by 230 Kv startup power. Diesel generator (DG) 2-2 and 2-3 are operable in standby. DG 2-1 is inoperable due to scheduled maintenance. The traveling screens for the safety-related auxiliary saltwater system (ASW) are not degraded and are managing the influx of jellyfish with no significantly elevated DP. Unit 2 is stable in Mode 3 at normal operating temperature and pressure. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B), 'RPS actuation,' and 50.72(b)(3)(iv)(A), 'Specified System Actuation.' Operators reduced power on Unit 1 in response to the potential loss of normal flow to the condenser due to the jellyfish influx on the traveling screens. Currently, the traveling screens are maintaining DP within limits and the unit is stable at 50% power. Unit 2 decay heat removal is being performed by Auxiliary Feed Water to four steam generators blowing down via the 10% steam dumps to atmosphere. No other safety related equipment was out of service at the time of the trip. The licensee notified the NRC Resident Inspector.
The licensee issued a press release regarding this issue. Notified R4DO (Deese). | Steam Generator Reactor Protection System Auxiliary Feedwater Decay Heat Removal Control Rod | 05000323/LER-2008-002 |
| ENS 44419 | 17 August 2008 07:12:00 | 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 Declared Due to Fire in Main Bank Tranformer | At 0012 the licensee declared a Notice of Unusual Event due to a fire in Main Bank Transformer. The licensee is fighting the fire along with offsite assistance from CAL FIRE. The unit is shutdown and stable in mode 3. All control rods fully inserted on the trip. Decay heat is being removed via auxiliary feedwater to the steam generators steaming to the main condenser via the main condenser steam dumps. Emergency buses remain powered via offsite power with emergency diesel generators available if required. The licensee will notify the NRC Resident Inspector. Notified DHS (Barnes), FEMA (Blankenship). * * * UPDATE FROM J. DILLIS TO J. KOZAL AT 0501 EDT ON 8/17/08 * * * The fire is out. Plant personnel are assessing the damage in the main transformer area. The plant is stable in mode 3. The licensee will notify the NRC Resident Inspector. * * * UPDATE FROM J. DILLIS TO P. SNYDER AT 0542 EDT ON 8/17/08 * * * The licensee has terminated the Notice of Unusual Event at 0231 PDT due to the fact that the fire was out and no reflash occurred. The plant remains stable in mode 3. The licensee will notify the NRC Resident Inspector. Notified R4DO (Lantz), IRD (McMurtray), NRR EO (Wermiel), R4 (Collins), NRR (Grobe), DHS (Wallace), FEMA (Biscoe).
This is an update of the Notification of Unusual Event due to Fire in the 'C' phase of the main bank transformer, and Automatic Reactor Trip. This update is to report automatic actuation of the auxiliary feedwater system (AFW) and is being reported as an 8-hour non-emergency event in accordance with 10 CFR 50.72(b)(3)(iv)(A) 'System Actuation.' The unit is stable in mode 3 (Hot Standby) with offsite power being supplied to all buses via the 230 Kv startup circuit. All rods fully inserted on the reactor trip. All systems performed as designed. Operators are preparing to place the unit in Mode 5, while preparing to assess the damage to the 500 Kv transformer. All emergency diesel generators remain operable in standby. Unit 1 was unaffected by this event and remains at 100% power. This notification also constitutes a late 4 hour notification for the RPS actuation. To summarize, on August 16, 2008, at 2356 PDT, the 'C' phase 500 Kv transformer failed resulting in a fire and an automatic reactor trip, and automatic actuation of AFW. An unusual event was declared on August 17, 2008, at 0012 PDT due to the fire. Cal Fire was called to assist and support the on-site fire brigade. The fire was extinguished and the NUE was downgraded at 0231 PDT. The NRC Resident Inspector has been notified as well as State and local government agencies. The licensee intends to issue a press release. R4DO (Lantz) notified. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Main Transformer Main Condenser Control Rod | 05000323/LER-2008-001 |
| ENS 43393 | 29 May 2007 05:50:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Actuation Due to Momentary Undervoltage Condition | Unit 1 was making preparations to parallel to the grid on 5/28/07. All busses had been transferred to the Startup power supply (230 KV System). Operations then started a large 12KV motor, Circulating Water pump 1-2. Starting this large motor caused a momentary dip in startup voltage. This voltage dip was sufficient to pick up the undervoltage relay on vital bus G, causing a start of that bus's Emergency Diesel Generator (EDG). The bus remained supplied by the startup power supply, the EDG did not load. This valid start of EDG 1-2 is being reported per 10CFR50.72(b)(3)(iv)(A). The EDG was subsequently shut down and returned to it's normal standby condition. There is nothing unusual or not understood about this event and all systems functioned as required. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator | 05000275/LER-2007-003 |
| ENS 43391 | 27 May 2007 18:17:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip in Mode 3 During Control Rod Testing | On May 27, 2007, at 11:17 PDT, with Diablo Canyon Power Plant (DCPP) Unit 1 in Mode 3 (Hot Standby) during day 28 of refueling outage 14, operators manually actuated the reactor protection system. All rods fully inserted. Operators were performing Surveillance Test Procedure (STP) R-1 C 'Digital Rod Position Indictor Functional Test'. While control bank C was being re-inserted, at 42 steps withdrawn, rod N-13 indicated position changed to 24 steps withdrawn. In response to this 18 step deviation, and in accordance with STP R-1C, operators opened the reactor trip breakers. At the time of the event, the reactor coolant system remained at normal operating temperature and pressure, and all other shutdown and control rod banks remained fully inserted. This event is being reported as an 8-hour non-emergency event in accordance with 10 CFR 50.72(b)(3)(iv)(A) for a valid actuation of the reactor protection system. Unit-1 is maintaining temperature with atmospheric steam dumps and aux feed water. Electrical power is from 500 KV backfeed (normal electrical shutdown lineup). Control rod N-13 was being tested along with seven other control rods. The licensee notified the NRC Resident Inspector.
(STP) R-1C 'Digital Position Indicator Functional Test' was re-performed on control bank C to obtain additional information. At 1706 hours, while control bank C was being re-inserted, at 138 steps withdrawn, rod N-13 indication dropped to 126 steps withdrawn. Rod motion was stopped to gather data. All other shutdown and control banks remained fully inserted. At 1725 hours, operators manually actuated a reactor trip to fully insert all control bank C rods. All rods fully inserted. (STP) R-1C 'Digital Rod Position Indicator Functional Test' was re-performed on control bank C for further evaluation. At 1838 hours, while control bank C was being re-inserted, at 168 steps withdrawn, rod N-13 indication dropped to 150 steps withdrawn. Rod motion was stopped to gather data. All other shutdown and control banks remained fully inserted. At 1840 hours, operators manually actuated a reactor trip to fully insert all control bank C rods. All rods fully inserted. At the time of the event, the reactor coolant system remained at normal operating temperature and pressure, and all other shutdown and control rod banks remained fully inserted." R4DO (C. Johnson) notified. The licensee will notify the NRC Resident Inspector.
On May 27, 2007, in response to the initial occurrence of Rod N-13 slipping at 11:17 PDT, PG&E contacted the vendor and researched operating experience to develop a troubleshooting plan. Management approved a plan based on the vendor recommendations, which consisted of exercising all rods successfully five cycles. It was acknowledged that additional slipping events could occur requiring mitigation by opening the reactor trip breakers. Upon implementation of this plan, two similar events occurred and are described in the update above (provided on 05/27 at 2301 EDT). Following these events, Rod N-13 was successfully cycled five times without any deviation. Reactor startup was then allowed to proceed and there have been no further rod position deviation events. The licensee notified the NRC Resident Inspector. Notified R4DO (C. Cain) | Reactor Coolant System Reactor Protection System Control Rod | 05000275/LER-2007-002 |
| ENS 43360 | 12 May 2007 17:25:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generators Start Due to Loss of Off Site Start Up Power | On May 12, 2007, at 10:25 am PST, offsite startup power was lost to Diablo Canyon Units 1 and 2. This is the 8-hour non-emergency event notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) for a valid actuation of five emergency diesel generators (EDGs) due to the loss of startup power. Unit 1 is in a refueling outage in No Mode, with all fuel offloaded to the spent fuel pool (SFP). Auxiliary power is cleared for maintenance and offsite power was being provided by startup power. EDG 1-3 was also cleared for maintenance. On the loss of startup power, EDGs 1-1 and 1-2 auto-started and powered their loads. Since SFP cooling is not automatically loaded on the EDGs, Operators re-started a SFP Cooling pump. This re-established decay heat removal via component cooling water (CCW) (two of three pumps remained operable) and auxiliary saltwater pump 1-2. SFP temperature remained at approximately 105 degrees F. Unit 2 remained in Mode 1 at 100 percent power with auxiliary power supplying Unit 2 equipment. The loss of startup power caused all three EDGs to auto-start but they did not load. At approximately 11:30am PDT, startup power was restored to the site via the Mesa substation. Operators completed restoration of startup power to plant equipment by approximately 14:30 PDT. The cause of the loss of startup power was reported as a failed Morro Bay - Diablo transmission line. The licensee notified the NRC Resident Inspector. | Emergency Diesel Generator Decay Heat Removal | 05000275/LER-2007-001 |
| ENS 43047 | 12 December 2006 21:22:00 | 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 | Circulating Water Pump Failure Causes Explosion in Intake Structure | On December 12, 2006, at 1322 PST, while conducting power ascension operations with Unit 2 at approximately 25 percent power, an electrical failure occurred in the Unit 2 circulating water pump motor enclosure (CWP) 2-1. A loud bang and explosion was reported to the Unit 2 control room. At 1322 PST, an electrical transient was experienced on Unit 2 12-Kv non-vital bus 'D' which in turn caused reactor coolant pump (RCP) 2-2 and RCP 2-4 motor breakers to trip on 12kV non-vital bus 'D' undervoltage, initiating an automatic reactor trip. The reactor trip signal was initiated when 2 out of 4 RCP motor breakers opened. All control rods fully inserted in response to the reactor trip and all plant systems functioned as required. The auxiliary feedwater system for Unit 2 was manually actuated per plant procedures, before an auto-start signal for this system was generated. At 1356 PST, DCPP Fire Department first responders reported that the fire was out. At 1340 the licensee declared a Notice of Unusual Event (NUE), number 23, 'Confirmed Explosion Onsite'. Decay heat is being removed via auxiliary feedwater to the steam generators steaming to the atmospheric dump valves. Emergency power is being supplied via offsite power. The NRC resident inspector has been informed and was onsite at the time of the Unit 2 reactor trip. Diablo Canyon Power Plant (DCPP) Unit 2 is presently in Mode 3. There were no injuries associated with this event. Investigations into the electrical transient are ongoing. DCPP Unit 2 will remain in Mode 3 pending the results of this investigation. An estimated restart of DCPP Unit 2 is not known at this time. DCPP Unit 1 is unaffected and remains in Mode 1 at 100% power. * * * UPDATE AT 1730 EST ON 12/12/06 FROM T. CHITWOOD TO P. SNYDER * * * The licensee terminated the Unusual Event at 1430 PST. The Notice of Unusual Event was terminated due to the fact the licensee assessed the damage to the circulating water pump and determined that the explosion was caused by a breaker in the local area. The licensee fire brigade was satisfied that the fire was extinguished and that the damage was confined to the local area of circulating water pump 2-1. The licensee notified the NRC Resident Inspector. Notified the R4DO (Smith), NRR EO (Highland), IRD (Wilson), DHS (Inzer), and FEMA (Discoe). * * * UPDATE AT 1841 EST ON 12/12/06 FROM T. CHITWOOD TO J. ROTTON * * * The licensee provided an update to make the formal 4 hour notification pursuant to 10 CFR 50.72 (b)(2)(iv)(B) due to the RPS actuation while critical, and the formal 8 hour notification pursuant to 10 CFR 50.72 (b)(3)(iv)(A) due to the actuation of the auxiliary feedwater system. The licensee notified the NRC Resident Inspector. Notified the R4DO (Smith), and NRR EO (Highland). | Steam Generator Auxiliary Feedwater Control Rod | 05000323/LER-2006-004 |
| ENS 43042 | 11 December 2006 00:08:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rapid Shutdown and Manual Reactor Trip Due to High Reactor Coolant Pump Stator Temperature | On December 10, 2006, at 1608 PST, operators manually tripped the reactor while it was subcritical. This is an 8 hour notification per 10 CFR 50.72(b)(3)(iv)(A) for actuation of the reactor protection system while subcritical. On December 10, 2006, with Unit 2 operating in Mode 1 at 100% power, operators initiated an unplanned reactor shutdown due to indications of increasing stator temperature on reactor coolant pump 2-2. In accordance with plant procedures, when the reactor coolant pump stator temperature reached 300 degrees Fahrenheit, operators manually tripped the reactor and tripped the coolant pump 2-2. The reactor was subcritical at the time of the reactor trip but all rods had not been fully inserted. All control rods fully inserted in response to the reactor trip and all systems functioned as required. The auxiliary feedwater system was already in service as part of the shutdown. The grid is stable and with the exception of RCP 2-2, all major equipment including the 3 emergency diesel generators remain operable. Unit 2 electrical loads are being supplied by the normal startup power. The other 3 RCPs are maintaining forced circulation of the reactor coolant system. Investigations into the high RCP stator temperature are ongoing, the unit will remain in Mode 3 pending the results of these investigations. An estimated restart date is not known at this time. The NRC Resident Inspector has been informed and was onsite at the time of the trip. Unit 1 is unaffected and remains in Mode 1. The licensee stated that a press release related to this event is likely.
Licensee believes the cause of the high RCP stator temperature was due to a failed RTD.
Unit 2 was in Mode 3 (Hot Standby) at 0 percent power at the time of the manual trip. At the time of the Initial notification, all procedural requirements for declaring Mode 3 had not been met, but upon further review, PG&E has verified K effective was less than 0.99. The licensee notified the NRC Resident Inspector. Notified R4DO (Proulx). The original report was modified by this change. The 'RX Crit' field was changed from 'Y' to 'N'. The 'Init Power' field was changed from '5' to '0'. The 'Curr Power' field was changed from 'Startup' to 'Hot Standby'. | Reactor Coolant System Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Control Rod | 05000323/LER-2006-003 |
| ENS 42157 | 19 November 2005 23:51:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Diesel Generator Starts on a Vaild Bus Undervoltage Signal | On November 19, 2005 at 1551 Pacific Standard Time Diablo Canyon Unit 2 (was) in Mode 1 (Power Operation) at approximately 100% power when the 230kv offsite power supply breaker (#212) opened. Diesel generators 2-1, 2-2, and 2-3 auto-started on a valid bus undervoltage signal. The immediate cause of the event was the opening of the 230kv breaker (#212) due to a valid (but unintentional) startup transformer 1-1 sudden pressure switch activation due to maintenance being performed on the Unit 1 Startup Transformer 1-1. All equipment performed as expected. None of the diesel generators picked up load since the buses remained energized. The safety buses were powered by offsite power. Unit 1 is in a refueling outage and was not adversely affected. The diesel generators are currently being returned to a standby condition. The licensee will notify the NRC Resident Inspector. | 05000323/LER-2005-002 | |
| ENS 41250 | 9 December 2004 10:04:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Actuation While Shutdown Due to Failed Fuse | During performance of Surveillance Testing on Intermediate Range Nuclear detector, a control power fuse blew. The plant was in Mode 3, and received a Reactor Trip signal which opened the Reactor Trip Breakers. All rods were on the bottom prior to the trip signal. Testing of Intermediate Range. Detectors has been suspended pending investigation. The plant remains in Mode 3. The licensee will inform the NRC resident inspector.
This is a retraction of NRC Event # 41250 'RPS Actuation While Shutdown.' On December 9, 2004, at 0204 PST, with Unit 2 in Mode 3 (Hot Standby), during testing of Intermediate Range Nuclear Detectors, a control power fuse blew, resulting in an actuation signal, and causing actuation of the Reactor Protection System (RPS). The reactor trip breakers opened, however, since all rods were fully inserted, no rod motion occurred. At 0448 PST, this event was reported as an eight hour non-emergency notification per 10 CFR 50.72 (b)(3) (iv)(A). Based on a subsequent review of the event against NUREG-1022 Rev. 2, the actuation was confirmed to be INVALID due to a blown fuse and not an actuation of the intermediate range neutron flux detector. Furthermore, with the unit in Mode 3 with all rods on the bottom, the RPS actuation had no effect, as the safety function had already been completed. Therefore, the event is not reportable under 10 CFR 50.72 or 10 CFR 50.73, and the event notification is retracted. The licensee notified the NRC Resident Inspector. Notified R4 DO (G. Pick) | Reactor Protection System | |
| ENS 41168 | 2 November 2004 19:05:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Diesel Generator Start on a Valid Bus Undervoltage Signal | On November 2, 2004, at 11:05 PST, with Diablo Canyon Unit 2 in Mode 6 (Refueling), Emergency Diesel Generator (EDG) 2-1 auto started on an unplanned actuation signal from a valid 4160 volt Bus G undervoltage signal. All equipment responded as designed. After the EDG started, the auxiliary feeder breaker opened, and loads were automatically sequenced onto the EDG. At the time of the event, test equipment was being connected in preparation for an instrumented manual test start of EDG 2-1 prior to maintenance. On November 2, 2004, at 1708 PST, operators transferred Bus G to auxiliary power and shutdown DG 2-1. Prior to the event, Bus G was being supplied by auxiliary power, with startup power cleared for planned maintenance. Bus G was being prepared to be cleared for maintenance, therefore, required equipment was in-service on the other buses. Buses F and H were unaffected by this event and remain operable on auxiliary power, with EDG 2-2 (Bus H) and 2-3 (Bus F) operable. The following decay heat removal trains are powered from Bus G: Residual Heat Removal Pump 2-1 remained in standby, Component Cooling Water Pump 2-2 started on the transfer to EDG, and Auxiliary Saltwater Pump 2-2 re-started. Unit 2 is in day 9 of a refueling outage, with the reactor head removed, the refueling cavity filled, and the upper internals installed with rods latched. Unit 1 was unaffected and continues to operate in Mode 1 (Power Operation) at 100 percent power. The cause of the Bus G undervoltage signal is being investigated. The licensee stated that no undervoltage was seen on any other equipment but the undervoltage relays on Bus G did sense an undervoltage. Residual Heat Removal was unaffected by the event. Bus G has been returned to its pre-event configuration and is considered operable. The other EDGs were fully operable at the time of the event and there were no significant LCOs at the time. All systems functioned as required. The licensee has notified the NRC Resident Inspector | Emergency Diesel Generator Residual Heat Removal Decay Heat Removal | 05000323/LER-2004-001 |