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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56877 | 2 December 2023 12:10:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Auto Start of Emergency Diesel Generator | The following information was provided by the licensee via email: At 0610 CST on 12/2/2023, with Unit 2 in Mode 1 at 100 percent power, the South Texas Project switchyard south electrical bus was de-energized. Emergency diesel generator (EDG) '22' automatically started in response to the loss of offsite power on the train 'B' engineered safety feature (ESF) electrical bus. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of an emergency AC electrical power system (50.72(b)(3)(iv)(B)(8)). All required loads were successfully started. 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 initial loss of the south electrical bus, partial loss of off-site power, put the plant in a 24 hour limiting condition for operation (LCO) in accordance with (IAW) technical specification (TS) 3.8.1.1.E. Power was restored to the train 'B' ESF bus via an alternate offsite power source and the EDG was returned to its automatic standby condition. Currently, the plant is in a 72 hour LCO IAW TS 3.8.1.1.A. | Emergency Diesel Generator | |
| ENS 55692 | 6 January 2022 12:03:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Auto Start Emergency Diesel Generator | At 0603 CST on 1/6/2022, with Unit 2 in Mode 1 at 100 percent power, the South Texas Project (STP) south switchyard electrical bus was de-energized momentarily and re-energized approximately 40 seconds later. Emergency Diesel Generators (EDG) 22 automatically started in response to loss of offsite power on Train B Engineered Safety Feature (ESF) Bus. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of an emergency AC electrical power system (50.72(b)(3)(iv)(B)(8)). 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: Unit 2 is in a 72 hour LCO per TS 3.8.1.1.A for the loss of one offsite power supply. The plant is in a normal electrical lineup. There was no impact on Unit 1. | Emergency Diesel Generator | |
| ENS 54605 | 24 March 2020 15:46:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation - Auto Start Emergency Diesel Generators | At 1046 CDT on 3/24/2020, with Unit 1 defueled and Unit 2 in Mode 1 at 100 percent power, the South Texas Project (STP) South switchyard electrical bus was deenergized. This resulted in a loss of power to Standby Transformer 2 which was supplying power to the Engineered Safety Features (ESF) 4160v busses for Unit 1 A- and C-Trains, and Unit 2 B-Train. Emergency Diesel Generators (EDGs) 11, 13, and 22 automatically started in response to the undervoltage condition. The cause for the loss of the South switchyard electrical bus was an error in relay testing in the switchyard. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of emergency AC electrical power systems (50.72(b)(3)(iv)(B)(8)) as well as the Unit 1 A- and C-Train and Unit 2 B-Train Reactor Containment Fan Coolers (50.72(b)(3)(iv)(B)(7)) and Unit 2 B-Train Auxiliary Feedwater pump (50.72(b)(3)(iv)(B)(6)). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 54351 | 25 October 2019 01:51:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Undervoltage Actuation During Edg Sequencer Maintenance | On October 24, 2019, at 2051 Central Time, while performing Train C Sequencer maintenance, a valid undervoltage actuation signal was sent to Unit 2 Emergency Diesel Generator (EDG) 23. The ESF Train C bus loads were shed but EDG 23 did not automatically start because it had been placed in Pull-To-Stop to support the sequencer maintenance activities. EDG 23 was taken out of Pull-To-Stop by Control Room staff to allow it to auto start and load the bus. As a result of the bus strip signal, the in service Spent Fuel Pool Cooling Pump secured. Spent Fuel Pool Cooling was restored with no measurable increase in pool temperature. The reactor was not critical and reactor decay heat removal was not challenged throughout the event. This actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A) due to the automatic actuation of a system listed in 10 CFR 50.72(b)(3)(iv)(B). The NRC Resident Inspector has been notified. | Emergency Diesel Generator Decay Heat Removal | |
| ENS 51897 | 2 May 2016 01:21:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Generator Lockout | At 2021 (CDT) on 05/01/2016 Unit 1 automatically tripped due to a generator lockout. Relay 86/G1 actuated. The generator lockout resulted in a Unit 1 turbine trip and a reactor trip. Auxiliary Feedwater and Feedwater Isolation actuated as designed. All Control Rods fully inserted. No primary or secondary relief valves opened. There were no electrical problems. Normal operating temperature and pressure is (being maintained). There were no significant TS LCO's entered. This event was not significant to the health and safety of the public based on all safety systems performed as designed. Unit 2 was not affected. Unit 1 is stable in Mode 3, with decay heat being removed via dump valves to the condenser. The cause of the generator lockout is under investigation. The licensee notified the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Control Rod | |
| ENS 51687 | 27 January 2016 05:25:00 | 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 Loss of Feedwater to a Single Steam Generator | At 2325 (CST) on 01/26/2016, Unit 1 was manually tripped due to loss of Feedwater on 'C' S/G (Steam Generator). The loss of Feedwater was a result of a failure on 'C' S/G Main Feedwater Regulating Valve that caused the valve to travel closed with no Operator action. Auxiliary Feedwater and Feedwater Isolation actuated as designed. All Control and Shutdown Rods fully inserted. Intermediate Range Nl 36 (Nuclear Instrument) failed above P10 and, as a result, Source Range Nuclear Instruments were manually energized. No primary or secondary relief valves opened. There were no electrical problems. Normal operating temperature and pressure (NOT/NOP) is 567 degrees F and 2235 psig. There were no significant TS LCOs entered. This event was not significant to the health and safety of the public based on all safety systems performed as designed. Unit 2 was not affected. Decay heat removal is being controlled via Steam Dumps. Offsite power is in the normal electrical lineup. The NRC Resident Inspector has been notified. | Steam Generator Feedwater Auxiliary Feedwater Decay Heat Removal | |
| ENS 51615 | 21 December 2015 21:33:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Feedwater Isolation | At 1519 (CST), the Main Turbine was tripped due to an Oscillating Governor Valve 2 (cause not known). At 1533, Unit 1 was manually tripped due to a feedwater isolation P-14 (caused by steam generator swell induced high steam generator level, resulting in) steam generator low level (after the isolation). Aux feedwater actuated as designed. All Control and Shutdown Rods fully inserted. Intermediate Range NI 36 failed above P-10 so SR-Nis (source range nuclear instruments) were manually energized. No primary relief valves lifted. All Steam Generator PORVs (power operated relief valves) opened. There were no electrical bus problems. Normal operating temperature and pressure (NOT/NOP) is 567F and 2235 psig. There were no significant TS LCOs (Technical Specification limiting conditions for operations) entered. This event was not significant to the health and safety of the public based on all safety systems performed as designed. Unit 2 was not affected and continues to operate at 100% power. The licensee has notified the NRC Resident Inspector. | Steam Generator Feedwater Main Turbine | |
| ENS 49666 | 20 December 2013 04:33:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Actuation of Main Steam Isolation Valves That Was Not Part of a Preplanned Sequence | While in Mode 3 in preparing the Unit 2 secondary plant for startup, conditions occurred where it became necessary to break vacuum on the main condenser. Procedures directed closing of the main steam isolation valves. Instead of shutting each main steam isolation valve individually, a manual main steam isolation actuation was initiated through the solid state protection system (SSPS) to close the valves. This actuation of SSPS was a valid signal. The actuation was not part of a preplanned sequence. This notification is supported by the guidance of NUREG-1022, Revision 3, 'Event Reporting Guidelines 10CFR50.72 and 50.73.' In part, the guidance states: 'The Commission is interested in both events in which a system was needed to mitigate the consequences of an event (whether or not the equipment performed properly) and events in which a system actuated unnecessarily.' The manual actuation was not initiated to mitigate the consequences of an actual event. However, the method of closing the main steam valves for this condition did not specifically require that the valves should be closed by initiating a main steam isolation signal and therefore, the safety system was unnecessarily actuated. Therefore, this notification is being made pursuant to 10 CFR 50.72(b)(3)(iv)(A) as an event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) that was not part of preplanned sequence during testing or reactor operation. The system listed in paragraph (b)(3)(iv)(B) is (2) main steam isolation valves. The licensee notified the NRC Resident Inspector. | Main Steam Isolation Valve Main Condenser Main Steam | |
| ENS 48659 | 8 January 2013 22:40:00 | 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 Main Transformer Fire | Fire in Unit 2 main transformer 2A. Reactor trip. Two train of offsite power lost to Unit 2. An Unusual Event was declared based on EAL HU-2 - Fire or explosion in protected area or switchyard which affects normal plant operations. At 1655 CST, South Texas Unit 2 declared an Unusual Event due to a main transformer fire. Unit 2 tripped from 100% power and is currently at 0% power in Mode 3. The transformer fire is out. In addition to the loss of the main transformer, several safety related electrical busses and non-safety electrical busses lost offsite power. The appropriate emergency diesel generators started and powered the safety related busses. Unit 2 is currently stable and on natural circulation due to the loss of power to the reactor coolant pumps. Auxiliary feedwater is functioning as required and decay heat is being removed through the steam generator atmospheric relief valves. Unit 1 was unaffected by the event. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email.
On January 8, 2013, at 1640 CST, a failure of the Unit 2 Main Transformer occurred which resulted in a Unit 2 automatic trip. The failure of the main transformer resulted in a fire and damage to the transformer. The onsite fire brigade responded to the fire. The fire was declared under control at 1649 CST and declared out at 1656 CST. No offsite assistance was required. An Unusual Event was declared at 1655 CST for initiating condition HU-2 (Fire or explosion in protected area or switchyard which affects normal plant operations) due to the main transformer fire. Due to the site electrical lineup at the time, the loss of the main transformer resulted in a loss of power to 4160 ESF buses 2A and 2C, and associated Standby Diesel Generators 21 and 23 started as required and loaded on to their respective buses. 4160 ESF bus 2B remained energized from offsite power during this event and Standby Diesel Generator 22 did not start since an undervoltage condition did not exist on its ESF bus. All three (3) motor-driven and the steam-driven Auxiliary Feedwater Pumps started as required. The Main Steam Isolation Valves were closed in accordance with procedure to limit plant cooldown. Decay heat is being removed via Auxiliary Feedwater with Steam Generator Power Operated Relief Valves. Following the reactor trip, Pressurizer Power Operated Relief Valve 656A momentarily lifted and re-closed. There were no personnel injuries and no radiological release as a result of this event. A press release has been issued. The plant is currently stable in Mode 3 and the cause of the event is under investigation. The Unusual Event was terminated at 1947 CST on 1/8/2013. The licensee notified the NRC Resident Inspector. Notified R4DO (Gaddy), NRR (Leeds), R4 (Reynolds), IRD (Gott), NRR EO (Hiland). Notified DHS SWO, FEMA, USDA, HHS, DOE, DHS NICC, EPA, and NuclearSSA via email. | Steam Generator Emergency Diesel Generator Main Steam Isolation Valve Auxiliary Feedwater Main Transformer | |
| ENS 48651 | 4 January 2013 15:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 2 Manually Tripped After Two Shutdown Control Rods Unexpectedly Dropped During Surveillance Testing | On January 4, 2013, at 0941 hours (CST), Unit 2 was manually tripped after 2 shutdown rods unexpectedly dropped during monthly control rod surveillance testing. Shutdown Bank C rods were being inserted in accordance with the surveillance procedure, when 2 rods in Shutdown Bank E (D-8 and M-8) unexpectedly dropped. This met the criteria for a manual reactor trip, which was immediately performed. The appropriate procedures were entered to mitigate the transient and all systems responded as designed. Unit 2 is currently in Mode 3 and the cause of the 2 dropped rods is under investigation. All three (3) motor-driven and the steam-driven Auxiliary Feedwater Pumps started as required and have since been secured. Decay heat is being removed using normal startup feedwater with steam discharge to the main condenser via the bypass valves. Unit 2 is in a normal post trip electrical lineup. The licensee informed the NRC Resident Inspector. | Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 47485 | 29 November 2011 09:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip on Generator Lockout | At 0329 CST on 11/29/2011, Unit 2 experienced an automatic Reactor Trip while the plant was stable at 100% power in Mode 1. All systems actuated as designed. The reactor trip was caused by Generator Lockout. All ESF systems actuated as designed. The following systems actuated: Auxiliary Feedwater (AFW) and Feedwater Isolation. All control rods fully inserted. Steam Dump system valve FV 7485 failed open and was manually isolated. This caused a letdown isolation that was restored. No primary/secondary relief valves lifted. There were no electrical bus transfer problems. Normal operating temperature and pressure (of) 567 degrees F and 2235 psig (is being maintained). There were no significant TS LCOs entered. The electrical grid is stable and is supplying power to the plant via a normal shutdown electrical line-up. Decay heat is being removed via steam dumps to the condenser with AFW supplying the steam generators. There was no effect on Unit 1. The licensee has notified the NRC Resident Inspector.
The licensee has issued a press release concerning this event. The NRC Resident Inspector will be notified. R4DO (Farnholtz) informed. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 46387 | 3 November 2010 15:21:00 | 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(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to an Explosion in the Protected Area Which Affects Normal Operations | On November 3, 2010 at 1021 CDT, operators attempted to start the Unit 2 startup feedpump to support a two hour maintenance run. The feeder breaker for the startup feedpump exploded causing an undervoltage condition on Auxiliary Bus 1H and Standby Bus 1H which resulted in an automatic reactor trip due to reactor coolant pump undervoltage. All control rods inserted into the core. At 1038 CDT, the site declared an UNUSUAL EVENT (HU-2) due to an explosion in the protected area which affects normal plant operations. The standby diesel generator (EDG-23) started and loaded to the 'C' train loads which sequenced properly. The auxiliary feedwater system automatically started as expected providing feedwater to the steam generators. The normal feedwater pumps were secured. Decay heat is being removed from Unit 2 using the normal steam dump valves to the main condenser. There is no primary to secondary leakage. The plant is stable and in MODE 3 with no challenges to reactor safety. There was no impact on Unit 1. At 1240 CDT the licensee terminated the UNUSUAL EVENT. The licensee is investigating the cause of the breaker explosion and if the other standby diesel generator (EDG-21) should have started due to the undervoltage condition. The licensee notified the NRC Resident Inspector.
A News Release is being planned so this condition is also being reported pursuant to 10 CFR 50.72(b)(2)(xi). During the electrical fault condition, some Train A components stopped running although no Train A low voltage ESF actuation occurred, The reason for this occurrence is not fully understood, The breaker malfunction did not result in a fire. The licensee will notify the NRC Resident Inspector. Notified: R4DO (Spitzberg); NRR (Thorp); R4RA (Collins); NRR (Boger); IRD (Gott) DHS (Hill); FEMA (Hollis)
The licensee called to correct an editorial error in the original report above. The second sentence of the original report above should refer to Auxiliary Bus 2H and Standby Bus 2H since this is a Unit 2 issue. Notified: R4DO (Spitzberg). | Steam Generator Feedwater Auxiliary Feedwater Main Condenser Control Rod | |
| ENS 46298 | 1 October 2010 00:04:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Standby Diesel Generator Autostart Due to Loss of a Switchyard Bus During Maintenance | At 1904 (CDT) on 9/30/2010, the South Texas Project (STP) North switchyard bus was lost due to a Transmission & Distribution Service Provider (TDSP) human performance error that occurred while performing maintenance on breaker Y-0530. This resulted in a loss of power to Standby transformer 1 which was supplying power to the Unit 1 B Train Engineered Safety Features (ESF) 4160v bus. The B Train Standby Diesel Generator automatically started due to the Loss of Offsite Power (LOOP) on its associated bus. The Mode II (LOOP) ESF loads sequenced onto the bus. All safety related equipment responded as expected. Action (e) of Technical Specification (TS) 3.8.1.1, 'AC Electrical Power Sources', was momentarily entered due to the loss of two independent offsite circuits while the North bus was de-energized. The North bus was de-energized for approximately 5 minutes. Action (a) of Technical specification (TS) 3.8.1.1, 'AC Electrical Power Sources', was entered due to the loss of one independent offsite circuit. All Technical Specification Limiting Condition of operation have been exited at this time. An 8-hour notification is required for this event due to the valid actuation of safety related equipment as described in 10CFR50.72 (b) (3) (iv) (A). (A notification is required for) any event or condition that results in valid actuation of any of the systems listed in paragraph (b) (3) (iv) (B) of this section, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The NRC Resident Inspector has been notified. Unit 2 briefly entered a Technical Specification Limiting Condition of Operation, while all electrical buses remained energized. | ||
| ENS 46191 | 20 August 2010 20:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to an Inadvertent Turbine Trip Signal During Testing | At 1525 (hrs. CDT) on 08/20/10, Unit 1 experienced an automatic reactor trip while the plant was stable at 100% power in Mode 1. All systems actuated as designed. The reactor trip was caused by an inadvertent turbine trip signal initiated during testing. All ESF (Engineered Safety Features) systems actuated as designed. The following systems actuated: Auxiliary Feed Water and Feed Water Isolation. All control and shutdown rods fully inserted. The plant is currently stable at normal operating pressure and temperature with decay heat being removed via steam dumps to the condenser. No primary or secondary relief valves lifted during the transient. The plant is in its normal shutdown electrical lineup with no problems noted. The trip was uncomplicated. The turbine trip signal was generated by a human performance error during reactor trip breaker testing. The licensee has notified the NRC Resident Inspector. | Auxiliary Feedwater | |
| ENS 44605 | 26 October 2008 04:42:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Safety Injection Signal Due to Maintenance Error | At approximately 2342 on 10/25/2006, while performing maintenance on the solid state protection system in Mode 5 cold shutdown, Unit 2 received an automatic safety injection signal which resulted in all three ESF Diesel Generators starting and a containment ventilation isolation and containment phase A isolation. All safety injection pumps were in pull-to-lock per plant conditions so that the pumps did not start and no water was discharged into the reactor coolant system. As a result, the ESF Diesel Generators started but did not load as designed. The residual heat removal pumps were stripped from the ESF electrical busses due to the actuation. The first residual heat removal pump was restored within 4 minutes upon the loss of residual heat removal cooling and the second pump was restored within 6 minutes. The residual heat removal system heat exchangers were bypassed at the time of the event and the plant was being allowed to heat up. The cause of the automatic safety injection signal was an inadvertent removal of the block for the low pressurizer pressure safety injection system during the maintenance. Therefore, the signal was a valid signal initiated in response to a parameter satisfying the requirements for initiation of the safety function of the system. Although the actuation was the result of a valid signal, safety injection was not required in this Cold Shutdown Mode of Operation. This notification is being made under 10CFR50.72(b)(3)(iv) as an event that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section. This actuation was initially determined to be due to an invalid signal. Upon further review, it was determined at 1745 on 10/27/08 that the actuation was due of a valid signal. The licensee notified the NRC Resident Inspector. | Reactor Coolant System Residual Heat Removal | |
| ENS 40473 | 23 January 2004 22:16:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to High-High Steam Generator Level | We make the following report per 10CFR50.72(b)(2)(iv)(B). At 16:16 CST Unit 1 Reactor automatically tripped from full power due to (actual) high-high level in 1B Steam Generator. Prior to the trip, vital distribution panel 1201 lost power when it's normal power supply inverter failed. Steam Generators 1A and 1B levels were selected to instruments from this power supply (and therefore generating a false low level output). Operators were in the process of taking manual control of 1A and 1B Main Feed regulating valves when the Main Turbine trip was actuated due to the high level in 1B Steam Generator. A reactor trip occurred due to the Turbine Trip above 50% power. The unit is stable at 567 degrees and 2235 pslg. We also make the following report per 10CFR50.72(b)(3)(iv)(A). Following the reactor trip the Auxiliary Feed Water System automatically actuated on (actual) low steam generator level. This is normal for a trip in the Unit 1 from full power. The following information is also provided: All control rods fully inserted. No primary reliefs lifted. Technical Specification 3.8.3.1 action b was entered due to the vital distribution panel not being energized from its normal source. (inverter). It is currently power from it's voltage regulator. Decay heat is currently being removed via the steam dumps. The plant electrical system responded normally and all emergency diesel generators remain in standby. All ECCS systems remain operable. There are no primary to secondary leaks. The licensee notified the NRC Resident Inspector. | Steam Generator Emergency Diesel Generator Auxiliary Feedwater Control Rod |