Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 40380 | 9 December 2003 21:38:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unable to Achieve Cold Shutdown Using Natural Circulation at a Cooldown Rate of 50 Degrees Per Hour | At 1538 on 12/09/03, it was determined that the plant could not achieve cold shutdown using Natural Circulation with the available volume of the Auxiliary Feed Water Storage Tank (AFWST) and at an allowed plant cooldown rate of 50 degrees F/hour. The plant has a functional objective to achieve Plant Shutdown from Operating Condition to Cold Shutdown Condition with the Natural Circulation Cooldown process with the volume of water available in the AFWST. Natural Circulation cooldown is required during Loss of Offsite Power (LOOP), long term cooling, Safety Grade Cold Shutdown and Appendix R type fire with LOOP. This condition also applies to Unit 2. STP (South Texas Project) changed from a Reactor Vessel Head (RVH) T-Hot Plant to a RVH T-Cold Plant after replacement of the plant's steam generators. As a result, the plant cooldown rate under natural circulation conditions was changed from 25 degrees F/hour to 50 degrees F/hour based on a Westinghouse evaluation. While performing confirmatory calculations to address a fire safe shutdown issue, it was determined that the plant can be cooled down with two steam generators using natural circulation. However, at a cooldown rate of 50 degrees F/hour, the two steam generators not being used for cooldown would stagnate with hot water in the primary side tubes. During the depressurization process, the hot water flashes to steam such that depressurization below approximately 1000 psig is significantly challenged. This results in filling the pressurizer and a loss of RCS pressure control. The fluid in the RCS (Reactor Coolant System) loops not receiving AFW will remain at saturated conditions until cooling is achieved by other means. Cooldown and depressurization of the RCS in this configuration will take substantially longer than assumed in the current analyses. Although there is adequate core cooling, the plant will be in a condition that is outside the existing design basis and not specifically addressed by the Emergency Operating Procedures. Therefore, this condition resulted in the plant being in an unanalyzed condition that significantly degraded plant safety. Thermal Hydraulic Analyses at a cooldown rate of 25 degrees F/hr demonstrated that natural circulation will occur in the two steam generators that are not being used for cooldown such that the stagnated hot water conditions would not occur in these steam generators. This would allow for depressurization of the reactor coolant system and successfully meet the functional objective. A compensatory action is in place to limit natural circulation cooldown rate to no greater than 25 degrees F/hr. The licensee has notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Auxiliary Feedwater | |
| ENS 40612 | 27 March 2004 04:30:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Unanalyzed Condition Regarding Hhsi Flush Line Isolation Valve Leakage - | While performing a High Head Safety Injection (HHSI) Train-B inservice test, a Plant Operator noted leakage coming through HHSI Flush Line Isolation Valve #SI-0120B. This valve is located on a flush line coming from the discharge of the HHSI pump and directs water to the Safety Injection Pump room sump located within the Fuel Handling Building. The operator investigated and determined that the leakage was due to the valve not being fully seated. The Licensee has come to the conclusion that the estimated leakage through the valve exceeded the allowable value to remain in compliance with 10 CFR 50, Appendix A, GDC 19 limits for Control Room Envelope during the post LOCA Containment recirc phase of a large break LOCA. This condition resulted in an unanalyzed condition that significantly degraded plant safety requiring a notification to the NRC within eight hours per 10 CFR 50.72 (b) (3) ii. Subsequently, flush line valve #SI-0120B has been fully seated, and the leakage stopped. The licensee considers this valve to be operable. The Licensee notified the NRC Resident Inspector. | Control Room Envelope | |
| ENS 45631 | 8 July 2008 03:47:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Loss of Fire Suppression Capability | This is an 8-hour notification being made in accordance with 10CFR 50.72(b)(3)(ii)(B) for an event or condition that results in the plant being in an unanalyzed condition that significantly degrades plant safety. This notification is being made as the result of the re-review of a July 7, 2008 occurrence which resulted in an inadvertent isolation of a large portion of the (fire suppression) ring header affecting all Unit 2 fire suppression and a portion of Unit 1 fire suppression. The Unit 1 ring header isolation did not have an affect on the fire safe shutdown capability of Unit 1. However, three areas in Unit 2 which credit the availability of fire suppression to assure that the safe shutdown capability could have been achieved did not have fire suppression for approximately 3 hours. A Licensee Event Report will be submitted within 60 days. The licensee notified the NRC Resident Inspector. | 05000499/LER-2010-002 | |
| ENS 46146 | 3 August 2010 20:06:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Potential Safety System Functional Failure of the Accident Mitigating Function | On 8/3/10 South Texas Project Unit 2 was in a scheduled A Train work week with the following equipment inoperable for planned maintenance; Essential Cooling Water Pump, Essential Chiller, Component Cooling Water Pump, Engineered Safety Function (ESF) Diesel Generator (DG), High Head Safety Injection (HHSI) pump, Low Head Safety Injection (LHSI) pump, and Containment Spray (CS) pump. At 0754 (CDT) on 8/3/10 the B train sequencer trouble alarm was received. The immediate operability determination was the sequencer remained operable. It was later identified during testing that the sequencer was inoperable. The B train sequencer was declared inoperable at 1506 (CDT) on 8/3/10. Due to loss of the automatic load sequencing support function, all associated train B safety equipment that is sequenced on the B train 14.16 kv bus during a Mode 1 Safety Injection (SI) was also declared inoperable. This condition resulted in an inoperable condition on two out of three safety trains for the accident mitigating function including the A and B train HHSI, LHSI, and CS pumps. All C train safety injection pumps remained operable. Pending a formal operability determination, this is conservatively considered to be a safety system functional failure of the accident mitigating function. This was determined to be reportable within 8 hours as required by 10 CFR 50.72(b)(3)(v)(D). The B train trouble alarm, an auto test feature, was discovered by operators during their rounds. The licensee entered their configuration risk management plan within the 1 hour as required. Currently, the licensee is working on completing the scheduled A train maintenance and restoring operability sometime in the morning. Also, a work package is under development to repair the faulty B train sequencer. The risk based time limit for restoring operability requires completion by 0449 (CDT) on 8/8/10. Unit 1 is unaffected and continues to operate at 100% power. The licensee informed the NRC Resident Inspector.
The purpose of this update is to retract the notification made in ENS Report #46146 (August 3, 2010). Following the ENS notification, troubleshooting determined the cause of the Train B sequencer alarm to be the failure of an Output Mode I Actuation Timing Switch Module. An engineering evaluation of the event has been completed and determined that a failure of this module did not affect the ability of the ESF load sequencer to perform its design function. Therefore, the Train B sequencer and associated Train B ESF equipment remained technically operable during the time that Train A equipment was inoperable due to scheduled maintenance, and a condition reportable per 10 CFR 50.72(b)(3)(v) did not exist. The licensee will notify the NRC Resident Inspector." Notified R4DO (Walker). | Containment Spray | |
| ENS 47491 | 20 November 2011 11:46:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Turbine Trip Protection Disabled While in Mode 3 | On November 20, 2011 at 0546 hours (CST), STP Unit 2 transitioned modes from Mode 4 to Mode 3. Prior to the mode change, all Solid State Protection System (SSPS) generated turbine trip signals were defeated by a maintenance work activity that installed a jumper in both channels (Train R and S) of non-class relays to the turbine trip circuit. The SSPS signals to the non-class relays that were defeated by the jumpers included the turbine trip from reactor trip breakers open (P4), turbine trip from a reactor trip signal (P-16), and the turbine trip from Steam Generator HI- HI (P-14). T.S. 3.3.2 Items 5a (P4) and 5b (P-14) are required in Modes 1, 2, and 3. The jumpers were removed around 0930 on November 20, 2011 with U2 still in Mode 3. Both the UFSAR and TS bases identify that the turbine trip mitigates the consequences of an accident. The TS bases states that an ESFAS initiated turbine trip mitigates the consequences of a steam line break or loss of coolant accident. The accident analysis for SGTR also assumes a turbine trip on a reactor trip to isolate the steam path. Although Unit 2 was in Mode 3, with the reactor trip breakers open, and turbine throttle valves closed while the jumpers were installed, this condition is conservatively considered to be a safety system functional failure. If not corrected, this condition could have prevented the fulfillment of the accident mitigating and control of the release of radiation safety functions. A review of the performance of this activity in previous outages was conducted. It was identified that during 2RE14 in April of 2010, a work package for this activity was not closed until after Mode 3. The 60 day LER will address if the jumpers were installed in Mode 3 in April, 2010. This was determined to be reportable within 8 hours as required by 10 CFR 50.72(b)(3)(v) parts (C) and (D). The licensee did not determine the reportability of this event until 1415 CST on 11/30/11. The licensee has notified the NRC Resident Inspector. | Steam Generator | |
| ENS 48277 | 22 August 2012 17:36:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Missing Flood Seal | During flooding walkdowns being performed on August 22, 2012, with the unit at 100 percent power, South Texas Project Unit 2 discovered the potential for water intrusion into the 10' Elevation Electrical Auxiliary Building (EAB) via a 2-inch underground conduit that was found to be missing its flood seal. It has been determined that the missing flood seal compromised the external flood design controls for the EAB. If flooding of the 10 (foot) EAB were to occur as a result of the missing flood seal, the operability of the Train A Engineered Safety Features (ESF) switchgear and the ESF Sequencers for all three Standby Diesel Generators could have been affected. Additionally, the Qualified Display Parameter System process cabinets (which control Auxiliary Feedwater flow and Steam Generator PORVs) and the Auxiliary Shutdown Panel are also located on the 10' Elevation. Repairs have been made and the 2-inch conduit is sealed. The event is being reported under 10 CFR 50.72(b)(3)(ii)(B) for Unit 2 being in an unanalyzed condition that significantly degraded plant safety, and under 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified.
The purpose of this call is to retract the notification made on 09/05/2012, Event Number 48277. Further analysis indicates that water intrusion resulting from the missing 2-inch conduit seal would not have been sufficient to affect the operability of the equipment located on the 10-foot elevation of the Unit 2 Electrical Auxiliary Building. It has been determined that the maximum water depth would not have exceeded 2 inches in depth and all safety related equipment on the 10-foot elevation is greater than 2 inches above the floor, therefore there would be no impact to any safety-related equipment. Accordingly, this event notification is being retracted. The licensee will notify the NRC Resident Inspector. Notified the R4DO (Geoffrey Miller). | Steam Generator Auxiliary Feedwater | |
| ENS 49490 | 31 October 2013 22:12:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Fire Event Could Result in a Hot Short That Could Adversely Impact Safe Shutdown Equipment | While performing a review of industry OE (operating experience) concerning unfused ammeter circuits on station batteries, it was discovered that the ammeter circuits for all of the non-1E batteries are of a similar design to that described in the OE. Also, while reviewing additional DC circuits, it was discovered that the control circuit for the Turbine Generator Emergency Lube Oil pump is unfused, protected only by the motor circuit breaker with a trip setting of 350 amps. The concern is that under the fire safe shutdown rules it is postulated that a fire in one fire area can damage these circuits and cause short circuits without protection that would overheat the cables and possibly result in secondary fires in other fire areas where the cables are routed. The secondary fires could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10CFR50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified. | 05000289/LER-2014-001 05000498/LER-2013-003 | |
| ENS 54190 | 30 July 2019 20:21:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | En Revision Imported Date 8/22/2019 | EN Revision Text: DISCOVERY OF CONDITION THAT COULD HAVE PREVENTED FULFILLMENT OF A SAFETY FUNCTION South Texas Project (STP) Unit 1 reactor head vent valve B1RCHCV0601 was declared inoperable on December 29, 2018, STP Unit 1 reactor head vent valve B1RCHCV0602 was declared inoperable on July 30, 2019. Technical Specification 3.3.3.5 requires one of two reactor head vent valves to be operable. This issue placed the plant in a 30-day Technical Specification Action. At 0741 CDT on July 31, 2019, South Texas Project Electric Generating Station (STPEGS) made a determination based on firm evidence that reactor head vent valve B1RCHCV0602 had been inoperable since June 24, 2019. This results in a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D), 'any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. The inoperable equipment is required for accident conditions and presents no danger to the health and safety of the public or the safe operation of the units. The NRC Resident Inspector has been notified.
The Event Date and Time provided in EN# 54190 should have been reported as 7/30/2019 at 1521 CDT, since this was the time of discovery of a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. The NRC Resident Inspector has been notified. Notified R4DO (Young). | ||
| ENS 54257 | 6 September 2019 02:15:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | En Revision Imported Date 11/8/2019 | EN Revision Text: CONTAINMENT PENETRATION DISCOVERED NOT ISOLATED At 2115 CDT on 9/5/2019, an inside containment test connection and inoperable outside containment isolation valve were discovered to be open for a containment air sample penetration. This resulted in the containment penetration not being isolated. The inside containment test connection was closed at 2322 CDT on 9/5/2019. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and (D) and 10 CFR 50.72(b)(3)(ii)(B). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
This event was originally reported on September 6, 2019 under 10 CFR 50.72(b)(3)(v)(C) and (D) and 10 CFR 50.72(b)(3)(ii)(B). Upon completion of the investigation of the event, it was determined that the event had insignificant safety consequences because the containment breach was disconnected from the Reactor Coolant System by a series of closed valves for the duration of the event. Additionally, the lines to the inside containment connection and the outside inoperable containment isolation valve that was found to be open as well as the main line connecting and passing through the penetration were one-inch diameter lines. Analysis determined that containment breaches that are less than a three-inch diameter do not lead to a large radiation release. The event did not place the plant in an unanalyzed condition that significantly degrades plant safety. Therefore, 10 CFR 50.72(b)(3)(ii)(B) did not apply to this event and this notification is to retract reporting under that criterion. The licensee notified the NRC Resident Inspector. Notified R4DO (Drake). | Reactor Coolant System | |
| ENS 56848 | 10 November 2023 20:13:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Essential Chiller Trains Inoperable | The following information was provided by the licensee via email: On 11/10/23 at 0642 CST, essential chiller 'B' train and cascading equipment was declared inoperable due to chill water temperature exceeding limits. At 1413 CST, essential chiller 'C' train and cascading equipment was declared inoperable due to discharge pressure exceeding limits. This condition resulted in an inoperable condition on two out of the three safety trains for the accident mitigating function including the 'B' and 'C' train high head safety injection, low head safety injection, containment spray, electrical auxiliary building HVAC, control room envelope HVAC, and essential chill water. All 'A' train equipment remained operable. This was determined to be reportable within 8 hours as required by 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Plant is in a 72 hour limiting condition for operation per technical specification 3.7.7. Restoration of 'B' train anticipated on 11/11/23 mid day. | HVAC Control Room Envelope Containment Spray | |
| ENS 56861 | 16 November 2023 21:41:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Essential Chiller Trains Inoperable | The following information was provided by the licensee via phone and email: 11/05/23, 2200 CST: Essential Chiller 'B' train and associated cascading equipment were declared INOPERABLE for planned maintenance. Unit 2 entered the Configuration Risk Management Program as required by Technical Specifications on 11/12/23 at 2200. 11/16/23, 1541: Essential Chiller 'C' train and associated cascading equipment were declared INOPERABLE due to an unexpected material condition causing the Essential Chiller to trip. The most limiting (Limiting Condition of Operability) LCO is 3.7.7, Action c. This condition resulted in the INOPERABILITY of two of the three safety trains required for the accident mitigating function including: High Head Safety Injection, Low Head Safety Injection, Containment Spray, Electrical Auxiliary Building HVAC, Control Room Envelope HVAC, Essential Chilled Water. This is an 8 hour reportable condition per 10CFR50.72(b)(3)(v)(D) because it could affect the ability to mitigate the consequences of an accident. A risk analysis was performed for the equipment INOPERABILITY and mitigating actions have been taken per site procedures. All 'A' train equipment remains operable. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The 'B' train Emergency Diesel Generator was also inoperable due to planned maintenance and continues to be inoperable. It was considered in the Configuration Risk Management Program and it was determined this condition could be maintained. LCO 3.7.7, Action c requires reactor shutdown within 72 hours. | Emergency Diesel Generator HVAC Control Room Envelope Containment Spray | |
| ENS 56944 | 24 January 2024 01:02:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Accident Mitigation - Relief Valves Inoperable | The following information was provided by the licensee via email: At 1640 CST on February 1, 2024, it was determined that a condition occurred that could have prevented the fulfillment of a safety function due to two of the four steam generator (SG) power-operated relief valves (PORVs) being simultaneously inoperable. In certain accident scenarios, more than two PORVs are needed to mitigate the consequences of an accident; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The first PORV was declared inoperable at 1025 on January 22, 2024, and the safety function is considered to have been lost when the second PORV was declared inoperable at 1902 on January 23, 2024. The safety function was restored at 2234 on January 23, 2024, when the first SG PORV was declared operable. There was no impact to unit 2. 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 plant remained in mode 3 for the duration of the condition. The causes for the two PORVs being inoperable were neither related nor systemic in nature. All SG PORVs have been restored to operation. | Steam Generator | |
| ENS 57019 | 10 March 2024 08:53:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Essential Chilled Water Trains Declared Inoperable | The following information was provided by the licensee via email: On 3/9/2024 at 2126 CST, train C essential cooling water was declared inoperable due to a through-wall leak on the discharge vent line. This would also cascade and cause train C essential chilled water to be inoperable. On 3/10/2024 at 0353 CDT, train B essential chilled water was declared inoperable due to chilled water outlet temperature greater than 52 degrees F following startup of essential chiller 12B. Chilled water outlet temperature was adjusted to less than 52 degrees F at 0440 CDT, and train B essential chilled water was declared operable. This condition resulted in the inoperability of two of the three safety trains required for the accident mitigating functions including: high head safety injection, low head safety injection, containment spray, electrical auxiliary building HVAC, control room envelope HVAC, and essential chilled water. This is an 8 hour reportable condition per 10CFR50.72(b)(3)(v)(D) because it could affect the ability to mitigate the consequences of an accident. The licensee notified the NRC Resident Inspector. | HVAC Control Room Envelope Containment Spray | |
| ENS 57146 | 26 May 2024 12:20:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Two of Three Essential Chilled Water Trains Declared Inoperable | The following information was provided by the licensee by phone and email: At 0210 CDT 5/24/24, essential chiller A train and cascading equipment was declared inoperable for maintenance to correct a temperature control malfunction. At 0720 CDT 5/26/24, essential cooling water B train and cascading equipment (including B train essential chiller) was declared inoperable due to a through wall leak discovered on the essential cooling water return header temperature element thermal well. This condition resulted in an inoperable condition on two out of three safety trains for the accident mitigating function, including the train A and train B high head safety injection, low head safety injection, containment spray, electrical auxiliary building heating ventilation and air conditioning (HVAC), and essential chilled water. All C train safety related equipment remains operable. This was determined to be reportable within 8 hours as required by 10CFR50.72(b)(3)(v)(D). NRC Resident Inspector has been notified. | HVAC Containment Spray |