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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56912 | 3 January 2024 17:57:00 | 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)(A), Seriously Degraded | Pressure Boundary Degraded / Both Trains of High Pressure Safety Injection Inoperable | At 1257 EST on January 3, 2024, it was determined that a class 1 system barrier had a through wall flaw with leakage. The leakage renders both trains of high pressure safety injection inoperable. The unit is being cooled down to cold shutdown to comply with technical specifications. This event is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(A) and 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 Officer Report Guidance: At the time of the discovery, the unit was shutdown in mode 3. The unit was experiencing signs of reactor coolant system leakage and a shutdown was initiated in order to search for possible sources. The unit is currently cooling down and proceeding to mode 5, where the safety function is not required.
The purpose of this notification update is to retract a portion of a previous report, made on 1/03/2024 at 1257 EST (EN 56912). Notification of the event to the NRC was initially made as a result of declaring both trains of unit 2 high pressure safety injection system inoperable due to reactor coolant barrier through wall leak on the vent line for the 2A2 safety injection tank. Subsequent to the initial report, Florida Power and Light has concluded that the through wall leak rate was insignificant, and therefore the safety injection system's safety-related function was maintained. Therefore, this portion of the event is not considered a safety system functional failure and is not reportable to the NRC pursuant 10 CFR 50.72(3)(v)(D). This update does not affect the original 10 CFR 50.72(b)(3)(ii)(A) report for the degraded condition related to the reactor coolant barrier through wall leak. The NRC Resident Inspector has been notified. Notified R2DO (Miller). | Reactor Coolant System | |
| ENS 56893 | 16 December 2023 01:45:00 | 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)(A), Seriously Degraded | Reactor Coolant System Leak | At 2045 EST on December 15, 2023, it was determined that the reactor coolant system barrier had a through wall flaw with leakage. The leakage is minor in nature and unquantifiable. The leakage is coming from the welded connection of a vent valve for safety injection tank 2A2 outlet valve rendering both trains of high-pressure safety injection inoperable. The unit is being cooled down to cold shutdown to comply with technical specifications. This event is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(A) and 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: The unit was heating up after a maintenance outage. The leak was discovered during mode 3 walkdown.
The purpose of this notification update is to retract a portion of a previous report, made on 12/16/2023 at 0404 EST (EN 56893). Notification of the event to the NRC was initially made as a result of declaring both trains of unit 2 high pressure safety injection system inoperable due to reactor coolant barrier through wall leak on the vent line for the 2A2 safety injection tank. Subsequent to the initial report, Florida Power and Light has concluded that the through wall leak rate was insignificant, and therefore the safety injection system's safety-related function was maintained. Therefore, this portion of the event is not considered a safety system functional failure and is not reportable to the NRC pursuant 10 CFR 50.72(3)(v)(D). This update does not affect the original 10 CFR 50.72(b)(3)(ii)(A) report for the degraded condition related to the reactor coolant barrier through wall leak. The NRC Resident Inspector has been notified. Notified R2DO (Miller). | Reactor Coolant System | |
| ENS 56135 | 30 September 2022 20:08:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Safety System Inoperability | The following information was provided by the licensee via email: At 1608 (EDT) on September 30, 2022, it was discovered that both trains of the chemical volume and control system were simultaneously inoperable due to an unisolable piping flaw inside containment detected during plant pressurization in preparation for startup following a refueling outage. St. Lucie Unit 2 was not affected and remains at 100 percent power. This event is being reported pursuant to 10CFR50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous report made on 09/30/2022 at 1713 EDT (EN 56135). Notification of the event to the NRC was initially made as a result of declaring both trains of U1 Chemical and Volume Control System inoperable due to a piping flaw detected during plant pressurization in preparation for startup following a refueling outage. Subsequent to the initial report, FPL (Florida Power and Light) has concluded that the flaw identified in line 2"-CH-109 did not exceed (with sufficient margin) the allowable axial flaw size utilizing the ASME Code Case N-869 methodology, and the Chemical and Volume Control System was operable but degraded for the period of concern. Therefore, this event is not considered a Safety System Functional Failure and is not reportable to the NRC as a Licensee Event Report (LER) per 10 CFR 50.73. The NRC Senior Resident Inspector has been notified. Notified R2DO (Miller). | ||
| ENS 56128 | 26 September 2022 21:41:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | Safety System Inoperability | The following information was provided by the licensee via email: At 1741 EDT on September 26, 2022, it was discovered that both trains of the chemical volume and control system were simultaneously inoperable due to an unisolable piping flaw detected during plant pressurization in preparation for startup following refueling outage. St. Lucie Unit 2 was not affected and remains at 100 percent power. This event is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: The purpose of this notification is to retract a previous report made on 09/26/2022 at 2239 EDT (EN 56128). Notification of the event to the NRC was initially made as a result of declaring both trains of U1 Chemical and Volume Control System inoperable due to a piping flaw detected during plant pressurization in preparation for startup following a refueling outage. Subsequent to the initial report, FPL (Florida Power and Light) has concluded that the flaw identified in line 2"-CH(1)104 did not exceed (with sufficient margin) the allowable axial flaw size utilizing the ASME Code Case N-869 methodology, and the Chemical and Volume Control System was Operable but degraded for the period of concern. Therefore, this event is not considered a Safety System Functional Failure and is not reportable to the NRC as a Licensee Event Report (LER) per 10 CFR 50.73. The NRC Senior Resident Inspector has been notified. Notified R2DO (Miller). | ||
| ENS 53598 | 11 September 2018 04:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Weld Defect on Pressurizer Interface | At 2200 EDT on September 11, 2018, during inspections as part of the pressurizer bare metal inspection program, an apparent weld defect was identified on Class 1 pressurizer level piping at the piping to pressurizer interface. The reactor is currently defueled in a refueling outage. Appropriate repairs will be made during the refueling outage. NRC Resident Inspector has been notified. | ||
| ENS 52523 | 31 January 2017 17:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant Pressure Boundary Piping Defect | At 1200 (EST) on January 31, 2017, during investigation of potential 1B2 Reactor Coolant Pump seal degradation, a through wall defect was identified on Class 1 piping servicing the Lower Seal Heat Exchanger, which is part of the Reactor Coolant Pressure Boundary (RCPB). The reactor is presently in Mode 3 with decay heat being removed by the atmospheric steam dump valves. The plant is being maneuvered to Mode 5 to affect appropriate repairs. The licensee has notified the NRC Resident Inspector. | ||
| ENS 50310 | 25 July 2014 15:29:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Technical Specification Required Shutdown Due to a Safety Injection Valve Defect | On July 25th, 2014 at 1129 EDT, during investigation of a small leak inside Saint Lucie Unit 2 containment, a defect was identified on a valve (V3811, Safety Injection Tank 281 Outlet Vent Valve). This valve is within the Q-Group A (ASME Class 1 equivalent) boundary of the safety injection line. The leak is currently isolated from the reactor coolant system by a Technical Specification credited isolation check valve. The affected Safety Injection Tank has been declared inoperable and the unit has entered 24 hour shutdown (Technical Specification) action statement 3.5.1.b. Unit shutdown is currently being planned for repair of the flaw. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(A). The licensee has notified the NRC Resident Inspector.
On July 26, 2014 at 0002 EDT, Saint Lucie Unit 2 commenced shutdown to repair the identified leak. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(A) and 10 CFR 50.72(b)(2)(i). The NRC Resident Inspector has been notified. Unit 2 current power is 88% at 0105 EDT and expect to trip the unit at approximately 0400 EDT. Notified the R2DO (Musser). | Reactor Coolant System | |
| ENS 45200 | 13 July 2009 11:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leakage Identified | On 07/13/2009 at 00:38 EDT a planned manual reactor trip was initiated on Unit 2 based on increased, unidentified RCS leakage. The RCS leakage had increased over a four day period. The downpower was initiated at 21:32 EDT from 100% power and proceeded in a controlled manner to 25% power. The reactor was manually tripped at 00:38 EDT, as planned. Following the reactor trip, EOP-1, Standard Post Trip Actions, and EOP-2, Reactor Trip Recovery procedures were completed and the unit was stabilized in Mode 3. Decay heat was removed by the Steam Bypass Control System (SBCS) to the condenser. The Main Feedwater Pump unexpectedly tripped on low suction pressure just prior to the reactor trip, but was recovered prior to levels reaching the Auxiliary Feedwater Actuation (AFAS) setpoint. There were no further major equipment failures identified post trip. On 7/13/2009 at 07:00 EDT a Containment walkdown confirmed the leak was caused by a J-joint weld defect on the 2B2 RCP Seal injection line. The leak is classified as reactor coolant pressure boundary leakage. The plant is in the process of cooling down to enter Mode 5 as required by Technical Specification 3.4.6.2.a. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(3)(ii)(A). Additional investigation is being performed to determine the cause of the leak. The initial leak rate was less than 0.2 gpm. The repair will require taking the plant to a reduced inventory. The licensee notified the NRC Resident Inspector | Feedwater Auxiliary Feedwater Steam Bypass Control System | |
| ENS 43941 | 29 January 2008 10:31:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Manual Reactor Trip to Repair Rcp "2B1" Seal Package | St Lucie Plant, Unit 2, manually tripped the reactor at 0531 hrs. EST, as part of a controlled reactor plant shutdown, in response to an RCS leak rate of 0.28 gpm, indicated to be associated with the 2B1 Reactor Coolant Pump seal package. All systems functioned as expected, satisfactorily. All CEAs fully inserted. Decay Heat removal is with Steam Generators and Main Feedwater. This report is made pursuant to 10 CFR 50.72(b)(2)(iv)(B). Unit 2 will cooldown to Mode 5 for repairs. The licensee informed the NRC Resident Inspector.
St. Lucie Plant, Unit 2, reported their manual reactor trip of 0531 hrs. EST, at 0630 hrs. EST, for an unidentified leak associated with the 2B1 RCP seal package (EN 43941). Subsequent investigation, reported to the STA (Shift Technical Advisor) at 1015 hrs. EST, determined that this leak constitutes an RCS Pressure Boundary Leak. This leak is at a pipe-to-flange weld on the outboard side of the first flanged coupling of the 2B1 RCP Upper Cavity Seal pressure sensing line. Investigation is in progress to determine the detailed configuration at the leak and root cause. There were NO ESFAS signals or actuations. Decay Heat Removal continues on Steam Generators, with MFW (Main Feedwater) and SBCS (Steam Bypass Control System); Off-Site power continues (to be) available and stable; Unit 1 at 100% power with operations and conditions normal. This report is made pursuant to 10 CFR 50.72(b)(3)(ii)(A). The licensee informed the NRC Resident Inspector of this update. Notified R2DO (Bonser). | Steam Generator Feedwater Decay Heat Removal | |
| ENS 43863 | 21 December 2007 19:45:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Rcs Pressure Boundary Leakage on Rcp Seal Injection Line | With unit shut down in Mode 3 with all CEAs inserted preparing for startup, RCS Pressure Boundary leakage was identified. Leakage appears to be from a weld to Reactor Coolant Pump Seal Injection Line. Leakage is much less than one gallon per minute. Implementing Technical Specification LCO 3.4.6.2 Reactor Coolant System Operational Leakage Action (a) to place the unit in Cold Shutdown within the following 30 hours. Commencing plant cooldown for repair. The licensee notified the NRC Resident Inspector. | Reactor Coolant System | |
| ENS 43580 | 19 August 2007 21:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Plant Shutdown Due to Unisolable Reactor Coolant System Leakage | On 08/18/2007 at 16:03 EDT a Planned Manual Reactor Trip was initiated on Unit 2 based on increased unidentified RCS leakage. The RCS leakage had increased over a two day period and surpassed the hard limit of 0.28 gpm established by a plant Operational Decision Making (ODM) plan. The downpower was initiated on 08/18/07 at 10:35 EDT from 100% power and proceeded in a controlled manner to 25% power. The Feedwater Control System was maintained in the automatic mode. Station Power was transferred by procedure from the Auxiliary to the Startup Transformers. The reactor was manually tripped at 16:03 EDT. Following the reactor trip, EOP-1, Standard Post Trip Actions, and EOP-2, Reactor Trip Recovery procedures were completed without contingencies and the unit was stabilized in Mode 3. Decay heat was removed by the Steam Bypass Control System (SBCS) to the condenser. The Steam Generators were fed by the MFW System and levels remained above Auxiliary Feedwater Actuation (AFAS) setpoint. There were no significant equipment failures identified post trip. The plant was cooled down and entered Mode 5 on 8/19/07 at 15:09 EDT. On 8/19/07 at 17:00 EDT while in Mode 5, a Containment walkdown confirmed the 2B1 RCP Seal injection line to be the only active RCS leakage source. The leak is unisolable from the RCS and is located at or adjacent to a socket weld on the 3/4" Class 1 Schedule 160 seal injection line. Removal of obstructions and insulation must occur before the specific details of the leak can be identified. This non-emergency notification is being made pursuant to 10 CFR 50.72(b)(3)(ii)(A). Additional investigation is being perform to determine the detailed configuration of the leak and the root cause. The licensee notified the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater Steam Bypass Control System | |
| ENS 43410 | 6 June 2007 17:10:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Containment Isolation Valves Found Misaligned | On 6/06/07 at 13:10 EDT while performing a surveillance to verify the position of administratively locked valves for Containment and Shield Building Integrity, two manual Containment Isolation valves were found Locked Open, versus the required Locked Closed position. The valves were restored to the Locked Closed position at 13:30 EDT of the same day. The subject valves, V18794 and V18796, provide Service Air (a non-safety class system) to the Containment during Mode 5 and 6 conditions through Containment Penetration #8. The subject valves are required to be returned to Locked Closed prior to entering Mode 4 in accordance with the Plant Technical Specifications 3.6.1.1 for Containment Integrity and plant procedures. The valves are to be verified in the Locked Closed position once every 31 days per Administrative Procedure 1-0010123 Appendix E, which was last performed on 5/09/07. Unit 1 exited Mode 5 and entered Mode 4 on 5/20/07 following a refueling outage and is currently in Mode 1 at 100% power. It is assumed the valves have been incorrectly positioned since entering Mode 4. The Service Air System is intact and pressurized at approximately 110 psig. There is no open path from Containment to the outside atmosphere. This event is being reported pursuant to 10CFR50.72(b)(3)(ii)(A), based on principal safety barriers being seriously degraded. The licensee notified the NRC Resident Inspector. | Shield Building | |
| ENS 41401 | 10 February 2005 21:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Rcs Leakage Identified During Plant Startup Walkdowns | On 2/10/2005, St. Lucie Unit 2 was in Mode 3 returning from refueling outage SL2-15. At 0605 during plant startup walkdowns, 2 leaks were discovered on the instrument line to the 2B1 safety injection header. Closing the instrument root isolation valve to PT-3339 isolated one leak. The second leak was under the insulation and later upon evaluation it was concluded to be ASME section Xl RCS pressure boundary leakage based on its location between the reactor coolant system loop check valves. The leakage source is from the 2B1 safety injection tank and not directly from the reactor coolant system because the first reactor coolant system loop check valve is maintaining pressure isolation. As a result this is not Technical Specification pressure boundary leakage. The pressure instrument root is downstream of the Safety Injection Tank (SIT) isolation valve and between the two reactor coolant system loop check valves. Isolating the SIT from the header isolated second leak. An engineering response team was formed and repairs are being planned. RCS heat removal is being accomplished via the Steam Generators and the Atmospheric Dump Valves. Feedwater is being supplied to the Steam Generator by the Auxiliary Feedwater System. Plant electrical loads are supplied from offsite sources through the Startup Transformers. The licensee informed the NRC Resident Inspector. | Steam Generator Reactor Coolant System Feedwater Auxiliary Feedwater |