Semantic search
Start date | Report date | Site | Reporting criterion | System | Event description | |
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ENS 56974 | 17 February 2024 13:37:00 | Brunswick | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Core Isolation Cooling Primary containment | The following information was provided by the licensee via email and phone call: At 0837 EST, on 02/17/2024, during a refueling outage at 0 percent power while performing local leak rate testing (LLRT) on the reactor core isolation cooling (RCIC) isolation valves, which is part of the containment boundary, it was determined that the Unit 1 primary containment leakage rate did not meet 10 CFR 50 Appendix J requirements specified in Technical Specification 5.5.12. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56959 | 11 February 2024 15:11:00 | Hatch | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Feedwater Primary containment | The following information was provided by the licensee via email: At 1011 EST on 02/11/2024, during a refueling outage at 0 percent power, while performing local leakage rate testing (LLRT) of the feedwater check valves (part of the containment boundary), it was determined that the Unit 1 primary containment leakage rate did not meet 10 CFR 50 Appendix J requirements specified in Technical Specification 5.5.12. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56912 | 3 January 2024 17:57:00 | Saint Lucie | 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 | 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). | |
ENS 56893 | 16 December 2023 01:45:00 | Saint Lucie | 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 | 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). | |
ENS 56834 | 5 November 2023 16:33:00 | Waterford | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Steam Generator | The following information was provided by the licensee via email: At 1033 CST on November 5, 2023, while in a refueling outage, it was determined that Waterford Steam Electric Station, Unit 3, did not meet the performance criteria for steam generator structural integrity in accordance with Technical Specification 6.5.9.b.1, Steam Generator Program, due to two tube failures in the number 1 steam generator. The condition was identified during performance of in-situ pressure testing. The affected tubes will be plugged. The plant is currently stable with all fuel in the spent fuel pool. Decay heat is being removed by normal spent fuel cooling system operation. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A) as a degraded condition. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56785 | 9 October 2023 22:07:00 | Palo Verde | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | The following information was provided by the licensee via email: On October 9, 2023, during the Palo Verde Nuclear Generating Station Unit 1 refueling outage, while performing a small nozzle inspection in support of boric acid walkdowns, boric acid leakage was found on the area of the weld of a pressurizer thermowell. At 1507 MST, non-destructive examination of the weld indicated leakage through the reactor coolant pressure boundary. The exam result constitutes welding or material defects in the primary coolant system that are unacceptable under ASME Section XI. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56774 | 3 October 2023 15:54:00 | North Anna | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | The following information was provided by the licensee via email: At 1154 EDT on 10/03/23, investigation into a boric acid indication was determined to be through a leak on a weld-o-let upstream of a pressurizer level transmitter isolation valve. Unit 2 is currently in MODE 6 with reactor coolant system (RCS) operational leakage limits not applicable. The leak is not quantifiable as it only consists of a small amount of dry boric acid at the location. The failure constitutes welding or material defects in the primary coolant system that are unacceptable under ASME Section XI. Therefore, this is a degraded condition reportable under 10 CFR 50.72(b)(3)(ii)(A). This condition does not affect the health and safety of the public or station employees. The Resident Inspector was notified. | |
ENS 56685 | 20 August 2023 20:00:00 | Fermi | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | The following information was provided by the licensee via email: On 8/20/2023 at 1600 EDT, during plant walkdowns in the drywell while in mode 3 to identify a cause of increasing unidentified leakage rate, reactor coolant system pressure boundary leakage (approximately 2 gpm) was identified on the reactor recirculation sample line between the reactor recirculation sample line inboard isolation valve (B3100F019) and where the sample line taps off the B reactor recirculation jet pump riser. This requires entry into technical specification 3.4.4 condition C, identification of pressure boundary leakage with a required action to be in mode 3 in 12 hours and mode 4 in 36 hours. At 1630 EDT, a technical specification required shutdown to mode 4, cold shutdown, was initiated. A press release by DTE is anticipated. This event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i), a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi), and an eight-hour, non-emergency notification 10 CFR 50.72(b)(3)(ii)(A) for the degraded condition of the pressure boundary. Investigation into the cause of the reactor coolant system pressure boundary leakage is still ongoing. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. | |
ENS 53456 | 15 July 2023 10:39:00 | Arkansas Nuclear | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | On June 12, 2018, at 1500 CDT, a Reactor Coolant System (RCS) Pressure Boundary leak was identified during a Mode 3, hot shutdown walkdown on a High Pressure Injection Line (HPI) to Reactor Coolant Pump (P32C) drain line weld near MU-1066A HPI Line Drain Valve and MU-1066B HPI Line Drain Valve. The 3/4 inch drain line containing drain valves MU-1066A and MU-1066B on the 'C' HPI header (CCA-5 pipe class) has a through-wall defect on the pipe stub or welds between the sockolet and valve MU-1066A. The leak location is in the ASME Class I RCS Pressure Boundary. The hot shutdown walkdown was being performed as part of a planned outage to investigate excessive Reactor Building Sump inleakage. Total unidentified RCS leakage prior to the investigation was determined to be at 0.165 gpm. After the initial investigation of the leakage, the following Tech Specs (TS) were determined be applicable: TS 3.4.5 - RCS Loops Mode 3, TS 3.4.13 - RCS Leakage, TS 3.5.2 - ECCS. Unit 1 is currently in Mode 3 and in progress of an RCS cooldown to comply with Tech Spec requirements. The licensee notified the NRC Resident Inspector. | |
ENS 56494 | 30 April 2023 05:00:00 | Hope Creek | 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 | Primary containment | The following information was provided by the licensee via email: At 0100 EDT on 04/30/23, it was determined that the primary containment integrity did not meet (Technical Specification) TS 4.6.1.1.d requirement, suppression chamber in compliance with TS 3.6.2.1 due to the inability to establish test conditions for the bypass leakage test in accordance with TS 4.6.2.1.f. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D) & 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56485 | 22 April 2023 23:42:00 | Beaver Valley | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | The following information was provided by the licensee via email: At 1942 EDT on April 22, 2023, during the Beaver Valley Power Station, Unit No. 2 refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per ASME Code Section XI. The reactor vessel vent line penetration will require repair prior to returning the vessel head to service. The indication was not through-wall as there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified | |
ENS 56474 | 15 April 2023 15:12:00 | Turkey Point | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | The following information was provided by the licensee via email: At 1112 EDT on 4/15/23, it was determined that the (reactor coolant system) RCS pressure boundary does not meet ASME Section XI, Table IWB-341 0-1, `Acceptable Standards,' due to through wall leak of the flux mapper seal table guide tube H-6. Corrective actions have been scheduled. `This event is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(A). A follow-up review of the condition revealed that 10 CFR 50.72 notification was applicable within 8 hours of the time of discovery on 04/15/23. The NRC Resident Inspector has been notified. | ||
ENS 56434 | 26 March 2023 20:03:00 | Susquehanna | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Secondary containment | The following information was provided by the licensee via email: On 03/26/2023 at 1603 EDT, while performing Appendix J local leak rate testing, it was determined that the Secondary Containment Bypass Leakage (SCBL) limit had been exceeded for Unit 2. During performance of the leak rate test, SE-259-027 for X-9B penetration, it was determined that the combined SCBL limit of 15 standard cubic feet per hour for the as-found minimum pathway was exceeded, as specified in Technical Specification, Surveillance Requirement 3.6.1.3.11. This event is being reported pursuant to 10CFR50.72(b)(3)(ii). The Resident Inspector has been notified. | |
ENS 56421 | 18 March 2023 18:10:00 | Nine Mile Point | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | The following information was provided by the licensee via email: On 3/18/2023 at 1410 EDT, with Nine Mile Point Nuclear Station Unit 1 in a planned refueling outage, the main control room was notified of the results of an automated examination of a dissimilar metal weld on reactor penetration N2E. The results indicate a defect present which cannot be found acceptable under ASME Section XI, IWB-3600. This 8-hour non-emergency report is being made based upon requirements of 10CFR50.72(b)(3)(ii)(A) which states, `The licensee shall notify the NRC ... of the occurrence of ... any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The NRC Senior Resident was informed. A repair plan is being developed. | ||
ENS 56411 | 15 March 2023 03:57:00 | Browns Ferry | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System Main Steam | The following information was provided by the licensee via email: At 2257 (CDT) on 3/14/2023 during the 2R22 refueling outage on Browns Ferry Nuclear Plant Unit 2, it was determined there was RCS boundary leakage from five of eight sensing lines that pass through containment penetrations X-30 and X-34 that did not meet the requirements of Section XI, of the ASME Boiler and Pressure Vessel Code. The condition will be resolved prior to plant startup. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. 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 Event Notification, EN 56411 reported on 3/14/23. Following the initial notification, further analysis of the condition was performed. It was determined that the leaking pipe weld was ASME Section XI Code Class 2 piping which falls under the requirements of ASME Section XI Subsection IWC and not Subsection IWB. Therefore, this condition does not represent a serious degradation of the nuclear power plant, including its principle safety barriers. Based upon the above, the leaks identified on the ASME Section XI Code Class 2 equivalent Main Steam sense lines are not reportable under 10 CFR 50.72(b)(3)(ii). Therefore, the NRC non-emergency 10 CFR 50.72(b)(3)(ii) report was not required and the NRC report 56411 can be retracted and no Licensee Event Report under 10 CFR 50.73(a)(2)(ii) is required to be submitted. Notified R2DO (Miller) | |
ENS 56371 | 18 February 2023 10:39:00 | Browns Ferry | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | The following information was provided by the licensee via email: On February 17, 2023 during the planned U2R22 outage on Browns Ferry Nuclear Plant Unit 2, personnel entered the Unit 2 drywell for leak identification. Personnel discovered a cracked weld on the 2A recirculation pump discharge isolation valve drain line. At 0439 CST on February 18, 2023, following engineering evaluation, this drain line was determined to be ASME Code Class 1 piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. | ||
ENS 56342 | 7 February 2023 22:38:00 | Hatch | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary containment | The following information was provided by the licensee via email: At 1738 EST on 02/07/2023, while in mode 5 at 0 percent power, it was determined during local leak rate testing (LLRT) that the primary containment leakage rate exceeded the allowable limit defined in 10 CFR 50, Appendix J, 'Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.' Both primary containment isolation valves in a penetration failed LLRT requirements which represents a failure to maintain primary containment integrity. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56257 | 3 December 2022 16:00:00 | Browns Ferry | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Shutdown Cooling Residual Heat Removal | The following information was provided by the licensee via email: On 12/2/2022 at 2330 (CST) during the planned F311 outage on Browns Ferry Nuclear Plant Unit 3, personnel entered the Unit 3 drywell for leak identification. Personnel discovered a through-wall piping leak on a 0.75 inch test line between the two test line isolation valves. This 0.75 inch test line is located on the residual heat removal (RHR) loop 1 shutdown cooling and RHR return line to the reactor vessel. On 12/3/2022 at 1000 CST, Engineering determined this location is classified as ASME Code Class 1 piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. | |
ENS 56221 | 14 November 2022 11:30:00 | Oconee | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | The following information was provided by the licensee via email: During a scheduled refueling outage, a walkdown inside containment discovered a small amount of boron on the 1B2 Reactor Coolant Pump (RCP) lower bearing temperature instrument. At 0730 EST on November 14, 2022, with Unit 1 in Mode 6, disassembly of the instrument indicated the source of the boron was from a leak in the thermowell. The thermowell is considered part of the reactor coolant system pressure boundary and as such the condition is reportable. Repairs for the condition are in progress. This event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56197 | 1 November 2022 18:33:00 | Oconee | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | The following information was provided by the licensee via email: At 1433 EDT on November 1, 2022, it was determined that a single relevant indication in the RCS pressure boundary did not meet the acceptance criteria under ASME, Section XI IWB-3514-2. 'Allowable Planar Flaws.' The condition will be resolved prior to plant startup. This event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56176 | 23 October 2022 15:30:00 | Diablo Canyon | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | The following information was provided by the licensee via email: At 0830 PDT on 10/23/2022, during routine outage inspections on Unit 2, it was determined that the RCS Pressure Boundary did not meet ASME Section XI acceptance criteria on a 2-inch vacuum refill connection line. This event is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 56165 | 15 October 2022 15:59:00 | Catawba | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | The following information was provided by the licensee via email: On 10/15/2022 at 1159 (EDT), during the Catawba Nuclear Station Unit 2 refueling outage, it was determined that the results of a planned surface examination Liquid Penetrant test (PT) performed on a previous overlay repair on nozzle number 74 of the reactor vessel closure head (RVCH) did not meet applicable acceptance standards. The examination was being performed to meet the requirements of Relief Request RA-21-0144, 'Proposed Alternative to Use Reactor Vessel Head Penetration Embedded Flaw Repair for Life of Plant'. The penetration required repairs for the discovered indications. The repairs have been completed in accordance with the ASME Code of Record prior to returning the vessel head to service. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The original indication that led to the overlay repair was discovered in April 2021, during ultrasonic testing and reported to the NRC and assigned EN55201. | ||
ENS 56153 | 7 October 2022 06:19:00 | Braidwood | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Control Rod | The following information was provided by the licensee via fax: Control Rod Drive Mechanism (CRDM) penetration 69 degraded. At 0119 (CDT) on October 7, 2022, it was determined that the CRDM penetration 69 was degraded because examination identified unacceptable indications in accordance with ASME Code Case N-729-6. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The notification is being corrected to state: At 0119 (CDT) on October 7, 2022, it was determined that the Control Rod Drive Mechanism (CRDM) penetration 69 was degraded because liquid penetrant testing, performed on the seal weld, identified unacceptable indications in accordance with ASME Section III and NRC approved licensee relief request for a previously performed embedded flaw repair. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. Notified R3DO (Ruiz). | |
ENS 55857 | 23 April 2022 13:54:00 | Byron | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | The following information was provided by the licensee via email: At 0854 (CDT) on April 23, 2022, while performing volumetric inspections required by ASME Code Case N-729-6, a rejectable indication on Reactor Vessel Head Penetration 75 Core Exit Thermocouple (CETC) was identified. The indication is located inboard of the J-groove weld and is OD-initiated (outer diameter - initiated). This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). 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 repair is scheduled during the refueling outage.
The following information was provided by the licensee via telephone and email: The rejectable indication on Reactor Vessel Head Penetration 75 Core Exit Thermocouple (CETC) initiated on the outside diameter (OD) of the nozzle in an area that was not surface stress mitigated (peened). The indication was found to be acceptable for continued operation under CFR and ASME requirements and will not be repaired during this outage. The licensee notified the NRC Resident Inspector. Notified R3DO (Ziolkowski). | ||
ENS 55706 | 16 January 2022 05:20:00 | Browns Ferry | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Shutdown Cooling | The following information was provided by the licensee via email: On 1/15/2022 at 2320 (EST) during the planned F108 outage on Browns Ferry Nuclear Plant Unit 1, personnel entered the Unit 1 Drywell for leak identification. Personnel discovered a through-wall piping leak on a 3/4 inch test line upstream of the test valve. This 3/4 inch test line is located on the RHR Shutdown Cooling & RHR Return Line to the reactor vessel and is classified as ASME Code Class 1 Piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC resident has been notified. | |
ENS 55574 | 12 November 2021 21:05:00 | Turkey Point | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | At 1605 EST on 11/12/21, it was determined that the RCS Pressure Boundary does not meet ASME Section XI, Table IWB-341 0-1, 'Acceptable Standards' due to a through wall leak of the Core Exit Thermocouple Nozzle Assembly. Measures have been taken to establish Mode 5 for corrective actions. This event is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: On 11/12/2021 EN 55574 reported possible evidence of pressure boundary through-wall leakage observed on a Core Exit Thermocouple (CET) tube. On 3/10/2022, based on laboratory analysis of the affected CET tube section, FPL Engineering determined that there was no pressure boundary through-wall leakage associated with this event. Analysis identified that the leakage likely originated from an adjacent threaded compression fitting on a tubing joint. This condition complies with ASME Section XI requirements and is therefore not reportable. This follow-up NRCOC notification is a retraction of EN 55574. The NRC Resident Inspector has been notified. Notified R2DO (Miller). | |
ENS 55560 | 3 November 2021 19:31:00 | Columbia | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary containment Residual Heat Removal | On 11/03/2021 Columbia Generating Station concluded the results for refueling outage 24 (R24) and 25 (R25) Local Leak Rate Testing as-found data was incorrect. At 1231 PDT on November 3rd, 2021 Columbia Generating Station determined the local leak rate tests (LLRT) for the X- 25B containment penetration did not meet Technical Specification requirements for LLRT acceptance criteria. The incorrect LLRT data identified for residual heat removal (RHR) B Suppression Pool Spray containment isolation valve (RHR-V-27B) was from the previous two refueling outages (R24 on 5/22/2019 & R25 on 6/512021) at which time primary containment was not required to be operable. The corrected leakage assigned to the X-25 penetration also resulted in total Type B and C leakage summation exceeding the maximum allowable leakage rate for the primary containment (1.0 La) for R24 and exceeding 0.6La in R25. The valve was flushed and retested satisfactory prior to entering the mode of applicability. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 55540 | 22 October 2021 05:04:00 | Beaver Valley | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | At 0104 EDT on October 22, 2021, during the Beaver Valley Power Station, Unit 2 refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that two penetrations could not be dispositioned as acceptable per ASME Code Section XI. Penetrations 28 and 40 will require repair prior to returning the vessel head to service. The indications were not through wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-6 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 55506 | 4 October 2021 19:33:00 | Arkansas Nuclear | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | At 1433 CDT, on October 4, 2021, Arkansas Nuclear One, Unit 2 (ANO-2) completed the analysis related to an indication revealed on head penetration 46 during Reactor Vessel Closure Head inspections. It was determined the indication is not acceptable under ASME code requirements. The indication displays characteristics consistent with primary water stress corrosion cracking. No leak path signal was identified during ultrasonic testing. The plant was in cold shutdown at 0 percent power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. This is the only indication that is currently present, however, if additional indications are found, they will also be repaired prior to the plant startup. The NRC Senior Resident Inspector has been notified. | ||
ENS 55457 | 12 September 2021 21:28:00 | North Anna | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | On September 12, 2021, at 1728 EDT, with Unit 1 in Mode 5 (Cold Shutdown) while performing inspections of the North Anna Power Station Unit 1 reactor vessel head flange area, a weld leak was identified on the reactor vessel flange leak-off line that connects to the flange between the inner and outer head o-rings. Entered TRM 3.4.6 Condition B for ASME Code Class 1,2, and 3 components. With known leakage past the inner head o-ring, this condition is reported since the fault in the tubing is considered pressure boundary (Reactor Coolant System) leakage. This event is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(A) for any event or condition that results in the condition of the nuclear power plant, including its principle safety barriers, being seriously degraded. The NRC Resident has been notified.
The condition identified in EN 55457, pursuant to 10 CFR 50.72 (b)(3)(ii)(A) has been evaluated, and has been determined not to be Reactor Coolant System (RCS) pressure boundary leakage. As such, the 8-hour report is being retracted, as it is not an event or condition that results in, 'the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The leakage was subsequently determined to be in a tubing connection downstream of the reactor vessel inner O-ring. Leakage past a seal or gasket is not considered to be pressure boundary leakage, as defined by Technical Specifications. The NRC Resident Inspector has been notified. Notified R2DO (Miller) | |
ENS 55275 | 25 May 2021 21:51:00 | Catawba | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary containment | At 1751 EDT on May 25, 2021, it was determined the local leak rate test (LLRT) for the 2EMF-IN containment penetration did not meet 10 CFR 50 Appendix J requirements for both the inboard and outboard containment isolation valves (2MISV5230 and 2MISV5231). The LLRT was performed during the previous refueling outage at which time primary containment was not required to be operable. The leakage assigned to the penetration also resulted in total leakage exceeding the allowed overall leakage. The valves were repaired and retested satisfactory prior to entering the mode of applicability, This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A), There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 55201 | 21 April 2021 02:30:00 | Catawba | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | During the performance of reactor vessel closure head (RVCH) examinations, at 2230 EDT on April 20, 2021, it was determined that the Unit 2 RVCH penetration nozzle number 74 did not meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME code case N-729-6 . All other RVCH penetration examinations have been completed per 10CFR50.55a(g)(6)(ii)(D) and ASME code case N-729-6 with no other relevant indications identified. The condition of the Unit 2 reactor vessel head penetration nozzle number 74 will be resolved prior to re-installation of the Unit 2 RVCH. This event is being reported as an eight-hour, non-emergency notification per 10CFR50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 55122 | 4 March 2021 08:23:00 | Calvert Cliffs | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | At time 0323 (EST) on March 04, 2021, it was determined that the Reactor Coolant System (RCS) pressure boundary did not meet the acceptance criteria under ASME, Section XI IWB-3600, "Analytical Evaluation of Flaws." This condition will be resolved prior to plant start up. This event is being reported as an eight hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident has been notified. | |
ENS 54994 | 11 November 2020 18:11:00 | Watts Bar | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Steam Generator Reactor Coolant System | At 1311 EST on November 11, 2020, it was determined, after evaluation of the Watts Bar Nuclear Plant (WBN) Unit 2 Steam Generator (SG) tube eddy current test data collected during the on-going refueling outage, that the WBN Unit 2 Reactor Coolant System pressure boundary did not meet the performance criteria for SG tube structural integrity. Specifically, SG number 3 failed the condition monitoring assessment for conditional burst probability. WBN has completed tube plugging and additional corrective actions are in progress. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 54971 | 29 October 2020 14:30:00 | Peach Bottom | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System Reactor Pressure Vessel | At 1030 EDT on Thursday, October 29, 2020, during the performance of Peach Bottom Atomic Power Station leakage testing of the reactor pressure vessel and associated piping, a through-wall leak (non-isolable) was identified on an instrument line connected to the N16A nozzle. The reactor will be maintained shutdown until pipe repairs and testing are complete. The NRC resident inspector has been informed. | |
ENS 54930 | 3 October 2020 14:13:00 | Point Beach | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Steam Generator Reactor Coolant System | During a scheduled refueling outage, a walkdown inside containment to investigate leakage revealed a pressure boundary leak upstream of 1RC-526B, HX-1B Steam Generator Channel Head Drain. This location would be considered part of the reactor coolant system as defined under 10 CFR 50.2. As such, this event is being reported pursuant to 10 CFR 50.72 (b)(3)(ii)(A). Unit 1 is currently in mode 4. Repairs for the condition are being determined. The NRC Resident Inspector has been notified. The leak rate was determined to be 0.138 gpm. | |
ENS 54897 | 11 September 2020 23:30:00 | Palisades | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | At 1930 EDT, on September 11, 2020, Palisades Nuclear Plant was conducting ultrasonic data analysis from reactor vessel closure head in-service inspections. During this analysis, signals that display characteristics consistent with primary water stress corrosion cracking were identified in head penetration 34. No leak path signal was identified during ultrasonic testing. The plant was in cold shutdown at 0% power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. This is the only indication that is currently present, however, if additional indications are found, they will also be repaired prior to the plant startup. The licensee notified the NRC Senior Resident Inspector. | ||
ENS 54707 | 13 May 2020 02:20:00 | Catawba | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | During the performance of reactor vessel closure head (RVCH) inspections, at 2220 EDT on May 12, 2020, it was determined that the Unit 1 RVCH penetration nozzle number 18 did not meet ASME code case N-729-4 requirements. A surface examination (penetrant test) identified a linear indication on nozzle number 18. The indication was not through-wall as determined by ultrasonic testing. The condition of the Unit 1 reactor vessel head penetration nozzle number 18 will be resolved prior to re-installation of the Unit 1 reactor vessel head. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 54680 | 24 April 2020 05:30:00 | Beaver Valley | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | At 0130 (EDT) on April 24, 2020, during the Beaver Valley Power Station, Unit 2 refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per ASME Code Section XI. Penetration 37 will require repair prior to returning the vessel head to service. The indication was not through wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-4 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 54614 | 26 March 2020 07:00:00 | McGuire | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Control Rod | EN Revision Imported Date : 3/30/2020 DEGRADED CONDITION On March 26, 2020, while McGuire Unit 2 was shut down for a scheduled refueling outage, the reactor vessel head penetrations were being examined in accordance with the in-service Inspection Program. Ultrasonic examinations identified a relevant indication in the Control Rod Drive Mechanism nozzle number 35 that did not meet the acceptance criteria under ASME, Section XI IWB-3600, 'Analytic Evaluation of Flaws.' Actions to address the relevant indication will be taken in accordance with the applicable codes, standards, and regulations. This event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). The relevant indication has no impact on the health and safety of the public or station employees. The NRC Resident Inspector has been notified.
McGuire is retracting the eight hour non-emergency notification made on March 26, 2020, at 10:45 ET (EN#54614). A subsequent evaluation determined that the suspect indication identified during ultrasonic examination of Control Rod Drive Mechanism nozzle number 35 is not service induced nor representative of primary water stress corrosion cracking (PWSCC). The indication has been classified as "non-relevant" and is not reportable as a degraded condition. The senior NRC Resident Inspection has been notified. Notified R2DO (Miller). | |
ENS 54463 | 4 January 2020 16:09:00 | Hatch | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Primary containment | At 1109 (EST) on 01/04/2020, it was determined that the primary containment leakage rate did not meet value La, defined in 10 CFR 50, Appendix J, 'Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors'. An additional, tested valve has been closed to maintain leakage below maximum allowable leakage, La. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
ENS 54337 | 17 October 2019 23:51:00 | Beaver Valley | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | At 1951 (EDT) on October 17, 2019, fretting indications on the reactor coolant system pressure boundary piping (pressurizer spray line) were identified. This condition does not appear to meet original construction code, ANSI B31.1, 1967 Edition thru summer 1971 Addenda. The condition will be resolved prior to plant startup. This event is being reported as an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. This condition will be corrected prior to the plant entering Mode 4.
An engineering evaluation has determined that the subject fretting is not considered a flaw, but instead is considered wear. Appendix F of Section III of the ASME Boiler and Pressure Vessel Code was applied and it was determined that the pressurizer spray line piping maintained its required design safety functions in the as-found condition. The wear has been repaired during the current refueling outage in accordance with the original construction code (ANSI B3l.l, 1967 Edition through summer 1971 Addenda) as well as Owner's Requirements. The NRC Resident Inspector has been notified. Notified R1DO (Young). | |
ENS 54031 | 26 April 2019 16:47:00 | River Bend | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Standby Liquid Control | At 1147 (CDT) on 4/26/19, a through wall leak (reported as 1 drop every 1 to 2 minutes) was identified and confirmed by operation and NDE (Non-Destructive Examination) personnel on the Standby Liquid Control injection line during pressure testing activities. The line is 1.5 inch in diameter and classified as an ASME Section Ill, Class 1 line. The leak is currently isolated from the reactor vessel by a danger tagged manual valve. The licensee notified the NRC Resident Inspector. | |
ENS 53959 | 26 March 2019 15:30:00 | Browns Ferry | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | On 3/26/2019 at 1030 CDT Engineering evaluation determined that Traversing lncore Probe (TIP) System test results related to Leak Rate Testing of 2-CKV-76-653, TIP Purge Header Check Valve, during the Unit 2 Refueling Outage resulted in a reportable condition. On 3/24/2019 at 1558 CDT, Leak Rate Testing identified a (local leak rate test) LLRT failure of 2-CKV-76-653. The gross leakage Leak Rate value exceeded the Technical Specification allowable value for Type C valves of less than 0.6 (allowable leakage) La. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The short-term corrective actions include repairing the valve such that it passes the test. The valve needs to be repaired before the unit can change modes. | ||
ENS 53837 | 18 January 2019 19:33:00 | Waterford | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | This is a non-emergency notification from Waterford 3. On January 18, 2019, a relevant indication was detected in the performance of Phased Array Ultrasonic Examinations of A600 Dissimilar Metal Piping Welds during planned inspections. The indication was observed during the analysis of data recorded of the Reactor Coolant System (RCS) Loop 2A Reactor Coolant Pump Suction Drain Nozzle to Safe-End Butt Weld (11-007). This indication does not meet applicable acceptance criteria under American Society of Mechanical Engineers (ASME) Section XI. The plant was in Mode 6 (Refueling) at 0 percent power for a planned refueling outage at the time of discovery. The condition will be resolved prior to plant startup. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50 .72(b)(3)(ii)(A), 'Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded,' because an indication was found that did not meet acceptance criteria referenced in ASME Section XI , IWB-3514-2 and Code Case N-770-2, 3132. The NRC Resident Inspector has been notified. Reference: CR-WF3-2019-01041 | |
ENS 53834 | 17 January 2019 06:00:00 | Waterford | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System | This is a non-emergency notification from Waterford 3. On January 17, 2019, a relevant indication was detected in the performance of Phased Array Ultrasonic Examinations of A600 Dissimilar Metal Piping Welds during planned inspections. The indication was observed during the analysis of data recorded of the Reactor Coolant System (RCS) Loop 1A Reactor Coolant Pump Suction Drain Nozzle to Safe-End Butt Weld (07-009). This indication does not meet applicable acceptance criteria under American Society of Mechanical Engineers (ASME) Section Xl. The plant was in Mode 6 (Refueling) at 0 percent power for a planned refueling outage at the time of discovery. The condition will be resolved prior to plant startup. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A), 'Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded,' because an indication was found that did not meet acceptance criteria referenced in ASME Section Xl. IWB-3514-2 and Code Case N-770-2, 3132. The NRC Resident Inspector has been notified. Reference: CR-WF3-2019-0967 | |
ENS 53749 | 21 November 2018 05:00:00 | Palisades | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | On November 21, 2018, during an extent of condition review, after completion of ultrasonic testing, further interrogation of reactor vessel closure head (RVCH) penetration 36 was performed using eddy current testing. The testing detected three repairable indications. No indication of boric acid leakage was identified at this location during the bare metal visual inspection. Extent of condition review is complete on all RVCH penetrations. The plant was in cold shutdown at 0 percent power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector. | ||
ENS 53734 | 11 November 2018 04:00:00 | Palisades | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | On November 11, 2018, during ultrasonic data analysis from reactor vessel closure head in-service inspections, signals that display characteristics consistent with primary water stress corrosion cracking in head penetration 33 were identified. No indications of boric acid leakage and no surface indications were detected at this location during bare metal visual inspection. The plant was in cold shutdown at 0% power and in Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector. | ||
ENS 53733 | 10 November 2018 05:00:00 | Palisades | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | On November 10, 2018, during a planned bare metal visual inspection of the reactor head, boric acid was discovered at a CRDM (Control Rod Drive Mechanism) nozzle to reactor head penetration. Investigation of the source of the boric acid is ongoing. The plant was in cold shutdown at 0% power and Mode 6 for a refueling outage at the time of discovery. Repair actions will be completed prior to plant startup from the outage. All other reactor vessel head penetrations have had a bare metal visual inspection completed with no other indications identified. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) for degradation of a principal safety barrier. The licensee notified the NRC Senior Resident Inspector. | ||
ENS 53711 | 1 November 2018 04:00:00 | Beaver Valley | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | On 11/01/2018, during the Beaver Valley Power Station Unit No. 2 (BVPS-2) refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per ASME Code Section XI. Penetration 27 will require repair prior to returning the vessel head to service. The indication was not through wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-4 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. The other 65 penetrations will be examined during the 2R20 (current) refueling outage. The plant is currently shutdown and in Undefined Mode. The reactor vessel head is not currently installed. Repairs are currently being planned and will be completed prior to startup. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as-found indications did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-4 to remain in-service without repair. The NRC Resident Inspector has been notified. |