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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56411 | 15 March 2023 03:57:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Reactor Coolant System (RCS) Boundary Degraded Condition | 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) | Reactor Coolant System Main Steam | |
| ENS 56371 | 18 February 2023 10:39:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition | 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 56257 | 3 December 2022 16:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Discovered on Shutdown Cooling Test Line | 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. | Shutdown Cooling Residual Heat Removal | |
| ENS 55706 | 16 January 2022 05:20:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Discovered on Shutdown Cooling Test Line | 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. | Shutdown Cooling | |
| ENS 53959 | 26 March 2019 15:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Report of Degraded Condition Due to Leaking Valve | 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 45107 | 31 May 2009 18:30:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Degraded Condition Due to Discovery of Pressure Boundary Leakage | |||
| ENS 44680 | 23 November 2008 18:00:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leak | During performance of Unit 1 vessel leak test in accordance with 1-SI-3.3.1.A - ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping, a pressure boundary leak was discovered on an instrument line connected to the reactor pressure vessel. This instrument line is a ASME Code Class 1 equivalent component (nozzle safe end) on pressure vessel nozzle N11B and connecting upstream of primary containment penetration X-29B. This caused entry into Technical Requirements Manual (TRM) 3.4.3 - Structural Integrity - Condition A - in which the applicability is at all times and the required action is to immediately restore the structural integrity of the affected component to within its limit or maintain the reactor in MODE 4 or 5 or the reactor coolant system less than 50?F above the minimum temperature required by NDT considerations, until each indication of a defect has been investigated and evaluated. The plant is currently in MODE 4. This event is reportable within 8 hours under 10CFR50.72 (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. This event is also a reportable within 60 days under 10CFR50.73(a)(2)(ii)(A) any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC Resident Inspector has been notified. This condition has been documented in the BFN Corrective Action program as PER# 157918. Unit 2 and 3 remain at 100% power and are not affected by this event. | Reactor Coolant System Primary containment Reactor Pressure Vessel | |
| ENS 43659 | 22 September 2007 17:45:00 | 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | Pressure Boundary Leakage Discovered During Drywell Inspection | As part of a planned outage for Browns Ferry Unit 3, initial drywell leak inspections were performed after shutdown (mode 3). This inspection identified a weld defect in Residual Heat Removal (RHR) piping. The defect was in a one inch test line near manually operated valve 3-74-638B. This is classified as pressure boundary leakage and the piping is rated as ASME code class 1. The leak rate was estimated by visual observation at less than 0.25 gpm. Investigation is continuing into the cause of the weld defect. Unit 1 and 2 remain at full power and are not affected by this event. This event is reportable within 8 hours under 10 CFR 50.72(b)(3)(ii)(A) and 10 CFR 50.73(a)(2)(ii)(A) as 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The licensee plans to continue to Mode 4 (Cold Shutdown) as required by Tech Specs for pressure boundary leakage. The NRC resident inspector has been notified. | Residual Heat Removal |