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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.

• * * RETRACTION ON 03/28/2023 AT 1059 EST FROM CASEY CARTWRIGHT TO THOMAS HERRITY * * *

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)

At 0315 MST on May 19, 2021, Unit 2 reactor automatically tripped during testing of the Plant Protection System. The Reactor Protection System actuated to trip the reactor on High Pressurizer Pressure, although no plant protection setpoints were exceeded. Main Steam Isolation Signal (MSIS), Safety Injection Actuation Signal (SIAS), and Containment Isolation Actuation Signal (CIAS) were received. No injection of water into the Reactor Coolant System occurred. Auxiliary Feedwater Actuation Signals (AFAS) 1 and 2 actuated on low Steam Generator water level post trip as designed. This event is being reported as a reactor protection system and a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Following the reactor trip, all (Control Element Assemblies) CEAs inserted fully into the core. All systems operated as expected. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. Unit 2 is stable and in Mode 3. Steam Generator heat removal is via the class 1 E powered motor driven auxiliary feedwater pump and Atmospheric Dump Valves. The NRC Senior Resident Inspector has been informed.

• * * UPDATE ON 5/19/21 AT 1351 EDT FROM JASON HILL TO BRIAN P. SMITH * * *

The Unit 2 reactor tripped because of actual High Pressurizer Pressure that occurred as a result of a Main Steam Isolation Signal actuation. At 0337 MST, both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) were made inoperable when the injection valves were overridden and closed in accordance with station procedures. At 0346 MST, in accordance with station procedures, both trains of Containment Spray, LPSI, and HPSI pumps were overridden and stopped, rendering Containment Spray inoperable as well. This represents a condition that would have prevented the fulfillment of a safety function required to mitigate the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). Additionally, at the time of the Safety Injection Actuation Signal (0315 MST), both trains of Emergency Diesel Generators actuated as required and both 4160 VAC busses remained energized from off-site power. The NRC Senior Resident Inspector has been informed. Notified R4DO (Young)

• * * UPDATE ON 7/02/21 AT 1943 EDT FROM YOLANDA GOOD TO JEFFREY WHITED * * *

The inoperability of both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) and both trains of Containment Spray (CS) following the Unit 2 reactor trip has been determined to be an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Additionally, inoperability of both trains of HPSI resulted in a reportable condition that could prevent fulfillment of its credited safety function to maintain the reactor in a safe shutdown condition per 10 CFR 50. 72(b)(3)(v)(A). The additional reporting criteria were discovered during review of the event and corresponding safety analyses. The NRC Senior Resident Inspector has been informed. Notified R4DO (Werner)

EN Revision Text: INOPERABILITY OF EQUIPMENT FOR CONTROL OF RADIOLOGICAL RELEASE At 2130 CST on 11/27/2018, Division 1 Main Steam Positive Leakage Control System (MS-PLCS) was declared inoperable because of a leaking check valve that caused excessive cycling of the associated air compressor. Division 2 MS-PLCS had been declared inoperable on 11/27/2018 at 1400 CST when a pressure control valve in the system exceeded the maximum allowable stroke time. Because MS-PLCS supplements the isolation function of the main steam isolation valves (MSIVs) by processing fission products that could leak through the closed MSIVs, both divisions of MS-PLCS inoperable at the same time represents a condition that could prevent the fulfillment of a safety function of an SSC (Structures, Systems and Components) that is needed to control the release of radioactive material. The station diesel air compressor is available to supply backup air to the safety relief valves as required by the Technical Requirements Manual." (This is associated with operability of the safety relief valves, due to the inoperable MS-PLCS air compressor.) The unit is in a 7 day shutdown Limiting Condition for Operation (LCO), 1-TS1-18-Div 1 & 2 MSPLCS-685, for the two divisions of MS-PLCS being inoperable. The licensee notified the NRC Resident Inspector.

• * * RETRACTION ON 12/03/18 AT 1551 EST FROM TIM GATES TO BETHANY CECERE * * *

This event was initially reported under 10 CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the Main Steam Positive Leakage Control System (MS-PLCS) from fulfilling its safety function to control the release of radioactive material. Division I was declared inoperable due to a failed component. Division II was declared inoperable due to a pressure control valve in the system exceeding the maximum allowable time to close by 0.50 seconds. An engineering evaluation has since been performed and concluded that the 2 second maximum allowable time to close was based on the pressure control valve being classified as a rapid closure valve and was established from the original baseline data of 0.50 seconds. This baseline data is an administrative target value per the In-Service Testing Program. There are no technical specification requirements associated with the 2 second closure time. The engineering evaluation also determined that the volume of air supplied through the pressure control valve during the extra 0.50 seconds of valve closure would have an inconsequential effect on the pressure within the volume of leakage barrier between the Main Steam Isolation Valves associated with the MS-PLCS pressure control valve or have any effect on containment over-pressurization. Based on the information provided by the engineering evaluation, the Division II MS-PLCS has been declared operable-degraded non-conforming since time of initial discovery. Consequently, this event is not reportable as a condition that could have prevented the Main Steam Positive Leakage Control System (MS-PLCS) from fulfilling its safety function. The (NRC) Resident Inspector has been notified via e-mail. Notified the R4DO (Gaddy).

On 11/7/2017 at 1957 (EST), VC Summer Nuclear Station automatically tripped due to a turbine trip. The cause of the turbine trip is under investigation at this time. All systems responded as expected. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. This event is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The unit is currently stable in Mode 3 with decay heat removal via the Main Steam to the Main Condenser. The NRC Resident Inspector has been notified. The licensee will notify the South Carolina State Emergency Management Division, the Fairfield, Richland, Lexington and Newberry Counties.

• * * UPDATE FROM BETH DALICK TO VINCE KLCO ON 11/8/17 AT 1409 EST * * *

All systems responded as expected, with the exception of 'B' Steam Generator Feedwater Isolation Valve XVG1611 B-FW. This valve did not appear to automatically close and was slow to indicate closed from the Main Control Board. All Control Rods fully inserted and all Emergency Feedwater pumps started as required. The plant is stable in Mode 3. Notified the R2DO (Musser).

A fault occurred on the unit 2 "B" main bank transformer resulting in an oil fire. The main turbine tripped resulting in a reactor trip. All control rods fully inserted and no safety or relief valves lifted. Decay heat is being removed via steam dumps to the main condenser and feeding steam generators with auxiliary feedwater. Electrical power is through the normal shutdown electrical lineup. Offsite assistance was requested from the county and off duty fire brigade members. At 2228, the fire was reported as out. Spray is continuing and a reflash watch is being set. Unit 1 continued to operate at 100% power throughout the event. Notified the DHS SWO, FEMA Ops Center, DHS NICC, FEMA National Watch Center (E-mail) and Nuclear SSA (E-mail).

• * * UPDATE AT 2352 EDT ON 08/30/2016 FROM MICHAEL BOTTORFF TO JEFF HERRERA * * *

On August 30, 2016, at 2110 EDT, Watts Bar Nuclear Plant Unit 2 reactor tripped due to an electrical fault affecting the 2B Main Bank Transformer, resulting in a fire in the transformer. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and main steam dump systems. Unit 2 is in a normal shutdown electrical alignment. The fire was out at 2230 EDT. The cause of the fire is currently under investigation. The fire was reported at 2149 EDT. Local Fire Departments responded to the site as requested. The reactor trip and system actuation is being reported under 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 1. The NOUE was exited at 2342. The NRC Senior Resident Inspector has been notified. The Licensee notified the State of Tennessee. Notified the R2DO (Bartley), IRD MOC (Stapleton), NRR EO (Miller), DHS SWO, FEMA Ops Center, DHS NICC, FEMA National Watch Center (E-mail) and Nuclear SSA (E-mail).

This four and eight hour notification is being made to report that at 0541 (EDT) on 7/28/16, Salem Unit 1 initiated a shutdown to comply with Technical Specifications due to the inoperability of both source range nuclear instruments. During a reactor startup, with Unit 1 in Mode 2, both source range instruments were reading approximately one decade lower than expected compared to intermediate range and Gamma-Metric instruments and due to the proximity to the estimated critical condition. The condition could also have prevented the fulfilment of the source range instruments safety function to trip the reactor when required. Salem Unit 1 is currently stable in Mode 3. Reactor Coolant system pressure is 2235 psig and reactor coolant system temperature is 547 F with decay heat removal via the main steam dump and auxiliary feedwater systems. Unit 1 has one active shutdown tech spec action statement in effect due to the inoperability of the containment radiation monitor 1R11A. The inoperability of this radiation monitor had no effect on the event. All control rods were manually inserted to place Unit 1 in Hot Standby (Mode 3). No ECCS (emergency core cooling system) or ESF (emergency safety features) systems were required to function during this event. No major secondary equipment was tagged for maintenance prior to this event. No personnel were injured during this event. The reactor was manually shut down and a shutdown margin calculation verified sufficient margin. The licensee notified the NRC Resident Inspector and the local township.

• * * RETRACTION FROM MATT MOG TO VINCE KLCO ON 9/26/16 AT 1519 EDT * * *

The purpose of this notification is to retract event report number 52129 made on 7/28/2016 at 0925 (EDT). Previously PSEG reported that Salem Unit 1 initiated a shutdown to comply with Technical Specifications (TS) due to the inoperability of both source range nuclear instruments. Additionally PSEG reported that the condition could have prevented the fulfillment of the safety function needed to, 'Shut down the reactor and maintain it in a safe shutdown condition.' Subsequent review identified that the condition did not meet either reporting criteria. Maintenance and Engineering evaluation of the source range nuclear instruments determined that the instruments were fully operable at the time of the event. TS 3.3.1.1, Reactor Trip Instrumentation remained met, no TS shutdown was required and the instruments were capable of performing their required function. Therefore PSEG is retracting the notifications made under 10 CFR 50.72(b)(2)(i) and 10 CFR 50.72(b)(3)(v)(A). The NRC Resident Inspector was notified of this retraction by the licensee. Notified the R1DO (Cook).