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At 0634 EDT on August 11, 2021 (high pressure coolant injection) HPCI was declared inoperable due to a pump flow controller problem. The cause of the controller problem is unknown at this time and is under investigation. (Reactor core isolation cooling) RCIC was verified operable per Tech Spec 3.5.1 E.1. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

• * * RETRACTION FROM WHITNEY HEMMINGWAY TO KAREN COTTON ON 10/6/2021 AT 1036 EDT * * *

The purpose of this notification is to retract a previous report made on August 11, 2021 (EN 55403). At 0634 EDT on August 11, 2021, an unplanned inoperability of the High Pressure Coolant Injection system (HPCI) was reported pursuant to 10 CFR 50.72(b)(3)(v)(D) by EN 55403. HPCI was declared inoperable due to receipt of an alarm associated with the pump flow controller. The HPCI system operating procedure states that HPCI should be declared inoperable when this alarm is received. The cause of the alarm, a loose transmitter connection, was identified and corrected. Following clearance of the alarm, HPCI was declared operable at approximately 1930 EDT on August 11, 2021. This alarm indicated a fault in the signal from the transmitter to the HPCI flow controller; in this case, the HPCI flow controller would have continuously called for maximum HPCI flow. The controller is configured with a high limiter to prevent an overspeed trip. An engineering evaluation of the event identified that HPCI was capable of performing its required safety functions while this alarm was present. The condition was that the HPCI flow controller would have continuously called for maximum HPCI flow upon HPCI initiation, however operators would be able to manually control HPCI flow upon HPCI initiation. Additionally HPCI would have run until Reactor Pressure Vessel (RPV) level reached Level 8 where it would trip until RPV level decreased to Level 2 then automatically restart. The licensee notified the NRC Resident Inspector. Notified R3DO (Peterson).

On 01/11/2019 at 0958 EST, the Fermi 2 Active Seismic Monitoring system was taken out of service for planned maintenance. During the maintenance activity, the Active Seismic Monitoring System failed a planned surveillance test and was not restored to operability within 72 hours. Compensatory measures to provide alternative methods for event classification of a seismic event were implemented in accordance with the Fermi 2 Emergency Plan procedures prior to the start of the planned maintenance outage. The planned outage time to restore operability exceeded 72 hours on January 14th, 2019, at 0958 EST. Repairs have been completed, the Active Seismic Monitoring System has been declared Functional at 1037 EST, January 14th, 2019, and declared Operable at 1109 EST, January 14th, 2019.

The loss of the Active Seismic Monitoring System is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). No seismic activity has been felt onsite and the United States Geological Survey (USGS) recorded no seismic activity in the area. The NRC Resident Inspector has been notified. Femi 2 has two seismic monitors, one on the Reactor Pressure Vessel Pedestal and one in the High Pressure Core Injection (HPCI) room. Only the HPCI room monitor was inoperable.

On May 2, 2017, while performing a past operability review associated with Combustion Turbine Generator (CTG) 11-1, it was determined that a past configuration of CTG 11-1 could not have assured all of the applicable Appendix R success criteria under all of the postulated scenarios described in the Updated Final Safety Analysis Report (UFSAR). From November 21, 2016 until March 18, 2017 when Mode 4 was entered, CTG 11-1 was in a configuration where it could not be started from the dedicated shutdown panel, although it could be started locally. One of the specific scenarios for Appendix R in the UFSAR credits CTG 11-1 to support a safe shutdown based on an assumed time required to start CTG 11-1 and then provide flow to the reactor pressure vessel using the Standby Feedwater System. During the time period where CTG 11-1 could only be started locally, this assumed time would have been exceeded. Therefore, this event is being reported as an 'unanalyzed condition that significantly affects plant safety' under 50.72(b)(3)(ii)(B). At the time of discovery, CTG 11-1 was fully operable and the described condition had already been corrected. The NRC Resident Inspector has been notified.

• * * RETRACTION PROVIDED BY GREG MILLER TO JEFF ROTTON AT 1532 EDT ON 05/19/2017 * * *

The purpose of this notification is to retract a previous report made on May 2, 2017 (EN 52724) under 10 CFR 50.72(b)(3)(ii)(B). The notification to the NRC involved an event where Combustion Turbine Generator (CTG) 11-1 could only be started locally such that required operator actions during an Appendix R safe shutdown scenario could be delayed. Subsequent to the initial notification, the event, additional site documentation, and the NRC guidance in NUREG-1022 pertaining to 10 CFR 50.72(b)(3)(ii)(B) were reviewed further. lt was verified that Fermi 2 procedures contained actions to ensure Appendix R safe shutdown capability under the plant conditions during the relevant time period. A time validation study was performed (May 9, 2017) which verified that the operator actions could have been completed within the time described in the Updated Final Safety Analysis Report (UFSAR) for initiating Standby Feedwater flow to the reactor pressure vessel. In addition, a review of the supporting design calculation identified margin in the required time described in the UFSAR. Based on this information, the condition of CTG 11-1 during the time period from November 21, 2016 until March 18, 2017 would not have prevented compliance with the Appendix R safe shutdown requirements. Under these circumstances, the event does not represent an 'unanalyzed condition that significantly affects plant safety' under 10 CFR 50.72(b)(3)(ii)(B) per the guidance in NUREG-1022. Therefore, EN 52724 can be retracted and no Licensee Event Report (LER) under 10 CFR 50.73(a)(2)(ii)(B) is required to be submitted. The licensee has notified the NRC Resident Inspector. Notified R3DO (Cameron).

At 2305 EDT on September 13, 2015, a manual scram was initiated in response to a loss of all Turbine Building Closed Cooling Water (TBCCW). All control rods fully inserted. The lowest Reactor Water Level (RWL) reached was 137 inches. All isolations and actuations for RWL 3 occurred as expected. Decay heat was initially being removed through the Main Turbine Bypass System to the Main Condenser, however, as a result of the loss of TBCCW, the Main Feed Pumps lost cooling and had to be secured. At 2310, Standby Feedwater was initiated and Main Feedwater was secured. The loss of TBCCW also caused all Station Air Compressors (SACs) to trip on loss of cooling. The loss of SACs caused the Instrument Air header pressure to degrade to the point at which the Secondary Containment isolation dampers drifted closed. This resulted in the Reactor Building vacuum exceeding the Technical Specification limit. At 2325, operators started the Standby Gas Treatment system and manually initiated a Secondary Containment isolation signal. Secondary Containment vacuum was promptly restored to within Technical Specification limits. Additionally, Operators were monitoring for expected MSIV drift due to the degraded Instrument Air header pressure. When outboard MSIVs were observed to be drifting, Operators closed the outboard and inboard MSIVs at 2345. At 2352, Safety Relief Valves (SRVs) reached the Low-Low Setpoint and began cycling to control reactor pressure. RWL is currently being maintained in the normal level band with the Standby Feedwater and Control Rod Drive systems. Reactor Pressure is being controlled with Safety Relief Valves. Operators are currently in the Emergency Operating Procedure for Reactor Pressure Vessel control. Investigation into the loss of TBCCW continues. No safety-related equipment was out of service at the time of the event. All offsite power sources were adequate and available throughout the duration of the event. The NRC resident inspector has been notified.

• * * UPDATE AT 0555 EDT AT 09/14/15 FROM CHRIS ROBINSON TO JEFF HERRERA * * *

At 0409 EDT the Reactor Core Isolation Cooling (RCIC) system was placed in service due to identification of an unisolable leak in the Standby Feedwater System. Reactor water level and pressure is now being controlled though the RCIC system and Safety Relief Valves. This event update is reportable as a valid manual initiation of a specified safety system under 10CFR50.72(b)(3)(iv)(A). The NRC resident inspector has been notified. The leak rate was reported as approximately 5-10 gallons per minute from a weld on the standby feedwater pump header drain valve F326. The licensee reported the leak stopped once RCIC was placed into service. The licensee is still investigating the issue. Notified the R3DO (Pelke), IRD Manager (Grant), NRR EO (Morris).

• * * UPDATE PROVIDED BY CHRIS ROBINSON TO JEFF ROTTON AT 2135 EDT ON 09/14/2015 * * *

At 1847 EDT on September 14, 2015, a valid automatic Reactor Protection System (RPS) actuation occurred due to Reactor Water Level 3 while shutdown in MODE 3. Operators were manually controlling Reactor Pressure Vessel (RPV) level and pressure with Reactor Core Isolation Cooling (RCIC) and Safety Relief Valves (SRV). While operators were cycling SRVs, the RPV level went below the Level 3 setpoint. Operators promptly restored RPV level by manual operation of RCIC. The Level 3 actuation and associated isolations were verified to operate properly. The scram signal has been reset. Fermi 2 remains in MODE 3 controlling RPV Level and Pressure through manual operation of RCIC and SRVs. This is the second occurrence of a valid specified safety system actuation reportable under 10CFR50.72(b)(3)(iv)(A) for this ongoing event. The NRC Resident Inspector has been notified. Notified R3DO (Riemer), IRD Manager (Grant), and NRR EO (Morris)

• * * UPDATE FROM BRETT JEBBIA TO JOHN SHOEMAKER AT 1446 EST ON 2/27/16 * * *

This update provides clarification of the applicable reporting criteria for this Event associated with primary containment isolation actuations. Upon the manual reactor scram at 2305 EDT on September 13, 2015, Reactor Protection System (RPS) Level 3 actuated and Primary Containment Isolation System (PCIS) Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for these actuations is 10 CFR 50.72(b)(3)(iv)(A). The applicable reporting criterion for the manual closure of the inboard and outboard main steam isolation valves at 2345 EDT on September 13, 2015, is also 10 CFR 50.72(b)(3)(iv)(A). In addition, the manual closures of all MSIV lead to a loss of condenser vacuum which resulted in the actuation of PCIS Group 1 at 0001 EDT on September 14, 2015, as expected. The applicable reporting criterion for this actuation is also 10 CFR 50.72(b)(3)(iv)(A). Upon reaching Level 3 at 1847 EDT on September 14, 2015, PCIS Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for this actuation is 10 CFR 50.72(b)(3)(iv)(A). The licensee informed the NRC Resident Inspector. Notified the R3DO (Stone).

Boroscopic inspection of the reactor head assembly during refuel outage A1R17 has determined that white powder residue observed at reactor head penetrations 41, 49, 61, 65, and 73 is evidence of past Reactor Coolant System pressure boundary leakage at one or more of these locations. This determination was made today at 0510 (CDT) after completion of boroscopic inspections and consultation with the reactor head vendor. These reactor head penetrations are control rod drive mechanism (CRDM) penetrations. The location of the identified leakage is suspected to be the omega seal welded threaded connection on one or more of these CRDM penetrations. Further inspections must be performed to confirm the exact leakage location. This condition will be repaired before U-1 returns to mode 4. The licensee has notified the NRC Resident Inspector.

• * * RETRACTION ON 9/27/13 AT 1529 EDT FROM MATTHEW HENSON TO DONG PARK * * *

The purpose of this report is to retract ENS report #49353 (September 18, 2013). This report was made after boroscopic inspection of the Unit 1 reactor head assembly during refueling outage A1R17 determined that white powder residue observed at five Reactor Pressure Vessel (RPV) head penetrations was evidence of past Reactor Coolant System (RCS) pressure boundary leakage at one or more of these locations. At 1400 CDT on Wednesday, September 25, 2013, the Braidwood Nuclear Power Station (BNPS) concluded that the prior ENS notification could be retracted because the Plant Operations Review Committee had concluded that the white powder residue observed at five Unit 1 RPV head penetrations was not evidence of past RCS pressure boundary leakage. A detailed examination plan was implemented, which included a disassembly of the Digital Rod Position Indication (DRPI) and Control Rod Drive Mechanism (CRDM) coils for all five penetrations and a detailed visual examination using remote video equipments. The visual examinations used remote video equipments to inspect the lower and intermediate canopy seal weld areas for the five CRDMs. In addition, an examination of the upper canopy seal weld areas and a bare-metal inspection of the Reactor Pressure Vessel Head did not identify any evidence of boric acid leakage. Based on these inspections and a review of available photos from previous outages, BNPS concluded that the white residues were from incomplete cleaning in prior outages. The licensee has notified the NRC Resident Inspector. Notified R3DO (Dickson).

At 1855 EDT on 10/29/2012, the licensee declared a Notice of Unusual Event (NOUE) per criteria HU4 for high water level in the station intake structure of greater than 4.5 feet. At the time of the notification, water level in the intake structure was approximate 4.8 feet and slowly rising. The cause of the increased water level was due to storm surge associated with Hurricane Sandy. No other station impacts were reported at the time. The licensee continues to monitor the intake levels and ocean tides. The licensee has notified the NRC Resident Inspector and the State of New Jersey.

• * * ALERT UPDATE ON 10/29/2012 AT 2141 EDT FROM STEVE SERPE TO RYAN ALEXANDER * * *

At 2044 EDT on 10/29/2012, the licensee escalated its emergency declaration to an Alert per criteria HA4 for high water level in the station intake structure of greater than 6.0 feet. At the time of the notification, water level in the intake structure was approximately 6.6 feet. The site also experienced a loss of offsite power event concurrent with the additional water level increase. Both emergency diesel generators started and are supplying power to the emergency electrical busses. Shutdown cooling and spent fuel pool cooling have been restored. Reactor pressure vessel level is steady at 584.7 inches. Intake levels continues to rise slowly and the licensee is monitoring. The licensee has notified the NRC Resident Inspector and the State of New Jersey. Notified DHS SWO, FEMA Ops Center, USDA Ops Center, HHS Ops Center, DOE Ops Center, DHS NICC Watch Officer, EPA EOC, and NuclearSSA via e-mail.

• * * UPDATE on 10/30/12 at 0414 EDT FROM GILBERT DEVRIES TO RYAN ALEXANDER * * *

The licensee updated this report with an 8-hour non-emergency notification of emergency diesel generator auto-actuation due to the actual loss of off-site power event (which occurred at 2018 EDT on 10/29/2012). This event caused a valid RPS actuation with automatic containment isolations that resulted in a temporary loss of shut-down cooling to the reactor. Shutdown cooling was subsequently restored with power provided by the emergency diesel generators. The licensee has notified the NRC Resident Inspector. Notified R1DO (Caruso).

• * * UPDATE AT 0357 EDT ON 10/31/12 FROM GILBERT A. DeVRIES TO S. SANDIN * * *

Termination of Alert. The Oyster Creek Station has terminated the Alert that was declared at 2044 (EDT) on 10/29/12 due to Intake Structure high water level greater than 6.0 ft. MSL (EAL HA4). Intake water level has returned to normal and is now below the Unusual Event EAL threshold (4.5 ft. MSL) and continues to lower. The licensee informed state and local agencies and the NRC Resident Inspector. Notified Region I IRC (Clifford), NRR (Evans), and IRD (Marshall). Notified DHS SWO, FEMA Ops Center, USDA Ops Center, HHS Ops Center, DOE Ops Center, DHS NICC Watch Officer, EPA EOC, and NuclearSSA via e-mail.