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The following information was provided by the licensee via email: During thermography of a reactor protection system (RPS) distribution panel, a circuit breaker (RPS-CB-7B) was inadvertently opened. This resulted in a partial loss of power to RPS Division B, which caused containment isolations to occur in multiple systems (Reactor Water Clean Up, Equipment Drains Radioactive, Floor Drains Radioactive, Reactor Recirculation, and Traversing lncore Probe). Specifically, RWCU-V-1, FDR-V-3, EDR-V-19, RRC-V-19, and TIP-V-15 all closed. All actuations occurred as designed upon the partial loss of RPS power. This event is being reported pursuant to 10 CFR 50.72(b)(3)(iv)(A) due to an unplanned valid actuation of a system pursuant to 10 CFR 50.72(b)(3)(iv)(B)(2). Additionally, this is being reported pursuant to 10 CFR 50.72 (b)(3)(xiii) for a major loss of emergency assessment capability due to the inability to assess primary containment identified and unidentified leakage rates. Emergency assessment capability was restored at 1008 PDT upon system restoration. The NRC resident was notified by the licensee.

• * * UPDATE FROM SIMEON MORALES TO DONALD NORWOOD AT 1547 EDT ON 6/16/2022 * * *

The following information was received via email: This event is being reported pursuant to 10 CFR 50.72 (b)(3)(xiii) only for a major loss of emergency assessment capability due to the inability to assess primary containment identified and unidentified leakage rates. The containment isolation was not due to actual plant conditions or parameters meeting design criteria for containment isolation. Therefore, this is considered an invalid actuation. Updated ENS Text: During thermography of a reactor protection system (RPS) distribution panel, a circuit breaker (RPS-CB-7B) was inadvertently opened. This resulted in a partial loss of power to RPS Division B, which caused containment isolations to occur in multiple systems (Reactor Water Clean Up, Equipment Drains Radioactive, Floor Drains Radioactive, Reactor Recirculation, and Traversing Incore Probe). Specifically, RWCU-V-1, FDR-V-3, EDR-V-19, RRC-V-19, and TIP-V-15 all closed. All actuations occurred as designed upon the partial loss of RPS power. This is being reported pursuant to 10 CFR 50.72 (b)(3)(xiii) for a major loss of emergency assessment capability due to the inability to assess primary containment identified and unidentified leakage rates. Emergency assessment capability was restored at 1008 PDT upon system restoration. The plant is stable, and all effected systems have been restored. There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified. Notified R4DO (Azua).

• * * UPDATE FROM TRACY HOWARD TO ERNEST WEST AT 1853 EDT ON 8/10/2022 * * *

The following information was received via email: At 0923 (PDT) on June 13, 2022, a partial loss of power to the Reactor Protection System (RPS) 'B' occurred due to the inadvertent opening of circuit breaker RPS-CB-7B during thermography of RPS-PP-C72/P001. The partial loss of RPS 'B' resulted in closure of primary containment isolation valves (PCIVs) in multiple systems. No plant parameters existed which would cause the opening of RPS-CB-7B or actuation of the primary containment isolation; therefore, this is considered to be an invalid actuation of a system listed in 10 CFR 50.73(a)(iv)(B). The closure of PCIVs were expected responses to the partial loss of RPS 'B'. Circuit breaker RPS-CB-7B was closed lo restore energy lo RPS 'B' at 1008 (PDT), containment isolation valves were opened, and the affected systems were returned to normal operating conditions for the current configuration per plant procedures. As indicated in 10 CFR 50.73(a)(1), in the case of an invalid actuation reported under 10 CFR 50.73(a)(2)(iv)(A), the licensee may, at its option, provide a telephonic notification lo the NRC Operations Center within 60 days of discovery of the event instead of submitting a written Licensee Event Report. This 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) for invalid actuations reported under 10 CFR 50.73 (a)(2)(iv)(A). This actuation was invalid since it was caused by maintenance activities and not the result of actual plant conditions warranting containment isolation. The following additional information is provided as specified in NUREG-1022: The following inboard containment isolation valves were actuated when personnel inadvertently bumped into RPS-CB-7B during the removal of a panel � RWCU-V-1 Reactor Water Cleanup Suction Inboard Isolation Valve � EDR-V-19 Drywell Equipment Drain Inboard Isolation Valve � FDR-V-3 Drywell Floor Drain Inboard Isolation Valve � RRC-V-19 Reactor Water Sample Inboard Isolation Valve � TIP-V-15 Traversing In-Core Probe Purge Isolation Valve All actuations occurred as designed upon the partial loss of RPS power. There were no actual safety consequences associated with this event since all affected equipment responded as designed. The NRG Residents have been notified. Notified R4DO (O'Keefe).

The Licensee provided the following information via fax: During performance of a surveillance of the High Pressure Core Spray (HPCS) service water system on January 1, 2022, the HPCS system was declared inoperable for performance of the surveillance. During the surveillance, pump discharge pressure and flow were above the action range curve specified in the surveillance. For the given flow rate, pump discharge pressure was too high. This condition prevents declaring the HPCS service water system and HPCS system operable. The HPCS service water and HPCS systems remain inoperable. The station entered Technical Specification (TS) 3.7.2.A and TS 3.5.1.B at 0910 (PST) on January 1, 2022. In accordance with TS 3.5.1.B, the Reactor Core Isolation Cooling (RCIC) system was verified to be operable. TS 3.5.1 Action B provides a 14-day completion time to restore HPCS to an operable status. All other Emergency Core Cooling systems (ECCS) are operable. This event is being reported as an event or condition that could have prevented the fulfillment of a safety function credited for mitigating the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). The HPCS system is a single train system at Columbia. The NRC resident has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee is investigating the cause of the high pump discharge pressure and verifying instrumentation accuracy.

• * * RETRACTION ON 1/6/22 AT 1715 EST FROM CHASE WILLIAMS TO TOM KENDZIA * * *

This Notification is to retract EN 55682, Unplanned High Pressure Core Spray (HPCS) Inoperability. On 1/1/2022 at (1735 EST), Columbia Generating Station notified the NRC under 10 CPR 50.72(b)(3)(v)(D) of the inoperability of a single train of safety system (HPCS) for performance of the surveillance. During the surveillance pump discharge pressure and flow were above the action range curve specified in the surveillance. Engineering performed an analysis of this event and concluded the HPCS was operable during the event and would have performed its required safety function. The results of initial IST testing of HPCS-P-2 via OSP-SW/IST-Q703 on 01/01/22 resulted in measured parameters falling outside of the acceptable range specified for this pump. Systematic error was suspected as the cause of the failure and the test was reperformed following taking actions to eliminate the suspected systematic errors. The second performance of the test on 01/01/22 resulted in acceptable pump performance. Evidence exists that the initial performance of the test failed due to imprecise averaging techniques due to difficulties in averaging continuously changing values on the test instrument. The second performance of OSP-SW/IST-Q703 should be considered a successful test and the test of record as the systematic error was eliminated and measured parameters are considered valid. The NRC Resident Inspector has been notified. The HOO notified R4DO (Rolando-Otero).

Nine Mile Point unit 2 experienced an unusual event due to a small fire in the turbine building that was immediately extinguished and then reflashed. The fire was declared out at 1119 (EST), 2/1/18. The fire was caused when steam leak repair injection equipment failed and leaked onto hot piping. There was no equipment damage or impact to plant operation. The fire was extinguished by the fire brigade. Offsite assistance was not required. The fire resulted from Furmanite repair of a Moisture Separator Reheater inlet flow control valve. The unusual event will be terminated when sufficient lagging is removed to verify the extent of leaked fluid. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA, DHS NICC, and NNSA (via e-mail).

• * * UPDATE AT 1240 EST ON 2/1/2018 FROM ANTHONY PETRELLI TO MARK ABRAMOVITZ * * *

The unusual event was terminated at 1211 EST. The licensee notified the NRC Resident Inspector. Notified the R1DO (Janda), NRR EO (Miller), IRD MOC (Grant), DHS SWO, FEMA, DHS NICC, and NNSA (via e-mail).

On August 20, 2017 at 1605 PDT, Columbia Generating Station was manually scrammed from 100 percent power due to a rise of Main Condenser back pressure. Manual scram of the unit is procedurally required upon a loss of Main Condenser back pressure. Preliminary investigations indicate that the Main Condenser air removal suction valve (AR-V-1) closed, resulting in the Condenser back pressure rising to within 1.0 inch Hg of the setpoint with reactor power greater than 25 percent. Further investigations continue. All control rods fully inserted. In addition to the closure of the air removal suction valve, one of two Reactor Feedwater startup flow control valves did not adequately operate to control Reactor vessel level and resulted in a high-level (Level-8) actuation tripping the Reactor Feedwater System. All other systems operated as expected. Reactor water level is currently being controlled manually with the start-up level control isolation valve. AR-V-1 has been manually opened with a jumper and temporary air supply. Reactor decay heat is being removed via bypass valves to the Main Condenser. This event is being reported under the following: 10 CFR 50.72(b)(2)(iv)(B), which requires a four-hour notification for any event or condition that results in actuation of the Reactor Protection System when the reactor is critical. The licensee notified the NRC Resident Inspector. The licensee plans to issue a press release.

• * * UPDATE ON 8/24/17 AT 1937 EDT FROM MATT HUMMER TO DONG PARK * * *

The licensee is updating the notification to include an 8 hour notification under 10 CFR 50.72(b)(3)(iv)(A) for a specified system actuation due to a Level 3 isolation signal which occurred approximately 20 minutes after the scram. The licensee is currently in cold shutdown to repair the Reactor Feedwater startup flow control valve. The licensee has notified the NRC Resident Inspector. Notified R4DO (Farnholtz).

This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented fulfillment of a safety function. On July 11th, 2017, it was discovered that the flow indicating switch for the high pressure core spray (HPCS) minimum flow valve was providing unreliable indication. There was no flow through the line at the time the condition was discovered. This switch provides the flow signal to the HPCS minimum flow valve logic. The switch was declared inoperable and the required actions of Technical Specification 3.3.5.1 were entered. This condition could have prevented the HPCS system, a single train safety system, from performing its specified safety function. Troubleshooting is underway to determine the cause of and correct the condition. The licensee notified the NRC Resident Inspector.

• * * RETRACTION FROM DAN SHARPE TO KARL DIEDERICH AT 1710 EDT ON 9/20/17 * * *

The condition reported in Event notification #52848 pursuant to 10 CFR 50.72(b)(3)(v)(D) has been evaluated, and determined not to have met the threshold for classification as an Event or Condition the Could Have Prevented Fulfillment of a Safety Function. Engineering analysis has concluded that the affected switch was capable of performing its required support function to provide the flow signal to the HPCS minimum flow valve logic. Thus, the HPCS system remained capable of performing its specific function for the identified condition. The NRC Resident Inspector has been notified. Notified R4DO (G. Miller).