Event Notification Report for March 01, 2024

EN Revision Imported Date: 3/1/2024

EN Revision Text: OFFSITE NOTIFICATION - WORKPLACE INJURY

The following information was provided by the licensee via email:

"On February 1, 2024, a contract worker was transported offsite for medical treatment due to a work-related injury that required the individual to be admitted to the hospital. The individual was free-released from the site prior to transport.

"The injury and hospitalization were reported by the contract worker's employer to OSHA per 29 CFR 1904.39(a)(2). Based upon that notification to another government agency, Tennessee Valley Authority is reporting this per 10 CFR 50.72(b)(2)(xi).

"The NRC Senior Resident Inspector has been notified of this event."

* * * RETRACTION ON 2/29/24 AT 12:29 EST FROM MATTHEW SLOUKA TO KAREN COTTON * * *

The following information was provided by the licensee via email:

The purpose of this notification is to retract a previous Event Notification, EN 56941, reported on 02/01/2024.

"On 02/01/2024, at 15:32 EST, Browns Ferry Nuclear Plant (BFN) made an Event Notification 56941 notifying the NRC of a notification to another government agency. During further review of NRC reporting guidance, BFN has concluded that the contract worker's employer report to OSHA was below the reporting threshold outlined in NUREG 1022, Revision 3.

"The NRC Resident Inspector has been notified."

AUTOMATIC REACTOR SCRAM

The following information was provided by the licensee via fax and email:

"At approximately 0839 [CST] with Unit 1 in Mode 1 at 100 percent power, the reactor automatically scrammed due to the depressurization of the SCRAM air header caused by an invalid signal that [occurred] during system testing. The SCRAM was uncomplicated with all systems responding as expected. The cause and details of the event are under investigation. Containment isolation valves actuated and closed on a valid Group 2 signal.

"Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B), and an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group 2 isolation signal.

"Operations responded using the emergency operating procedure and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves.

"There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. State as well as Wright and Sherburne Counties will be notified."

The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:

The Anticipated Transient Without Scram (ATWS) circuit was being tested when an invalid signal was sent to depressurize the SCRAM air header.

SPECIFIED SYSTEM ACTUATION - AUTOMATIC START OF EMERGENCY DIESEL GENERATORS

The following information was provided by the licensee via phone and email:

"At 1350 EST on 2/28/2024, with Calvert Cliffs Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 65 percent power, an actuation of the '1A' and '2A' emergency diesel generators' auto-start occurred due to an undervoltage condition on the number 11 and number 21 4kV buses which are fed from the number 11 13kV bus. The '1A' and '2A' emergency diesel generators automatically started as designed when the 4kV buses' undervoltage signals were received.

"This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the '1A' and '2A' emergency diesel generators.

"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 undervoltage condition was caused by the feeder breaker to the number 11 13 kV bus opening during electrical maintenance.

PART 21 - SERVICE WATER BOOSTER PUMP BEARING HOUSING COVER OIL LEAK

The following information was provided by the licensee via phone and email:

"On August 27, 2022, while in Mode 1 at 100 percent power, Nebraska Public Power District (NPPD) Cooper Nuclear Station (CNS) identified a 3 drops per minute (dpm) oil leak from the radial (inboard) bearing housing cover on one of four residual heat removal (RHR) service water booster pumps (SWBP); specifically, SWBP-D.

"Subsequent analysis determined that the leak was a result of a deviation with the configuration of the labyrinth seal drain path. This deviation was due to an error introduced in a manufacturing drawing used by the vendor for the fabrication of four new replacement pumps. The bearing housing cover is a unique design to CNS for the RHR SWBPs.

"On March 1, 2024, NPPD completed a substantial safety hazard evaluation and determined that the manufacturing drawing error could cause a substantial safety hazard. The NRC Resident Senior Resident has been notified. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i). A written notification will be provided within 30 days."

The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:

All four SWBPs had a similar housing design. Three of the four pump housings have been replaced. No other nuclear plant is affected due to the housing design being unique to CNS.

INADVERTENT SIREN ACTIVATION

The following information was provided by the licensee via email:

"At 1330 CST, on March 1, 2024, an equipment vendor was coordinating with Wright County performing maintenance on an emergency siren when the county operator mistakenly sent an alarm signal instead of cancel signal, activating all Wright County emergency sirens for approximately 17 seconds. At 1345 CST, the Monticello Nuclear Generating Plant (MNGP) emergency planning coordinator received a notification from the vendor and notified the duty shift manager [of the inadvertent activation].

"Wright County officials are planning to make a public notification via social media to local residents.

"No press release by the licensee is planned at this time.

"This event is reportable per 10 CFR 50.72(b)(2)(xi), 'News Release or Notification of Other Government Agencies.' This is a 4-hour Reporting requirement.

"The NRC Resident has been notified."

AUTOMATIC REACTOR TRIP DUE TO LOSS OF MAIN FEEDWATER PUMP SUCTION

The following information was provided by the licensee via email:

"At 1142 CST on 3/3/2024, with Unit 2 in Mode 1 at 29 percent power, the reactor automatically tripped due to a turbine trip caused by a loss of suction to the 22 main feedwater pump. All systems responded normally post trip. Decay heat is being removed via the auxiliary feedwater water system. Secondary steam control mechanism is the steam generator PORVs [power operated relief valves]. Unit 1 remains at 100 percent power and is unaffected.

"This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The resident NRC inspector has been notified."

The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:

The trip occurred while the licensee was returning to power operations after a refueling outage. During the trip, all rods inserted into the core. The plant is in a normal shutdown electrical lineup with offsite power available. The plant will be maintained at normal operating temperature and pressure. There is no known primary to secondary leakage. The cause of the loss of 22 main feedwater pump suction is under investigation.

AUTOMATIC REACTOR SCRAM DUE TO MAIN TURBINE TRIP ON LOW CONDENSER VACUUM

The following information was provided by the licensee via email:
"On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram.

"Additionally, following the scram a low RPV [reactor pressure vessel] level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR [residual heat removal] shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred.

"All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system.

"This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'.

"Unit 1 is not affected. 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 cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration.