Event Notification Report for July 08, 2021
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
07/07/2021 - 07/08/2021
EVENT NUMBERS
55346
55346
Power Reactor
Event Number: 55346
Facility: Davis Besse
Region: 3 State: OH
Unit: [1] [] []
RX Type: [1] B&W-R-LP
NRC Notified By: Nicholas Buehler
HQ OPS Officer: Kerby Scales
Region: 3 State: OH
Unit: [1] [] []
RX Type: [1] B&W-R-LP
NRC Notified By: Nicholas Buehler
HQ OPS Officer: Kerby Scales
Notification Date: 07/09/2021
Notification Time: 00:44 [ET]
Event Date: 07/08/2021
Event Time: 21:54 [EDT]
Last Update Date: 07/09/2021
Notification Time: 00:44 [ET]
Event Date: 07/08/2021
Event Time: 21:54 [EDT]
Last Update Date: 07/09/2021
Emergency Class: Non Emergency
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS Actuation - Critical 50.72(b)(3)(iv)(A) - Valid Specif Sys Actuation 50.72(b)(2)(iv)(A) - Eccs Injection
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS Actuation - Critical 50.72(b)(3)(iv)(A) - Valid Specif Sys Actuation 50.72(b)(2)(iv)(A) - Eccs Injection
Person (Organization):
KUNOWSKI, MICHAEL (R3)
KUNOWSKI, MICHAEL (R3)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | N | 100 | Power Operation | 0 | Hot Standby |
EN Revision Imported Date: 8/6/2021
EN Revision Text: AUTOMATIC REACTOR TRIP DUE TO MAIN TURBINE TRIP
"At 2154 EDT on 7/8/2021, with the Unit in Mode 1 at 100% power, the reactor automatically tripped due to trip of the main turbine, caused by failure of a non-safety related breaker during functional testing. Following the reactor trip the Steam Feed Rupture Control System automatically initiated on low Steam Generator 1 level, actuating both turbine-driven Auxiliary Feedwater Pumps. The operators subsequently started the high pressure injection pumps manually per procedure in response to overcooling indications.
"Operations responded and stabilized the plant. Decay heat was initially being removed via the Main Condenser. During post-trip response actions, while attempting to shut down the Auxiliary Feedwater Pumps, a low pressure condition was experienced in Steam Generator 2, resulting in isolation of the Main Condenser and steam being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage.
"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). This event is also being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a four-hour, non-emergency notification of emergency core cooling system (ECCS) discharge into the reactor coolant system, and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an eight-hour, non-emergency notification of an event that results in a valid actuation of the Auxiliary Feedwater System.
"There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
EN Revision Text: AUTOMATIC REACTOR TRIP DUE TO MAIN TURBINE TRIP
"At 2154 EDT on 7/8/2021, with the Unit in Mode 1 at 100% power, the reactor automatically tripped due to trip of the main turbine, caused by failure of a non-safety related breaker during functional testing. Following the reactor trip the Steam Feed Rupture Control System automatically initiated on low Steam Generator 1 level, actuating both turbine-driven Auxiliary Feedwater Pumps. The operators subsequently started the high pressure injection pumps manually per procedure in response to overcooling indications.
"Operations responded and stabilized the plant. Decay heat was initially being removed via the Main Condenser. During post-trip response actions, while attempting to shut down the Auxiliary Feedwater Pumps, a low pressure condition was experienced in Steam Generator 2, resulting in isolation of the Main Condenser and steam being discharged through the Atmospheric Vent Valves for decay heat removal. There is no known primary to secondary leakage.
"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). This event is also being reported in accordance with 10 CFR 50.72(b)(2)(iv)(A) as a four-hour, non-emergency notification of emergency core cooling system (ECCS) discharge into the reactor coolant system, and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an eight-hour, non-emergency notification of an event that results in a valid actuation of the Auxiliary Feedwater System.
"There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."