Event Notification Report for September 17, 2024
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
09/16/2024 - 09/17/2024
Power Reactor
Event Number: 57330
Facility: Point Beach
Region: 3 State: WI
Unit: [1] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: Bob Murrell
HQ OPS Officer: Robert A. Thompson
Region: 3 State: WI
Unit: [1] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: Bob Murrell
HQ OPS Officer: Robert A. Thompson
Notification Date: 09/18/2024
Notification Time: 13:30 [ET]
Event Date: 09/17/2024
Event Time: 00:00 [CDT]
Last Update Date: 09/18/2024
Notification Time: 13:30 [ET]
Event Date: 09/17/2024
Event Time: 00:00 [CDT]
Last Update Date: 09/18/2024
Emergency Class: Non Emergency
10 CFR Section:
21.21(d)(3)(i) - Defects And Noncompliance
10 CFR Section:
21.21(d)(3)(i) - Defects And Noncompliance
Person (Organization):
Ziolkowski, Michael (R3DO)
Part 21/50.55 Reactors, - (EMAIL)
Ziolkowski, Michael (R3DO)
Part 21/50.55 Reactors, - (EMAIL)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
| 2 | N | Y | 96 | Power Operation | 96 | Power Operation |
PART 21 REPORT - DEFECTIVE RELAY IDENTIFIED DURING PRE-INSTALLATION TESTING
The following information was provided by NextEra Energy Point Beach, LLC (NextEra) via phone and email:
"NextEra makes the following notification under 10 CFR 21.21(d)(3)(i) of a defect found in a Westinghouse relay, model NBFD31S, during pre-installation bench testing. Specifically, the relay was found to not function as required due to its internal plunger not operating properly. This malfunctioning caused the plunger to not fully extend and cause the normally open contacts to remain closed. Investigations completed by Westinghouse determined that the plunger would not function properly because its kickout spring was misaligned due to human error. This relay was procured from Westinghouse for safety related nuclear applications.
"NextEra has concluded that this defect constitutes a substantial safety hazard (SSH). A SSH exists because of the nature of the defect was such that the relay would not be able to perform its safety function if installed, and would result in a loss of redundancy in a safety related system, in this case, the reactor protection system.
"On September 17, 2024, the Point Beach Site Vice President was notified of the requirement to report this event under 10 CFR 21.21. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i). A written notification in accordance with 10 CFR 21.21(d)(3(ii) will be provided within 30 days.
"Since this defect was discovered prior to installation, in accordance with station requirements for bench testing, and the vendor has concluded that this event is an isolated case, there were no actual impacts on safety related equipment.
"The NRC Resident Inspector has been notified."
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
Responsible corporate officer:
Michael Durbin
Site Vice President
(920) 755-7854
The following information was provided by NextEra Energy Point Beach, LLC (NextEra) via phone and email:
"NextEra makes the following notification under 10 CFR 21.21(d)(3)(i) of a defect found in a Westinghouse relay, model NBFD31S, during pre-installation bench testing. Specifically, the relay was found to not function as required due to its internal plunger not operating properly. This malfunctioning caused the plunger to not fully extend and cause the normally open contacts to remain closed. Investigations completed by Westinghouse determined that the plunger would not function properly because its kickout spring was misaligned due to human error. This relay was procured from Westinghouse for safety related nuclear applications.
"NextEra has concluded that this defect constitutes a substantial safety hazard (SSH). A SSH exists because of the nature of the defect was such that the relay would not be able to perform its safety function if installed, and would result in a loss of redundancy in a safety related system, in this case, the reactor protection system.
"On September 17, 2024, the Point Beach Site Vice President was notified of the requirement to report this event under 10 CFR 21.21. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i). A written notification in accordance with 10 CFR 21.21(d)(3(ii) will be provided within 30 days.
"Since this defect was discovered prior to installation, in accordance with station requirements for bench testing, and the vendor has concluded that this event is an isolated case, there were no actual impacts on safety related equipment.
"The NRC Resident Inspector has been notified."
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
Responsible corporate officer:
Michael Durbin
Site Vice President
(920) 755-7854
Power Reactor
Event Number: 57328
Facility: Summer
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-AP1000,[3] W-AP1000
NRC Notified By: Lauren Anderson
HQ OPS Officer: Tenisha Meadows
Region: 2 State: SC
Unit: [1] [] []
RX Type: [1] W-3-LP,[2] W-AP1000,[3] W-AP1000
NRC Notified By: Lauren Anderson
HQ OPS Officer: Tenisha Meadows
Notification Date: 09/17/2024
Notification Time: 23:38 [ET]
Event Date: 09/17/2024
Event Time: 20:05 [EDT]
Last Update Date: 09/18/2024
Notification Time: 23:38 [ET]
Event Date: 09/17/2024
Event Time: 20:05 [EDT]
Last Update Date: 09/18/2024
Emergency Class: Non Emergency
10 CFR Section:
50.72(b)(3)(v)(D) - Accident Mitigation
10 CFR Section:
50.72(b)(3)(v)(D) - Accident Mitigation
Person (Organization):
Suber, Gregory (R2DO)
Suber, Gregory (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 91 | Power Operation | 91 | Power Operation |
STEAM PROPAGATION DOOR INOPERABLE
The following information was provided by the licensee via email and phone:
"At 2005 EDT on 9/17/2024, it was discovered that steam propagation door DRCB/501 would not latch properly; thus making the door inoperable. Door DRCB/501 is required as a steam propagation barrier to protect both trains of engineered safety feature equipment from effects of a postulated steam line break. Due to this inoperability, the plant was in a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v).
"There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
"Steam propagation door DRCB/501 was repaired and maintained in the closed and latched position at 2032 EDT on 9/17/2024."
The following information was provided by the licensee via email and phone:
"At 2005 EDT on 9/17/2024, it was discovered that steam propagation door DRCB/501 would not latch properly; thus making the door inoperable. Door DRCB/501 is required as a steam propagation barrier to protect both trains of engineered safety feature equipment from effects of a postulated steam line break. Due to this inoperability, the plant was in a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v).
"There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
"Steam propagation door DRCB/501 was repaired and maintained in the closed and latched position at 2032 EDT on 9/17/2024."
Power Reactor
Event Number: 57326
Facility: Vogtle 3/4
Region: 2 State: GA
Unit: [3] [] []
RX Type: [3] W-AP1000,[4] W-AP1000
NRC Notified By: Jason Hayes
HQ OPS Officer: Sam Colvard
Region: 2 State: GA
Unit: [3] [] []
RX Type: [3] W-AP1000,[4] W-AP1000
NRC Notified By: Jason Hayes
HQ OPS Officer: Sam Colvard
Notification Date: 09/17/2024
Notification Time: 04:48 [ET]
Event Date: 09/17/2024
Event Time: 01:27 [EDT]
Last Update Date: 09/19/2024
Notification Time: 04:48 [ET]
Event Date: 09/17/2024
Event Time: 01:27 [EDT]
Last Update Date: 09/19/2024
Emergency Class: Non Emergency
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS Actuation - Critical 50.72(b)(2)(iv)(A) - ECCS Injection 50.72(b)(3)(iv)(A) - Valid Specif Sys Actuation
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS Actuation - Critical 50.72(b)(2)(iv)(A) - ECCS Injection 50.72(b)(3)(iv)(A) - Valid Specif Sys Actuation
Person (Organization):
Suber, Gregory (R2DO)
Crouch, Howard (IR)
Russell Felts (NRR EO) (NRR EO)
Suber, Gregory (R2DO)
Crouch, Howard (IR)
Russell Felts (NRR EO) (NRR EO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | A/R | Y | 100 | Power Operation | 0 | Safe Shutdown |
EN Revision Imported Date: 9/20/2024
EN Revision Text: AUTOMATIC REACTOR TRIP AND MANUAL SAFEGUARDS ACTUATION
The following information was provided by the licensee via phone and email:
"At 0127 EDT on 9/17/2024, with Unit 3 in mode 1 at 100% power, the reactor automatically tripped due to the passive residual heat removal heat exchanger outlet flow control valve failing open. A manual safeguards actuation was initiated due to the lowering pressurizer water level resulting from the reactor coolant system cooldown that was caused by the passive residual heat removal heat exchanger outlet flow control valve failing open. The trip was not complex, with all safety systems responding normally post-trip.
"Operations responded and stabilized the plant. Decay heat is being removed by the passive residual heat removal heat exchanger. Units 1, 2, and 4 are not affected.
"Due to the core makeup tank actuation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). The reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is reportable per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid containment isolation actuation and a valid passive residual heat removal heat exchanger actuation.
"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 failure of the control valve does not inhibit the residual heat removal system from functioning as it is passive. The reactor coolant system maximum allowable cooldown rate was exceeded (Technical Specification 3.4.3). The limit is 100 degrees F per hour above 350 degrees F. The maximum observed cooldown rate was 226 degrees F per hour. At time 0458 EDT, reactor coolant system temperature is 369.1 degrees F, reactor pressure is 900 psig. Currently, the plant is cooling down and proceeding toward placing shutdown cooling online.
EN Revision Text: AUTOMATIC REACTOR TRIP AND MANUAL SAFEGUARDS ACTUATION
The following information was provided by the licensee via phone and email:
"At 0127 EDT on 9/17/2024, with Unit 3 in mode 1 at 100% power, the reactor automatically tripped due to the passive residual heat removal heat exchanger outlet flow control valve failing open. A manual safeguards actuation was initiated due to the lowering pressurizer water level resulting from the reactor coolant system cooldown that was caused by the passive residual heat removal heat exchanger outlet flow control valve failing open. The trip was not complex, with all safety systems responding normally post-trip.
"Operations responded and stabilized the plant. Decay heat is being removed by the passive residual heat removal heat exchanger. Units 1, 2, and 4 are not affected.
"Due to the core makeup tank actuation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). The reactor protection system actuation while critical is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is reportable per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid containment isolation actuation and a valid passive residual heat removal heat exchanger actuation.
"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 failure of the control valve does not inhibit the residual heat removal system from functioning as it is passive. The reactor coolant system maximum allowable cooldown rate was exceeded (Technical Specification 3.4.3). The limit is 100 degrees F per hour above 350 degrees F. The maximum observed cooldown rate was 226 degrees F per hour. At time 0458 EDT, reactor coolant system temperature is 369.1 degrees F, reactor pressure is 900 psig. Currently, the plant is cooling down and proceeding toward placing shutdown cooling online.
Power Reactor
Event Number: 57345
Facility: Prairie Island
Region: 3 State: MN
Unit: [1] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: Timothy Thomas
HQ OPS Officer: Ernest West
Region: 3 State: MN
Unit: [1] [2] []
RX Type: [1] W-2-LP,[2] W-2-LP
NRC Notified By: Timothy Thomas
HQ OPS Officer: Ernest West
Notification Date: 09/27/2024
Notification Time: 11:43 [ET]
Event Date: 09/17/2024
Event Time: 17:00 [CDT]
Last Update Date: 09/27/2024
Notification Time: 11:43 [ET]
Event Date: 09/17/2024
Event Time: 17:00 [CDT]
Last Update Date: 09/27/2024
Emergency Class: Non Emergency
10 CFR Section:
26.719 - Fitness For Duty
10 CFR Section:
26.719 - Fitness For Duty
Person (Organization):
Havertape, Joshua (R3DO)
FFD Group, (EMAIL)
Havertape, Joshua (R3DO)
FFD Group, (EMAIL)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 0 | Refueling |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
FITNESS FOR DUTY
The following information was provided by the licensee via email:
"On September 17, 2024, the site identified that an individual assigned to perform fitness for duty (FFD) program duties, who should have been part of the fitness for duty program random testing pool, had been inadvertently removed during a recent computer system upgrade. The individual was reprocessed and placed back into the FFD program on September 18, 2024. This was determined to be an isolated incident as it was confirmed that no other individuals required to be in the program were removed."
The following information was provided by the licensee via email:
"On September 17, 2024, the site identified that an individual assigned to perform fitness for duty (FFD) program duties, who should have been part of the fitness for duty program random testing pool, had been inadvertently removed during a recent computer system upgrade. The individual was reprocessed and placed back into the FFD program on September 18, 2024. This was determined to be an isolated incident as it was confirmed that no other individuals required to be in the program were removed."
Power Reactor
Event Number: 57346
Facility: Monticello
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: Brandon Kent
HQ OPS Officer: Ernest West
Region: 3 State: MN
Unit: [1] [] []
RX Type: [1] GE-3
NRC Notified By: Brandon Kent
HQ OPS Officer: Ernest West
Notification Date: 09/27/2024
Notification Time: 12:10 [ET]
Event Date: 09/17/2024
Event Time: 17:00 [CDT]
Last Update Date: 09/27/2024
Notification Time: 12:10 [ET]
Event Date: 09/17/2024
Event Time: 17:00 [CDT]
Last Update Date: 09/27/2024
Emergency Class: Non Emergency
10 CFR Section:
26.719 - Fitness For Duty
10 CFR Section:
26.719 - Fitness For Duty
Person (Organization):
Havertape, Joshua (R3DO)
FFD Group, (EMAIL)
Havertape, Joshua (R3DO)
FFD Group, (EMAIL)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
FITNESS FOR DUTY
The following information was provided by the licensee via email:
"On September 17, 2024, the site identified that an individual assigned to perform fitness for duty (FFD) program duties, who should have been part of the fitness for duty program random testing pool, had been inadvertently removed during a recent computer system upgrade. The individual was reprocessed and placed back into the FFD program on September 18, 2024. This was determined to be an isolated incident as it was confirmed that no other individuals required to be in the program were removed."
The following information was provided by the licensee via email:
"On September 17, 2024, the site identified that an individual assigned to perform fitness for duty (FFD) program duties, who should have been part of the fitness for duty program random testing pool, had been inadvertently removed during a recent computer system upgrade. The individual was reprocessed and placed back into the FFD program on September 18, 2024. This was determined to be an isolated incident as it was confirmed that no other individuals required to be in the program were removed."