Event Notification Report for February 16, 2018
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
02/15/2018 - 02/16/2018
Power Reactor
Event Number: 53216
Facility: INDIAN POINT
Region: 1 State: NY
Unit: [] [3] []
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: JOHN DIGNAM
HQ OPS Officer: JEFF HERRERA
Region: 1 State: NY
Unit: [] [3] []
RX Type: [2] W-4-LP,[3] W-4-LP
NRC Notified By: JOHN DIGNAM
HQ OPS Officer: JEFF HERRERA
Notification Date: 02/16/2018
Notification Time: 04:43 [ET]
Event Date: 02/16/2018
Event Time: 02:01 [EST]
Last Update Date: 02/16/2018
Notification Time: 04:43 [ET]
Event Date: 02/16/2018
Event Time: 02:01 [EST]
Last Update Date: 02/16/2018
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
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
FRED BOWER (R1DO)
FRED BOWER (R1DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 3 | A/R | Y | 100 | Power Operation | 0 | Hot Standby |
AUTOMATIC TRIP DUE TO LOSS OF EXCITATION ON MAIN GENERATOR
"On February 16, 2018 at 02:01 EST Indian Point Unit 3 automatically tripped on a turbine trip due to a loss of main generator excitation. All control rods fully inserted and all plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv)(B).
"The auxiliary feedwater system actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. The plant is stable, in Mode 3, at no load operating temperature and pressure. Decay heat removal is via the steam generators to the main condenser via the condenser steam dumps. No radiation was released.
"Indian Point Unit 2 was unaffected by this event and remains at 100 percent power.
"A post trip investigation is in progress.
"The licensee has notified the NRC Resident Inspector."
"On February 16, 2018 at 02:01 EST Indian Point Unit 3 automatically tripped on a turbine trip due to a loss of main generator excitation. All control rods fully inserted and all plant equipment responded normally to the unit trip. This is reportable under 10 CFR 50.72(b)(2)(iv)(B).
"The auxiliary feedwater system actuated following the automatic trip as expected. This is reportable under 10 CFR 50.72(b)(3)(iv)(A). During the event offsite power remained available and stable. No primary or secondary reliefs lifted. The plant is stable, in Mode 3, at no load operating temperature and pressure. Decay heat removal is via the steam generators to the main condenser via the condenser steam dumps. No radiation was released.
"Indian Point Unit 2 was unaffected by this event and remains at 100 percent power.
"A post trip investigation is in progress.
"The licensee has notified the NRC Resident Inspector."
Power Reactor
Event Number: 53217
Facility: MCGUIRE
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BRIAN JETTON
HQ OPS Officer: DONG HWA PARK
Region: 2 State: NC
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: BRIAN JETTON
HQ OPS Officer: DONG HWA PARK
Notification Date: 02/16/2018
Notification Time: 13:58 [ET]
Event Date: 02/16/2018
Event Time: 10:14 [EST]
Last Update Date: 02/16/2018
Notification Time: 13:58 [ET]
Event Date: 02/16/2018
Event Time: 10:14 [EST]
Last Update Date: 02/16/2018
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
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
OMAR LOPEZ (R2DO)
OMAR LOPEZ (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 100 | Power Operation | 0 | Hot Standby |
AUTOMATIC REACTOR TRIP DURING SOLID STATE PROTECTION SYSTEM TESTING
"At 1014 [EST] hours on 2/16/18, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped when the Reactor Trip Breakers opened during Train B Solid State Protection System (SSPS) testing. The trip was uncomplicated with all systems responding normally post-trip. Operations manually started the motor driven auxiliary feedwater
pumps. The turbine driven auxiliary feedwater pump (TDCAP) auto-started on low steam generator level. A Feedwater Isolation occurred as designed due to the Reactor Trip and Lo Tave condition. Operations stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected.
"Due to the Reactor Protection System actuation while critical, actuation of the TDCAP and motor driven auxiliary feedwater pumps along with the Feedwater Isolation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A).
"There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. "
"At 1014 [EST] hours on 2/16/18, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped when the Reactor Trip Breakers opened during Train B Solid State Protection System (SSPS) testing. The trip was uncomplicated with all systems responding normally post-trip. Operations manually started the motor driven auxiliary feedwater
pumps. The turbine driven auxiliary feedwater pump (TDCAP) auto-started on low steam generator level. A Feedwater Isolation occurred as designed due to the Reactor Trip and Lo Tave condition. Operations stabilized the plant. Decay heat is being removed by the condenser. Unit 2 is not affected.
"Due to the Reactor Protection System actuation while critical, actuation of the TDCAP and motor driven auxiliary feedwater pumps along with the Feedwater Isolation, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A).
"There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. "
Non-Power Reactor
Event Number: 53222
Rep Org: UNIV OF FLORIDA
Licensee: UNIVERSITY OF FLORIDA
Region: 0
City: GAINESVILLE State: FL
County: ALACHUA
License #: R-56
Agreement: Y
Docket: 05000083
NRC Notified By: DAN CRONIN
HQ OPS Officer: DAVID AIRD
Licensee: UNIVERSITY OF FLORIDA
Region: 0
City: GAINESVILLE State: FL
County: ALACHUA
License #: R-56
Agreement: Y
Docket: 05000083
NRC Notified By: DAN CRONIN
HQ OPS Officer: DAVID AIRD
Notification Date: 02/20/2018
Notification Time: 16:45 [ET]
Event Date: 02/16/2018
Event Time: 08:55 [EST]
Last Update Date: 02/21/2018
Notification Time: 16:45 [ET]
Event Date: 02/16/2018
Event Time: 08:55 [EST]
Last Update Date: 02/21/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
10 CFR Section:
Person (Organization):
DUANE HARDESTY (NRR)
ELIZABETH REED (NRR)
DUANE HARDESTY (NRR)
ELIZABETH REED (NRR)
TECHNICAL SPECIFICATIONS VIOLATION DUE TO INCORRECTLY CALIBRATED REACTOR TRIP INSTRUMENTATION
"On 2/16/18, the calorimetric portion of Surveillance Requirement (SR) 3.2.3.3 for Limiting Condition for Operations (LCO) 3.2.3.1 was performed in accordance with University of Florida Test Reactor (UFTR) procedures. Reactor power was stabilized at an indicated power of approximately 94 percent with actual thermal power (calculated by heat balance) at approximately 98kW (98 percent RTP) [rated thermal power]. Indicated power was then adjusted to approximately 99 percent to ensure adequate margin to the license limit of 100kW thermal steady-state (100 percent RTP). At no time was actual reactor power allowed to exceed the license limit. The reactor was then shutdown and secured to perform post-calorimetric verifications. SR 3.2.2.1 and 3.2.3.1 were performed satisfactorily just prior to operation to verify the High Reactor Power trip function and setpoint. Reactor Power Channel 1 was set to actuate at 110 percent and Reactor Power Channel 2 was set to actuate at 108 percent.
"At about 1400 on 2/19/18, during review of surveillance parameters and discussion with the operators, UFTR management determined that Limited Safety System Setting 2.2.1 and LCO 3.2.2.1 had been violated. The Technical Specifications allowable value for the automatic High Reactor Power trip function is less than or equal to 110 percent RTP. There was a period of approximately 5 hours of operation prior to adjustment of the power channels, however, during which actual thermal power was greater than indicated power (but less than 100kW thermal). During this period, if a power excursion had occurred, the High Reactor Power trip would have actuated at approximately 114 percent RTP. This condition is reportable in accordance with Sections 6.7.2.a.3 and 6.7.2.a.4 of the UFTR Technical Specifications. As described above, this condition was corrected by adjusting indicated power on both reactor power channels to greater than actual thermal power.
"The NRC Project Manager, assigned inspector, and Non-Power Reactor Licensing Branch Chief (acting) have been notified."
"On 2/16/18, the calorimetric portion of Surveillance Requirement (SR) 3.2.3.3 for Limiting Condition for Operations (LCO) 3.2.3.1 was performed in accordance with University of Florida Test Reactor (UFTR) procedures. Reactor power was stabilized at an indicated power of approximately 94 percent with actual thermal power (calculated by heat balance) at approximately 98kW (98 percent RTP) [rated thermal power]. Indicated power was then adjusted to approximately 99 percent to ensure adequate margin to the license limit of 100kW thermal steady-state (100 percent RTP). At no time was actual reactor power allowed to exceed the license limit. The reactor was then shutdown and secured to perform post-calorimetric verifications. SR 3.2.2.1 and 3.2.3.1 were performed satisfactorily just prior to operation to verify the High Reactor Power trip function and setpoint. Reactor Power Channel 1 was set to actuate at 110 percent and Reactor Power Channel 2 was set to actuate at 108 percent.
"At about 1400 on 2/19/18, during review of surveillance parameters and discussion with the operators, UFTR management determined that Limited Safety System Setting 2.2.1 and LCO 3.2.2.1 had been violated. The Technical Specifications allowable value for the automatic High Reactor Power trip function is less than or equal to 110 percent RTP. There was a period of approximately 5 hours of operation prior to adjustment of the power channels, however, during which actual thermal power was greater than indicated power (but less than 100kW thermal). During this period, if a power excursion had occurred, the High Reactor Power trip would have actuated at approximately 114 percent RTP. This condition is reportable in accordance with Sections 6.7.2.a.3 and 6.7.2.a.4 of the UFTR Technical Specifications. As described above, this condition was corrected by adjusting indicated power on both reactor power channels to greater than actual thermal power.
"The NRC Project Manager, assigned inspector, and Non-Power Reactor Licensing Branch Chief (acting) have been notified."