Event Notification Report for October 26, 2017
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/25/2017 - 10/26/2017
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 53038
Facility: GRAND GULF
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: DAVID W. BURRUS
HQ OPS Officer: HOWIE CROUCH
Region: 4 State: MS
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: DAVID W. BURRUS
HQ OPS Officer: HOWIE CROUCH
Notification Date: 10/26/2017
Notification Time: 00:00 [ET]
Event Date: 10/26/2017
Event Time: 00:00 [CDT]
Last Update Date: 12/15/2017
Notification Time: 00:00 [ET]
Event Date: 10/26/2017
Event Time: 00:00 [CDT]
Last Update Date: 12/15/2017
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):
RAY AZUA (R4DO)
RAY AZUA (R4DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 100 | Power Operation | 100 | Power Operation |
BOTH DIVISIONS OF DRYWELL VACUUM RELIEF DECLARED INOPERABLE
"At 1055 [CDT], drywell purge supply/initial vacuum relief 1E61F003B was declared INOPERABLE for Drywell Vacuum Relief System while performing a monthly surveillance. 1E61F003B is a Division II powered valve. Division 1 Emergency Diesel Generator is INOPERABLE due to a tagout. At 1455, under LCO 3.8.1.B.2 the station declared both divisions of Drywell Vacuum Relief INOPERABLE. GGNS [Grand Gulf Nuclear Station] identified that a loss of Safety Function occurred due to a loss of two 10-inch vacuum relief lines from the Drywell required by Technical Specification 3.6.5.6 and therefore could have prevented fulfillment of its safety function. That event is being reported under 10CFR50.72(b)(3)(v)(D), as any event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.
"
Both divisions inoperable has placed the plant in a 72-hr. LCO shutdown action statement.
The licensee has notified the NRC Resident Inspector.
* * * RETRACTION FROM RALPH FLICKINGER TO HOWIE CROUCH AT 1643 EST ON 12/15/17 * * *
"At 1055 [CDT on 10/26/17], Drywell purge supply/initial vacuum relief 1E61F003B was declared INOPERABLE for Drywell Vacuum Relief System while performing a monthly surveillance. 1E61F003B is a Division II powered valve. Division 1 Emergency Diesel Generator was INOPERABLE due to a tagout. At 1455 [CDT on 12/26/17], under LCO 3.8.1.B.2, the station declared both divisions of Drywell Vacuum Relief INOPERABLE. GGNS identified that a loss of Safety Function occurred due to a loss of two 10-inch vacuum relief lines from the Drywell required by Technical Specification 3.6.5.6 and therefore could have prevented fulfillment of its safety function.
"This was initially reported under 10 CFR 50.72(b)(2)(v)(D). Div. 1 EDG was initially taken out of service at 1455 on 10/26 for preplanned maintenance (OP-EVAL). It was subsequently declared INOPERABLE-INOP due to a visible flaw indication in the exhaust manifold (1117 on 10/27). A subsequent Maintenance Functional Failure Evaluation and Past Operability Determination concluded the diesel was capable of performing its intended function for the required mission time, and therefore met the definition of OPERABLE.
"NUREG-1022 provides clarification for 10 CFR 50.72(b)(2)(v). NUREG Section 3.2.7, paragraph 4, states '...unless a condition is discovered that would have resulted in the system [otherwise] being declared inoperable, reports are not required when systems are declared inoperable solely as a result of Required Actions for which the bases is the assumption of an additional random single failure (i.e., . ..LCO 3.8.1, 'AC Sources Operating,' Required Actions .., B.2, or C.1). Per ACTION 3.8.1 .B.2, both trains of Drywell vacuum and Drywell Purge were inoperable for the purposes of Tech Specs. However, the normal power supply was available to Division 1 and there were no conditions which would have rendered the Division 1 diesel inoperable. Therefore, per Section 3.2.7 of NUREG-1022, this was not a Loss of Safety Function and was not reportable under 50.72(b)(2)(v)(D)."
The licensee has notified the NRC Resident Inspector.
Notified R4DO (Deese).
"At 1055 [CDT], drywell purge supply/initial vacuum relief 1E61F003B was declared INOPERABLE for Drywell Vacuum Relief System while performing a monthly surveillance. 1E61F003B is a Division II powered valve. Division 1 Emergency Diesel Generator is INOPERABLE due to a tagout. At 1455, under LCO 3.8.1.B.2 the station declared both divisions of Drywell Vacuum Relief INOPERABLE. GGNS [Grand Gulf Nuclear Station] identified that a loss of Safety Function occurred due to a loss of two 10-inch vacuum relief lines from the Drywell required by Technical Specification 3.6.5.6 and therefore could have prevented fulfillment of its safety function. That event is being reported under 10CFR50.72(b)(3)(v)(D), as any event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.
"
Both divisions inoperable has placed the plant in a 72-hr. LCO shutdown action statement.
The licensee has notified the NRC Resident Inspector.
* * * RETRACTION FROM RALPH FLICKINGER TO HOWIE CROUCH AT 1643 EST ON 12/15/17 * * *
"At 1055 [CDT on 10/26/17], Drywell purge supply/initial vacuum relief 1E61F003B was declared INOPERABLE for Drywell Vacuum Relief System while performing a monthly surveillance. 1E61F003B is a Division II powered valve. Division 1 Emergency Diesel Generator was INOPERABLE due to a tagout. At 1455 [CDT on 12/26/17], under LCO 3.8.1.B.2, the station declared both divisions of Drywell Vacuum Relief INOPERABLE. GGNS identified that a loss of Safety Function occurred due to a loss of two 10-inch vacuum relief lines from the Drywell required by Technical Specification 3.6.5.6 and therefore could have prevented fulfillment of its safety function.
"This was initially reported under 10 CFR 50.72(b)(2)(v)(D). Div. 1 EDG was initially taken out of service at 1455 on 10/26 for preplanned maintenance (OP-EVAL). It was subsequently declared INOPERABLE-INOP due to a visible flaw indication in the exhaust manifold (1117 on 10/27). A subsequent Maintenance Functional Failure Evaluation and Past Operability Determination concluded the diesel was capable of performing its intended function for the required mission time, and therefore met the definition of OPERABLE.
"NUREG-1022 provides clarification for 10 CFR 50.72(b)(2)(v). NUREG Section 3.2.7, paragraph 4, states '...unless a condition is discovered that would have resulted in the system [otherwise] being declared inoperable, reports are not required when systems are declared inoperable solely as a result of Required Actions for which the bases is the assumption of an additional random single failure (i.e., . ..LCO 3.8.1, 'AC Sources Operating,' Required Actions .., B.2, or C.1). Per ACTION 3.8.1 .B.2, both trains of Drywell vacuum and Drywell Purge were inoperable for the purposes of Tech Specs. However, the normal power supply was available to Division 1 and there were no conditions which would have rendered the Division 1 diesel inoperable. Therefore, per Section 3.2.7 of NUREG-1022, this was not a Loss of Safety Function and was not reportable under 50.72(b)(2)(v)(D)."
The licensee has notified the NRC Resident Inspector.
Notified R4DO (Deese).
Power Reactor
Event Number: 53036
Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: CHARLES PIKE
HQ OPS Officer: BETHANY CECERE
Region: 2 State: FL
Unit: [] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: CHARLES PIKE
HQ OPS Officer: BETHANY CECERE
Notification Date: 10/26/2017
Notification Time: 05:54 [ET]
Event Date: 10/26/2017
Event Time: 02:12 [EDT]
Last Update Date: 10/26/2017
Notification Time: 05:54 [ET]
Event Date: 10/26/2017
Event Time: 02:12 [EDT]
Last Update Date: 10/26/2017
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):
SHANE SANDAL (R2DO)
SHANE SANDAL (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | A/R | Y | 100 | Power Operation | 0 | Hot Standby |
AUTOMATIC REACTOR TRIP FOLLOWING A LOSS OF LOAD
"On October 26, 2017 at 0212 EDT St. Lucie Unit 2 experienced a reactor trip due to a loss of load event resulting in an RPS [Reactor Protection System] actuation. The cause of the loss of load is currently under investigation. Following the reactor trip, an Auxiliary Feedwater Actuation Signal occurred due to low level in the 2A Steam Generator. One of the two Main Feed Isolation Valves to the 2A Steam Generator did not close on the Auxiliary Feedwater Actuation Signal. 2A Steam Generator level was restored by Auxiliary Feedwater. The 2B Steam Generator level is being maintained by Main Feedwater. All CEAs [Control Element Assemblies] fully inserted into the core. Decay heat removal is being accomplished through forced circulation with stable conditions from Auxiliary Feedwater/Main Feedwater and Steam Bypass Control System. Currently maintaining pressurizer pressure at 2250 psia and Reactor Coolant System temperature at 532 degrees F.
"St. Lucie Unit 1 was unaffected and remains in Mode 1 at 100 percent power.
"This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72(b)(3)(iv)(A) for the Specified System Actuation."
The licensee notified the NRC Resident Inspector.
"On October 26, 2017 at 0212 EDT St. Lucie Unit 2 experienced a reactor trip due to a loss of load event resulting in an RPS [Reactor Protection System] actuation. The cause of the loss of load is currently under investigation. Following the reactor trip, an Auxiliary Feedwater Actuation Signal occurred due to low level in the 2A Steam Generator. One of the two Main Feed Isolation Valves to the 2A Steam Generator did not close on the Auxiliary Feedwater Actuation Signal. 2A Steam Generator level was restored by Auxiliary Feedwater. The 2B Steam Generator level is being maintained by Main Feedwater. All CEAs [Control Element Assemblies] fully inserted into the core. Decay heat removal is being accomplished through forced circulation with stable conditions from Auxiliary Feedwater/Main Feedwater and Steam Bypass Control System. Currently maintaining pressurizer pressure at 2250 psia and Reactor Coolant System temperature at 532 degrees F.
"St. Lucie Unit 1 was unaffected and remains in Mode 1 at 100 percent power.
"This report is submitted in accordance with 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip and 10 CFR 50.72(b)(3)(iv)(A) for the Specified System Actuation."
The licensee notified the NRC Resident Inspector.