Event Notification Report for August 03, 2019
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
08/02/2019 - 08/03/2019
Power Reactor
Event Number: 54198
Facility: HOPE CREEK
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: STEVEN BIER
HQ OPS Officer: HOWIE CROUCH
Region: 1 State: NJ
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: STEVEN BIER
HQ OPS Officer: HOWIE CROUCH
Notification Date: 08/03/2019
Notification Time: 23:33 [ET]
Event Date: 08/03/2019
Event Time: 19:47 [EDT]
Last Update Date: 08/04/2019
Notification Time: 23:33 [ET]
Event Date: 08/03/2019
Event Time: 19:47 [EDT]
Last Update Date: 08/04/2019
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 |
|---|---|---|---|---|---|---|
| 1 | M/R | Y | 37 | Power Operation | 0 | Hot Shutdown |
MANUAL REACTOR TRIP AND MANUAL ACTUATION OF REACTOR CORE ISOLATION COOLING
"At 1947 [EDT] on 8/3/19, with Hope Creek in Mode 1 at 37 percent power, the reactor was manually scrammed due to loss of condenser vacuum. All control rods fully inserted into the core. All safety systems responded as designed and expected. Reactor level was stabilized using Reactor Core Isolation Cooling (RCIC) and Reactor Feedwater Pumps.
"Currently reactor water level is being maintained by the feedwater system and decay heat is being removed by the main condenser using the main turbine bypass valves. 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). Due to the manual actuation of RCIC, this event is also being reported as an eight-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."
The plant is in its normal shutdown electrical lineup with all safe shutdown equipment available. The licensee will be notifying the state of Delaware, state of New Jersey and the Lower Alloway Creek township.
"At 1947 [EDT] on 8/3/19, with Hope Creek in Mode 1 at 37 percent power, the reactor was manually scrammed due to loss of condenser vacuum. All control rods fully inserted into the core. All safety systems responded as designed and expected. Reactor level was stabilized using Reactor Core Isolation Cooling (RCIC) and Reactor Feedwater Pumps.
"Currently reactor water level is being maintained by the feedwater system and decay heat is being removed by the main condenser using the main turbine bypass valves. 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). Due to the manual actuation of RCIC, this event is also being reported as an eight-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."
The plant is in its normal shutdown electrical lineup with all safe shutdown equipment available. The licensee will be notifying the state of Delaware, state of New Jersey and the Lower Alloway Creek township.
Power Reactor
Event Number: 54197
Facility: CLINTON
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: DAVID LIVINGSTON
HQ OPS Officer: HOWIE CROUCH
Region: 3 State: IL
Unit: [1] [] []
RX Type: [1] GE-6
NRC Notified By: DAVID LIVINGSTON
HQ OPS Officer: HOWIE CROUCH
Notification Date: 08/03/2019
Notification Time: 06:47 [ET]
Event Date: 08/03/2019
Event Time: 02:26 [CDT]
Last Update Date: 08/06/2019
Notification Time: 06:47 [ET]
Event Date: 08/03/2019
Event Time: 02:26 [CDT]
Last Update Date: 08/06/2019
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
10 CFR Section:
50.72(b)(2)(iv)(A) - ECCS INJECTION 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION 50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
PATRICIA PELKE (R3DO)
PATRICIA PELKE (R3DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 98 | Power Operation | 0 | Hot Shutdown |
AUTOMATIC REACTOR SCRAM ON LOW REACTOR WATER LEVEL
"At 0226 [CDT], an automatic scram on low reactor water level occurred due to a trip of the 'B' Reactor Feed pump. All control rods fully inserted. Reactor water level 2 was reached and the High Pressure Core Spray system, Reactor Core Isolation Cooling system, Division 3 diesel generator, Standby Gas Treatment Systems 'A' and 'B' and all shutdown safety related service water pumps started as expected. Reactor Core Isolation Cooling and High Pressure Core Spray injected as expected. All level 2 containment isolation signals occurred as expected and all level 2 containment valves closed as expected. Reactor water level is currently being controlled in band by condensate. Reactor pressure is being maintained by main turbine Bypass Valves.
"This event is being reported under 10 CFR 50.72(b)(2)(iv)(A), for ECCS discharge to RCS; 10 CFR 50.72(b)(2)(iv)(B), for RPS actuation, and 10 CFR 50.72(b)(3)(iv)(A), for specified system actuation.
"The NRC Senior Resident Inspector has been notified."
No safety relief valves lifted during the transient. The plant is in a normal shutdown electrical lineup with all safety equipment available. The licensee notified the Illinois Emergency Management Agency per their communications protocol.
* * * UPDATE FROM DAVID LIVINGSTON TO HOWIE CROUCH AT 0321 EDT ON 8/4/19 * * *
"Following automatic initiation of the High Pressure Core Spray (HPCS) System as described above, the HPCS System was manually secured following station procedures after verification that additional RPV [reactor pressure vessel] injection was no longer required. Securing HPCS injection in this manner prevents automatic restart of the system in the event of a subsequent low RPV level condition, rendering it inoperable. As the HPCS system is considered a single train safety system, this meets the reportability requirements of 10 CFR 50.72(b)(3)(v)(D). This reportable condition was identified following review of post-scram actions. The HPCS system has been restored to a Standby lineup."
The licensee will be notifying the NRC Resident Inspector.
Notified R3DO (Pelke).
* * * UPDATE FROM JAMES FORMAN TO KERBY SCALES AT 1545 EDT ON 8/6/19 * * *
"Following the scram, the Primary Containment to Secondary Containment and the Drywell to Primary Containment differential pressure limits were exceeded. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.4, Primary Containment Pressure, and 3.6.5.4, Drywell Pressure, Actions A.1, B.1, and B.2 were entered. Primary Containment to Secondary Containment differential pressure and Drywell to Primary Containment differential pressure were restored to within the LCO limits at 1505 on 8/3/19 and the associated TS Actions were exited. This event is reportable under 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that could have prevented the fulfillment of the primary containment function due to being outside the initial conditions to ensure that drywell and containment pressures remain within design values during a loss of coolant accident. This event is also reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of the drywell and primary containment functions to control the release of radioactive material for the same reason."
The licensee notified the NRC Resident Inspector.
Notified R3DO (Pelke).
"At 0226 [CDT], an automatic scram on low reactor water level occurred due to a trip of the 'B' Reactor Feed pump. All control rods fully inserted. Reactor water level 2 was reached and the High Pressure Core Spray system, Reactor Core Isolation Cooling system, Division 3 diesel generator, Standby Gas Treatment Systems 'A' and 'B' and all shutdown safety related service water pumps started as expected. Reactor Core Isolation Cooling and High Pressure Core Spray injected as expected. All level 2 containment isolation signals occurred as expected and all level 2 containment valves closed as expected. Reactor water level is currently being controlled in band by condensate. Reactor pressure is being maintained by main turbine Bypass Valves.
"This event is being reported under 10 CFR 50.72(b)(2)(iv)(A), for ECCS discharge to RCS; 10 CFR 50.72(b)(2)(iv)(B), for RPS actuation, and 10 CFR 50.72(b)(3)(iv)(A), for specified system actuation.
"The NRC Senior Resident Inspector has been notified."
No safety relief valves lifted during the transient. The plant is in a normal shutdown electrical lineup with all safety equipment available. The licensee notified the Illinois Emergency Management Agency per their communications protocol.
* * * UPDATE FROM DAVID LIVINGSTON TO HOWIE CROUCH AT 0321 EDT ON 8/4/19 * * *
"Following automatic initiation of the High Pressure Core Spray (HPCS) System as described above, the HPCS System was manually secured following station procedures after verification that additional RPV [reactor pressure vessel] injection was no longer required. Securing HPCS injection in this manner prevents automatic restart of the system in the event of a subsequent low RPV level condition, rendering it inoperable. As the HPCS system is considered a single train safety system, this meets the reportability requirements of 10 CFR 50.72(b)(3)(v)(D). This reportable condition was identified following review of post-scram actions. The HPCS system has been restored to a Standby lineup."
The licensee will be notifying the NRC Resident Inspector.
Notified R3DO (Pelke).
* * * UPDATE FROM JAMES FORMAN TO KERBY SCALES AT 1545 EDT ON 8/6/19 * * *
"Following the scram, the Primary Containment to Secondary Containment and the Drywell to Primary Containment differential pressure limits were exceeded. Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.4, Primary Containment Pressure, and 3.6.5.4, Drywell Pressure, Actions A.1, B.1, and B.2 were entered. Primary Containment to Secondary Containment differential pressure and Drywell to Primary Containment differential pressure were restored to within the LCO limits at 1505 on 8/3/19 and the associated TS Actions were exited. This event is reportable under 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that could have prevented the fulfillment of the primary containment function due to being outside the initial conditions to ensure that drywell and containment pressures remain within design values during a loss of coolant accident. This event is also reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of the drywell and primary containment functions to control the release of radioactive material for the same reason."
The licensee notified the NRC Resident Inspector.
Notified R3DO (Pelke).