Event Notification Report for February 26, 2005
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
02/25/2005 - 02/26/2005
Power Reactor
Event Number: 41445
Facility: CALVERT CLIFFS
Region: 1 State: MD
Unit: [] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: ROBERT PACE
HQ OPS Officer: JANELLE BATTISTE
Region: 1 State: MD
Unit: [] [2] []
RX Type: [1] CE,[2] CE
NRC Notified By: ROBERT PACE
HQ OPS Officer: JANELLE BATTISTE
Notification Date: 02/26/2005
Notification Time: 20:08 [ET]
Event Date: 02/26/2005
Event Time: 15:30 [EST]
Last Update Date: 02/26/2005
Notification Time: 20:08 [ET]
Event Date: 02/26/2005
Event Time: 15:30 [EST]
Last Update Date: 02/26/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
PAMELA HENDERSON (R1)
PAMELA HENDERSON (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Cold Shutdown | 0 | Refueling |
DEGRADED CONDITION OF REACTOR COOLANT BOUNDARY WELD
"IRE-003-469 identified flaws (one circumferential and two axial indications) on 2-DR-2007 weld of hot leg drain line. Engineering has evaluated the flaws using the procedures of ASME Code Section XI, IWB-3600. Although the circumferential flaw was found to be acceptable in accordance with ASME section XI, IWB-3600, the calculated growth rates of the postulated axial flaws are large and do not meet the ASME Section XI Code requirements per NUREG 1022, Rev. 2 sect. 3.2.4 (degraded or unanalyzed condition). 'Any event or condition resulting in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' Currently a repair plan is being developed. "
The licensee indicated that they will not be able to enter mode (4) until the repair is completed (using weld overlay as the repair method) and an inspection of the repair is conducted.
The licensee will perform a design engineering evaluation of extent of condition to determine the impact of this on the operating unit (Unit 1).
The NRC Resident Inspector will be notified.
See similar event #41448
"IRE-003-469 identified flaws (one circumferential and two axial indications) on 2-DR-2007 weld of hot leg drain line. Engineering has evaluated the flaws using the procedures of ASME Code Section XI, IWB-3600. Although the circumferential flaw was found to be acceptable in accordance with ASME section XI, IWB-3600, the calculated growth rates of the postulated axial flaws are large and do not meet the ASME Section XI Code requirements per NUREG 1022, Rev. 2 sect. 3.2.4 (degraded or unanalyzed condition). 'Any event or condition resulting in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' Currently a repair plan is being developed. "
The licensee indicated that they will not be able to enter mode (4) until the repair is completed (using weld overlay as the repair method) and an inspection of the repair is conducted.
The licensee will perform a design engineering evaluation of extent of condition to determine the impact of this on the operating unit (Unit 1).
The NRC Resident Inspector will be notified.
See similar event #41448
Power Reactor
Event Number: 41446
Facility: FT CALHOUN
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] CE
NRC Notified By: RICH LOWERY
HQ OPS Officer: CHAUNCEY GOULD
Region: 4 State: NE
Unit: [1] [] []
RX Type: [1] CE
NRC Notified By: RICH LOWERY
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 02/27/2005
Notification Time: 01:26 [ET]
Event Date: 02/26/2005
Event Time: 21:34 [CST]
Last Update Date: 02/27/2005
Notification Time: 01:26 [ET]
Event Date: 02/26/2005
Event Time: 21:34 [CST]
Last Update Date: 02/27/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
DALE POWERS (R4)
DALE POWERS (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | A/R | Y | 12 | Power Operation | 0 | Hot Standby |
PLANT HAD A REACTOR TRIP FROM 16.1% POWER DUE TO A LOSS OF TURBINE LOAD SIGNAL
The following information, in addition to the phone report, was obtained from the licensee via facsimile:
"Fort Calhoun Station Unit 1 experienced a reactor trip from 16.1% Reactor Power on a Loss of Turbine Load during a scheduled plant shutdown for a refueling outage.
"The turbine generator was tripped offline at 2101 per refueling outage schedule. A feedwater transient occurred [cause being investigated] during turbine testing, resulting in power rising rapidly from 12.6% to 16.1% . At which point the reactor tripped on loss of load due to being greater than 15% power with all turbine stop valves closed.
"All plant equipment functioned as designed. A Reactor Coolant System (RCS) cooldown resulted due to a feedwater transient. Main Feedwater was isolated, which terminated the cooldown. Auxiliary Feedwater (AFW) was established with the diesel driven auxiliary feed pump. Emergency Boration was manually initiated due to the RCS cooldown and secured after shutdown margin verification. Plant is currently stable in mode 3.
"This report is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for an event that resulted in an event that resulted in an unplanned RPS actuation while the reactor was critical."
All rods fully inserted, no ECCS actuation and no relief valves lifted. Steam generators are being used for heat sink.
The NRC resident Inspector was notified
The following information, in addition to the phone report, was obtained from the licensee via facsimile:
"Fort Calhoun Station Unit 1 experienced a reactor trip from 16.1% Reactor Power on a Loss of Turbine Load during a scheduled plant shutdown for a refueling outage.
"The turbine generator was tripped offline at 2101 per refueling outage schedule. A feedwater transient occurred [cause being investigated] during turbine testing, resulting in power rising rapidly from 12.6% to 16.1% . At which point the reactor tripped on loss of load due to being greater than 15% power with all turbine stop valves closed.
"All plant equipment functioned as designed. A Reactor Coolant System (RCS) cooldown resulted due to a feedwater transient. Main Feedwater was isolated, which terminated the cooldown. Auxiliary Feedwater (AFW) was established with the diesel driven auxiliary feed pump. Emergency Boration was manually initiated due to the RCS cooldown and secured after shutdown margin verification. Plant is currently stable in mode 3.
"This report is being made pursuant to 10 CFR 50.72(b)(2)(iv)(B) for an event that resulted in an event that resulted in an unplanned RPS actuation while the reactor was critical."
All rods fully inserted, no ECCS actuation and no relief valves lifted. Steam generators are being used for heat sink.
The NRC resident Inspector was notified