Event Notification Report for March 02, 2010
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/01/2010 - 03/02/2010
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 45737
Facility: CALVERT CLIFFS
Region: 1 State: MD
Unit: [1] [] []
RX Type: [1] CE,[2] CE
NRC Notified By: CHARLES MORGAN
HQ OPS Officer: MARK ABRAMOVITZ
Region: 1 State: MD
Unit: [1] [] []
RX Type: [1] CE,[2] CE
NRC Notified By: CHARLES MORGAN
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 03/02/2010
Notification Time: 11:30 [ET]
Event Date: 03/02/2010
Event Time: 07:30 [EST]
Last Update Date: 03/15/2010
Notification Time: 11:30 [ET]
Event Date: 03/02/2010
Event Time: 07:30 [EST]
Last Update Date: 03/15/2010
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):
JOHN WHITE (R1DO)
JOHN WHITE (R1DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
PRESSURIZER SAFETY NOZZLE WELD DEGRADATION
"A Unit-1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld, (#4SR-1006-1) was discovered to have an indication not acceptable under ASME Section XI, IWB-3600. The weld is on a 4 inch diameter line to the pressurizer safety relief valve (RV-201). The weld is in our dissimilar metal weld inspection program and was scheduled to be examined this RFO [refueling outage] in accordance with MRP-139 requirements. Initial inspections on 3/1/10 found an indication and the indication was confirmed by a second NDE level 3 examiner. Indication is circumferential, initiated at ID and propagates approximately 1.8 inches circumferentially to about approximately 70% through wall. ASME IWB-3500 allows up to 12.5% through wall for class 1 welds. The defect was found using phased array Ultrasonic Testing (UT). The NDE report is being reviewed and characterized at this time.
"At this time, we believe the most probable cause of the indication is primary water stress corrosion cracking. No active leaks have been identified, the defect is not through wall.
"Unit 1 is in Mode 6. A team has been established and repair options are being investigated."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION PROVIDED AT 0922 EDT ON 03/15/2010 BY ROBERT MARTIN TO JEFF ROTTON * * *
"On March 7, 2010, manually indexed phased-array ultrasonic examination of the U1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld was performed. Based on these inspections, it was determined on March 13, 2010 that the indication does not exhibit stress corrosion cracking characteristics and is not consistent with ultrasonic responses associated with inside diameter (ID) connected geometry. Therefore, the U1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld is acceptable from an ASME Code Section XI perspective. On this basis, the indication did not result in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded and the issue is not reportable under 10 CFR 50.72(b)(3)(ii)(A).
"The NRC resident was notified of this retraction."
Notified the R1DO (Chris Cahill)
"A Unit-1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld, (#4SR-1006-1) was discovered to have an indication not acceptable under ASME Section XI, IWB-3600. The weld is on a 4 inch diameter line to the pressurizer safety relief valve (RV-201). The weld is in our dissimilar metal weld inspection program and was scheduled to be examined this RFO [refueling outage] in accordance with MRP-139 requirements. Initial inspections on 3/1/10 found an indication and the indication was confirmed by a second NDE level 3 examiner. Indication is circumferential, initiated at ID and propagates approximately 1.8 inches circumferentially to about approximately 70% through wall. ASME IWB-3500 allows up to 12.5% through wall for class 1 welds. The defect was found using phased array Ultrasonic Testing (UT). The NDE report is being reviewed and characterized at this time.
"At this time, we believe the most probable cause of the indication is primary water stress corrosion cracking. No active leaks have been identified, the defect is not through wall.
"Unit 1 is in Mode 6. A team has been established and repair options are being investigated."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION PROVIDED AT 0922 EDT ON 03/15/2010 BY ROBERT MARTIN TO JEFF ROTTON * * *
"On March 7, 2010, manually indexed phased-array ultrasonic examination of the U1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld was performed. Based on these inspections, it was determined on March 13, 2010 that the indication does not exhibit stress corrosion cracking characteristics and is not consistent with ultrasonic responses associated with inside diameter (ID) connected geometry. Therefore, the U1 Pressurizer Safety Relief Valve Nozzle to Safe-end Weld is acceptable from an ASME Code Section XI perspective. On this basis, the indication did not result in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded and the issue is not reportable under 10 CFR 50.72(b)(3)(ii)(A).
"The NRC resident was notified of this retraction."
Notified the R1DO (Chris Cahill)
Power Reactor
Event Number: 45739
Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: LANCE LANE
HQ OPS Officer: DONG HWA PARK
Region: 4 State: KS
Unit: [1] [] []
RX Type: [1] W-4-LP
NRC Notified By: LANCE LANE
HQ OPS Officer: DONG HWA PARK
Notification Date: 03/02/2010
Notification Time: 18:48 [ET]
Event Date: 03/02/2010
Event Time: 14:58 [CST]
Last Update Date: 03/02/2010
Notification Time: 18:48 [ET]
Event Date: 03/02/2010
Event Time: 14:58 [CST]
Last Update Date: 03/02/2010
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):
CHUCK CAIN (R4DO)
CHUCK CAIN (R4DO)
| 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 CAUSED BY LOW STEAM GENERATOR WATER LEVEL
"Wolf Creek experienced a reactor trip at 1458 CST. The first out annunciator was Steam Generator Level Lo Lo Reactor Trip.
"The trip was caused by the loss of the 'A' Main Feed Pump. The cause of the loss of the feed pump was due to the loss of 120 VAC non-safety instrument inverter PN09. PN09 supplies the Main Feed Pump Speed Control Circuitry.
"The loss of the PN09 also resulted in the loss of the ability to dump steam to the main condenser. Initial post trip decay heat was being removed with the Steam Generator Atmospheric Relief Valves and Auxiliary Feed Water. The Atmospheric Relief Valves cycled from approximately 1458 CST until approximately 1504 CST. Primary to Secondary leakage is less than 2.68 gallons per day.
"PN09 was re-energized at 1554 CST.
"All systems functioned as designed with the exception of the instrumentation powered by PN09. At the time of the trip the 'A' Emergency Diesel Generator and the 'A' Class IE Air Conditioning Unit were out of service for maintenance.
"The plant is being maintained at normal Mode 3 pressure and temperature."
The reactor trip was uncomplicated and decay heat is currently being removed by steam dumps to the main condensers.
The licensee has notified the NRC Senior Resident Inspector.
"Wolf Creek experienced a reactor trip at 1458 CST. The first out annunciator was Steam Generator Level Lo Lo Reactor Trip.
"The trip was caused by the loss of the 'A' Main Feed Pump. The cause of the loss of the feed pump was due to the loss of 120 VAC non-safety instrument inverter PN09. PN09 supplies the Main Feed Pump Speed Control Circuitry.
"The loss of the PN09 also resulted in the loss of the ability to dump steam to the main condenser. Initial post trip decay heat was being removed with the Steam Generator Atmospheric Relief Valves and Auxiliary Feed Water. The Atmospheric Relief Valves cycled from approximately 1458 CST until approximately 1504 CST. Primary to Secondary leakage is less than 2.68 gallons per day.
"PN09 was re-energized at 1554 CST.
"All systems functioned as designed with the exception of the instrumentation powered by PN09. At the time of the trip the 'A' Emergency Diesel Generator and the 'A' Class IE Air Conditioning Unit were out of service for maintenance.
"The plant is being maintained at normal Mode 3 pressure and temperature."
The reactor trip was uncomplicated and decay heat is currently being removed by steam dumps to the main condensers.
The licensee has notified the NRC Senior Resident Inspector.