Event Notification Report for March 24, 2001
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
03/23/2001 - 03/24/2001
Power Reactor
Event Number: 37862
Facility: COLUMBIA GENERATING STATION
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: C. PERINO
HQ OPS Officer: BOB STRANSKY
Region: 4 State: WA
Unit: [2] [] []
RX Type: [2] GE-5
NRC Notified By: C. PERINO
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/24/2001
Notification Time: 08:35 [ET]
Event Date: 03/24/2001
Event Time: 03:08 [PST]
Last Update Date: 03/24/2001
Notification Time: 08:35 [ET]
Event Date: 03/24/2001
Event Time: 03:08 [PST]
Last Update Date: 03/24/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
JOHN PELLET (R4)
JOHN PELLET (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | Y | 100 | Power Operation | 99 | Power Operation |
PLANT SHUTDOWN REQUIRED BY TECHNICAL SPECIFICATIONS
"This report is being made in accordance with 10 CFR 50.72(b)(2)(i).
"Columbia Generating Station started a plant shutdown 03:08. At 23:15 yesterday, a rise in drywell pressure was observed Coolant leakage was confirmed by a rising trend in drywell particulate monitors. Upon investigation, this leakage was confirmed to be due to cycling RCIC-V-63, the RCIC steam supply inboard isolation valve, during the performance of a Surveillance Procedure.
"At 23:30, a rise in drywell unidentified leakage rate was observed. At 01:56 an entry into Technical Specification 3.4.5.B for greater than a 2 gpm rise in a 24 hour period was made. At 03:08, an entry into Technical Specification 3.4.5 A was made for unidentified leakage greater than 5 gpm. At this time, the decision to begin a controlled reactor shutdown was made.
"RCIC-V-63 was backseated to terminate leakage at 03:55. At 03:58 it was determined the backseating successfully stopped the leakage. The downpower was terminated at 04:19. Indications of leakage are trending down. At 04:30, the leakage had decreased below the Technical Specification limits and reactor power was increased to 100%."
The NRC resident inspector has been informed of this event by the licensee.
"This report is being made in accordance with 10 CFR 50.72(b)(2)(i).
"Columbia Generating Station started a plant shutdown 03:08. At 23:15 yesterday, a rise in drywell pressure was observed Coolant leakage was confirmed by a rising trend in drywell particulate monitors. Upon investigation, this leakage was confirmed to be due to cycling RCIC-V-63, the RCIC steam supply inboard isolation valve, during the performance of a Surveillance Procedure.
"At 23:30, a rise in drywell unidentified leakage rate was observed. At 01:56 an entry into Technical Specification 3.4.5.B for greater than a 2 gpm rise in a 24 hour period was made. At 03:08, an entry into Technical Specification 3.4.5 A was made for unidentified leakage greater than 5 gpm. At this time, the decision to begin a controlled reactor shutdown was made.
"RCIC-V-63 was backseated to terminate leakage at 03:55. At 03:58 it was determined the backseating successfully stopped the leakage. The downpower was terminated at 04:19. Indications of leakage are trending down. At 04:30, the leakage had decreased below the Technical Specification limits and reactor power was increased to 100%."
The NRC resident inspector has been informed of this event by the licensee.
Power Reactor
Event Number: 37863
Facility: COMANCHE PEAK
Region: 4 State: TX
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: STEVEN SEWELL
HQ OPS Officer: BOB STRANSKY
Region: 4 State: TX
Unit: [1] [2] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: STEVEN SEWELL
HQ OPS Officer: BOB STRANSKY
Notification Date: 03/24/2001
Notification Time: 10:39 [ET]
Event Date: 03/24/2001
Event Time: 00:30 [CST]
Last Update Date: 03/24/2001
Notification Time: 10:39 [ET]
Event Date: 03/24/2001
Event Time: 00:30 [CST]
Last Update Date: 03/24/2001
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
Person (Organization):
JOHN PELLET (R4)
JOHN PELLET (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | Y | 88 | Power Operation | 88 | Power Operation |
| 2 | N | Y | 100 | Power Operation | 100 | Power Operation |
OFFSITE NOTIFICATION REGARDING DISCHARGE OF WATER OFFSITE
"On 3/24/01, at 0845, a determination was made that a report to the TNRCC [Texas Natural Resource Conservation Commission] is required for a discharge of water to Squaw Creek Reservoir. An estimated 1000 gallons of fire protection water was discharged due to a ruptured pipe in warehouse C. The water contains no hazardous constituents that have assigned reportable quantities; however, a report to the TNRCC will be made in accordance with water quality regulations."
The NRC resident inspector has been informed of this report by the licensee.
"On 3/24/01, at 0845, a determination was made that a report to the TNRCC [Texas Natural Resource Conservation Commission] is required for a discharge of water to Squaw Creek Reservoir. An estimated 1000 gallons of fire protection water was discharged due to a ruptured pipe in warehouse C. The water contains no hazardous constituents that have assigned reportable quantities; however, a report to the TNRCC will be made in accordance with water quality regulations."
The NRC resident inspector has been informed of this report by the licensee.
Power Reactor
Event Number: 37864
Facility: ARKANSAS NUCLEAR
Region: 4 State: AR
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: FRED VANBUSKIRK
HQ OPS Officer: FANGIE JONES
Region: 4 State: AR
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: FRED VANBUSKIRK
HQ OPS Officer: FANGIE JONES
Notification Date: 03/24/2001
Notification Time: 19:24 [ET]
Event Date: 03/24/2001
Event Time: 16:01 [CST]
Last Update Date: 03/24/2001
Notification Time: 19:24 [ET]
Event Date: 03/24/2001
Event Time: 16:01 [CST]
Last Update Date: 03/24/2001
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 PELLET (R4)
JOHN PELLET (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
PRESSURE BOUNDARY DEGRADATION FOUND DURING TESTING AFTER SHUTDOWN
"On March 18, 2001, following shutdown for a scheduled refueling outage, Arkansas Nuclear One (ANO) Unit I performed a routine visual inspection of the Reactor Vessel (RV) head area per NRC Generic Letter 88-05. This inspection revealed an increase in the amount of boric acid crystals in the area of one Control Rod Drive Mechanism (CRDM) nozzle over the amount that had been present during the previous inspection. Results of the visual inspection were inconclusive regarding the source of the leakage. No other CRDM nozzle showed similar boron buildup. Prior to unit shutdown, on March 16 the total unidentified Reactor Coolant System leak rate was determined to be 0.087 gpm, significantly less than the 1 gpm allowed by Technical Specifications. After the RV head was removed and placed on the storage stand, additional inspections were conducted. At approximately 1601 on March 24, 2001, eddy current and ultrasonic testing revealed an RCS pressure boundary leak. Using ultrasonic test (UT) methods one crack was identified in the wall of CRDM nozzle #56. The UT data indicates that the crack is on the downhill side of the nozzle. It extends approximately 0.8 inches below the weld and extends upward to approximately 1.0 inches above the weld. The depth of the crack is approximately 0.2 inches. This crack is consistent with industry experience associated with Primary Water Stress Corrosion Cracking (PWSCC) in similar components that have been evaluated as part of the NRC Genetic Letter 97-01 response. These cracks do not pose a significant safety risk during plant operation. ANO plans to perform repairs during the current outage. Since the crack resulted in leakage through the Reactor Coolant System pressure boundary, this condition is being reported as a non-emergency eight-hour report in accordance with 10 CFR 50.72(b)(3)(ii), degradation of a principal safety barrier."
The licensee notified the NRC Resident Inspector.
"On March 18, 2001, following shutdown for a scheduled refueling outage, Arkansas Nuclear One (ANO) Unit I performed a routine visual inspection of the Reactor Vessel (RV) head area per NRC Generic Letter 88-05. This inspection revealed an increase in the amount of boric acid crystals in the area of one Control Rod Drive Mechanism (CRDM) nozzle over the amount that had been present during the previous inspection. Results of the visual inspection were inconclusive regarding the source of the leakage. No other CRDM nozzle showed similar boron buildup. Prior to unit shutdown, on March 16 the total unidentified Reactor Coolant System leak rate was determined to be 0.087 gpm, significantly less than the 1 gpm allowed by Technical Specifications. After the RV head was removed and placed on the storage stand, additional inspections were conducted. At approximately 1601 on March 24, 2001, eddy current and ultrasonic testing revealed an RCS pressure boundary leak. Using ultrasonic test (UT) methods one crack was identified in the wall of CRDM nozzle #56. The UT data indicates that the crack is on the downhill side of the nozzle. It extends approximately 0.8 inches below the weld and extends upward to approximately 1.0 inches above the weld. The depth of the crack is approximately 0.2 inches. This crack is consistent with industry experience associated with Primary Water Stress Corrosion Cracking (PWSCC) in similar components that have been evaluated as part of the NRC Genetic Letter 97-01 response. These cracks do not pose a significant safety risk during plant operation. ANO plans to perform repairs during the current outage. Since the crack resulted in leakage through the Reactor Coolant System pressure boundary, this condition is being reported as a non-emergency eight-hour report in accordance with 10 CFR 50.72(b)(3)(ii), degradation of a principal safety barrier."
The licensee notified the NRC Resident Inspector.