Event Notification Report for October 07, 2002
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
10/06/2002 - 10/07/2002
Power Reactor
Event Number: 39253
Facility: COMANCHE PEAK
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MIKE SMITH
HQ OPS Officer: STEVE SANDIN
Region: 4 State: TX
Unit: [1] [] []
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: MIKE SMITH
HQ OPS Officer: STEVE SANDIN
Notification Date: 10/07/2002
Notification Time: 12:14 [ET]
Event Date: 10/07/2002
Event Time: 09:35 [CDT]
Last Update Date: 10/07/2002
Notification Time: 12:14 [ET]
Event Date: 10/07/2002
Event Time: 09:35 [CDT]
Last Update Date: 10/07/2002
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
WILLIAM JONES (R4)
WILLIAM JONES (R4)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
AUTOSTART OF EDG FOLLOWING TRANSFER OF "1E" SWITCHGEAR TO ALTERNATE POWER SUPPLY
"During Design Modification testing in the switchyard, an unplanned contact picked up which de-energized the train 'C' electrical switchgear and the '1E' switchgear to transfer to the Alternate Power (-2 bkr). Core Alterations were in progress and were suspended."
Train "C" consists of four (4) electrical buses; 1A1, 1A2, 1A3 and 1A4. The -2 bkr transferred power from the 1EA1 to the 1EA2 bus which is energized from the XST1 transformer. During the slow transfer, the operating Spent Fuel Pool (SFP) pump and RHR "B" train were de-energized. There was no observed increase in SFP or appreciable increase in RCS temperatures. RHR cooling was re-established in approximately eight (8) minutes. Emergency Diesel Generator "1-02" autostarted on low voltage but did not supply power to the affected 1E buses. The cause of the EDG autostart is under investigation. Core Alterations are suspended pending completion of the investigation.
The licensee informed the NRC resident inspector.
"During Design Modification testing in the switchyard, an unplanned contact picked up which de-energized the train 'C' electrical switchgear and the '1E' switchgear to transfer to the Alternate Power (-2 bkr). Core Alterations were in progress and were suspended."
Train "C" consists of four (4) electrical buses; 1A1, 1A2, 1A3 and 1A4. The -2 bkr transferred power from the 1EA1 to the 1EA2 bus which is energized from the XST1 transformer. During the slow transfer, the operating Spent Fuel Pool (SFP) pump and RHR "B" train were de-energized. There was no observed increase in SFP or appreciable increase in RCS temperatures. RHR cooling was re-established in approximately eight (8) minutes. Emergency Diesel Generator "1-02" autostarted on low voltage but did not supply power to the affected 1E buses. The cause of the EDG autostart is under investigation. Core Alterations are suspended pending completion of the investigation.
The licensee informed the NRC resident inspector.
Power Reactor
Event Number: 39254
Facility: ARKANSAS NUCLEAR
Region: 4 State: AR
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: HOGUE
HQ OPS Officer: CHAUNCEY GOULD
Region: 4 State: AR
Unit: [1] [] []
RX Type: [1] B&W-L-LP,[2] CE
NRC Notified By: HOGUE
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 10/07/2002
Notification Time: 12:38 [ET]
Event Date: 10/07/2002
Event Time: 05:07 [CDT]
Last Update Date: 10/07/2002
Notification Time: 12:38 [ET]
Event Date: 10/07/2002
Event Time: 05:07 [CDT]
Last Update Date: 10/07/2002
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):
WILLIAM JONES (R4)
CHRIS GRIMES (EO)
WILLIAM JONES (R4)
CHRIS GRIMES (EO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
BORIC ACID CRYSTALS DETECTED IN THE AREA OF THE CONTROL ROD DRIVE MECHANISM NOZZLE
On October 7, 2002, following shutdown for a scheduled refueling outage, Arkansas Nuclear One (ANO) Unit 1 personnel performed a routine visual inspection of the Reactor Vessel (RV) head area per NRC Generic Letter 88-05. This inspection revealed a small nodule of boric acid crystals in the area of Control Rod Drive Mechanism (CRDM) Nozzle #56. The boric acid residue was detected on the down hill side of the nozzle (extending 180 degrees In the nozzle annulus area), with a small nodule of boric acid at the most downhill point. Prior to unit shutdown, on October 4. the total unidentified Reactor Coolant System (RCS) leak rate was determined to be significantly less than the 1 gpm allowed by Technical Specifications. This indication is consistent with industry experience associated with Primary Water Stress Corrosion Cracking (PWSCC) in similar components that has been evaluated as part of the NRC Generic 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 visual inspection indicates leakage through the Reactor Coolant System pressure boundary, this condition is being reported as a non-emergency eight-hour report in accordance with 10CFR50.72(b)(3)(II), degradation of a principal safety barrier. Reactor vessel head inspection is continuing. Any additional findings will be reported as an update to this notification.
The NRC Resident Inspector was notified.
On October 7, 2002, following shutdown for a scheduled refueling outage, Arkansas Nuclear One (ANO) Unit 1 personnel performed a routine visual inspection of the Reactor Vessel (RV) head area per NRC Generic Letter 88-05. This inspection revealed a small nodule of boric acid crystals in the area of Control Rod Drive Mechanism (CRDM) Nozzle #56. The boric acid residue was detected on the down hill side of the nozzle (extending 180 degrees In the nozzle annulus area), with a small nodule of boric acid at the most downhill point. Prior to unit shutdown, on October 4. the total unidentified Reactor Coolant System (RCS) leak rate was determined to be significantly less than the 1 gpm allowed by Technical Specifications. This indication is consistent with industry experience associated with Primary Water Stress Corrosion Cracking (PWSCC) in similar components that has been evaluated as part of the NRC Generic 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 visual inspection indicates leakage through the Reactor Coolant System pressure boundary, this condition is being reported as a non-emergency eight-hour report in accordance with 10CFR50.72(b)(3)(II), degradation of a principal safety barrier. Reactor vessel head inspection is continuing. Any additional findings will be reported as an update to this notification.
The NRC Resident Inspector was notified.
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
!!!!! THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor
Event Number: 39255
Facility: VERMONT YANKEE
Region: 1 State: VT
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: BROOKS
HQ OPS Officer: CHAUNCEY GOULD
Region: 1 State: VT
Unit: [1] [] []
RX Type: [1] GE-4
NRC Notified By: BROOKS
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 10/07/2002
Notification Time: 19:02 [ET]
Event Date: 10/07/2002
Event Time: 17:15 [EDT]
Last Update Date: 12/04/2002
Notification Time: 19:02 [ET]
Event Date: 10/07/2002
Event Time: 17:15 [EDT]
Last Update Date: 12/04/2002
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 (R1)
JOHN WHITE (R1)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 1 | N | N | 0 | Refueling | 0 | Refueling |
PLANT FAILED THEIR INTEGRATED LEAKRATE TESTING
During the normal outage integrated leakrate testing, it was determined the total primary containment leakage rate is in excess of allowable leakage rate. The calculated leakage rate is 216.55 scfh while the allowable leakage rate is 192.126 scfh. Repair work on the valves has been initiated.
The NRC Resident Inspector will be notified.
* * * RETRACTION FROM DENNIS MAY TO HOWIE CROUCH ON 12/4/02 AT 1430 EST * * *
"BASIS FOR RETRACTION:
"An NRC 8-Hour Non-Emergency Event Notification (#39255) was conservatively made following MSIV [Main Steam Isolation Valve] leak rate testing based upon exceeding the combined local leak rate test acceptance criterion based on a calculated maximum pathway leak rate pending further review by the Leak Rate Testing (Appendix "J") Coordinator.
"NUREG 1022, Rev. 2 provides guidance for the evaluation of containment leak rate test data under the 50.72(b)(3)(ii)(A) reporting criteria that specifies the use of the total containment as-found minimum pathway leak rate.
"The Technical Specification limit for the combined local leak rate test (Type B and C test) acceptance criterion is less than or equal to 0.6 La (192.126 scfh). The combined local leak rate test, calculated on a minimum pathway basis was 112.807 scfh. This value is less than the Technical Specification limit and therefore is not reportable in accordance with 10CFR50.72(b)(3)(ii)(A) or 50.73(a)(2)(ii).
Therefore, ENS Event Number 39255, made on 10/07/02, is being retracted."
Headquarters Operations Officer notified R1DO (Michele Evans). The licensee will notify the NRC Resident.
During the normal outage integrated leakrate testing, it was determined the total primary containment leakage rate is in excess of allowable leakage rate. The calculated leakage rate is 216.55 scfh while the allowable leakage rate is 192.126 scfh. Repair work on the valves has been initiated.
The NRC Resident Inspector will be notified.
* * * RETRACTION FROM DENNIS MAY TO HOWIE CROUCH ON 12/4/02 AT 1430 EST * * *
"BASIS FOR RETRACTION:
"An NRC 8-Hour Non-Emergency Event Notification (#39255) was conservatively made following MSIV [Main Steam Isolation Valve] leak rate testing based upon exceeding the combined local leak rate test acceptance criterion based on a calculated maximum pathway leak rate pending further review by the Leak Rate Testing (Appendix "J") Coordinator.
"NUREG 1022, Rev. 2 provides guidance for the evaluation of containment leak rate test data under the 50.72(b)(3)(ii)(A) reporting criteria that specifies the use of the total containment as-found minimum pathway leak rate.
"The Technical Specification limit for the combined local leak rate test (Type B and C test) acceptance criterion is less than or equal to 0.6 La (192.126 scfh). The combined local leak rate test, calculated on a minimum pathway basis was 112.807 scfh. This value is less than the Technical Specification limit and therefore is not reportable in accordance with 10CFR50.72(b)(3)(ii)(A) or 50.73(a)(2)(ii).
Therefore, ENS Event Number 39255, made on 10/07/02, is being retracted."
Headquarters Operations Officer notified R1DO (Michele Evans). The licensee will notify the NRC Resident.