Event Notification Report for May 31, 2009
U.S. Nuclear Regulatory Commission
Operations Center
EVENT REPORTS FOR
05/30/2009 - 05/31/2009
EVENT NUMBERS
45107
Power Reactor
Event Number: 45107
Facility: BROWNS FERRY
Region: 2 State: AL
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: WILLIAM BAKER
HQ OPS Officer: HOWIE CROUCH
Region: 2 State: AL
Unit: [] [2] []
RX Type: [1] GE-4,[2] GE-4,[3] GE-4
NRC Notified By: WILLIAM BAKER
HQ OPS Officer: HOWIE CROUCH
Notification Date: 05/31/2009
Notification Time: 17:39 [ET]
Event Date: 05/31/2009
Event Time: 13:30 [CDT]
Last Update Date: 06/02/2009
Notification Time: 17:39 [ET]
Event Date: 05/31/2009
Event Time: 13:30 [CDT]
Last Update Date: 06/02/2009
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):
GEORGE HOPPER (R2DO)
GEORGE HOPPER (R2DO)
| Unit | SCRAM Code | RX Crit | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode |
|---|---|---|---|---|---|---|
| 2 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown |
DEGRADED CONDITION DUE TO DISCOVERY OF PRESSURE BOUNDARY LEAKAGE
"During the performance of 2-SI-3.3.1.A ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping (ASME Section III, Class 1 and 2), a pressure boundary leak was identified on the RHR Shutdown Cooling Root Valve (2-SHV-074-0049) at a weld between the valve body and a 3/4 inch schedule 160 pipe nipple. The pipe nipple is a 6 inch long capped valve leak off line. This valve and line are ASME Code Class 1 equivalent components located between the A recirculation piping and the inboard Shutdown Cooling isolation valve (2-FCV-074-0048). This condition caused entry into Technical Requirements Manual (TRM) 3.4.3 - Structural Integrity - Condition A - in which the applicability is at all times and the required action is to immediately restore the structural integrity of the affected component to within its limit or maintain the reactor in MODE 4 or 5 or the reactor coolant system less than 50?F above the minimum temperature required by NDT considerations, until each indication of a defect has been investigated and evaluated. The plant is currently in MODE 4 with Shutdown Cooling in service.
"This event is reportable within 8 hours under 10CFR50.72 (b)(3)(ii)(A), 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' This event is also a reportable within 60 days under 10CFR50.73(a)(2)(ii)(A), 'Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The NRC Resident Inspector has been notified. This condition has been documented in the BFN [Browns Ferry Nuclear] Corrective Action program as PER# 172551.
"Unit 1 and 3 remain at 100% power and are not affected by this event."
* * * UPDATE FROM W. BAKER TO P. SNYDER AT 1839 ON 6/2/09 * * *
"Additional review of available data and inspection results revealed that the original identification of a 6 inch long ¥ inch diameter schedule 160 pipe nipple was part of the original design of the 2-SHV-074-0049. This connection to the valve body was subsequently modified by a design change documented in DCN S17640C and installed in 1992 by Work Order 92-49415-00 as a 2.55 inch long plug with a 0.800 inch tapered end, fabricated from bar stock to fit the existing valve socket weld cavity. All other details described in the original (REV 0) event notification remain as stated. The BFN Corrective Action PER# 172551 will be amended to include this additional information.
"The NRC [Resident Inspector] has been notified of this event notification revision."
Notified R2DO (Lesser).
"During the performance of 2-SI-3.3.1.A ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping (ASME Section III, Class 1 and 2), a pressure boundary leak was identified on the RHR Shutdown Cooling Root Valve (2-SHV-074-0049) at a weld between the valve body and a 3/4 inch schedule 160 pipe nipple. The pipe nipple is a 6 inch long capped valve leak off line. This valve and line are ASME Code Class 1 equivalent components located between the A recirculation piping and the inboard Shutdown Cooling isolation valve (2-FCV-074-0048). This condition caused entry into Technical Requirements Manual (TRM) 3.4.3 - Structural Integrity - Condition A - in which the applicability is at all times and the required action is to immediately restore the structural integrity of the affected component to within its limit or maintain the reactor in MODE 4 or 5 or the reactor coolant system less than 50?F above the minimum temperature required by NDT considerations, until each indication of a defect has been investigated and evaluated. The plant is currently in MODE 4 with Shutdown Cooling in service.
"This event is reportable within 8 hours under 10CFR50.72 (b)(3)(ii)(A), 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' This event is also a reportable within 60 days under 10CFR50.73(a)(2)(ii)(A), 'Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.' The NRC Resident Inspector has been notified. This condition has been documented in the BFN [Browns Ferry Nuclear] Corrective Action program as PER# 172551.
"Unit 1 and 3 remain at 100% power and are not affected by this event."
* * * UPDATE FROM W. BAKER TO P. SNYDER AT 1839 ON 6/2/09 * * *
"Additional review of available data and inspection results revealed that the original identification of a 6 inch long ¥ inch diameter schedule 160 pipe nipple was part of the original design of the 2-SHV-074-0049. This connection to the valve body was subsequently modified by a design change documented in DCN S17640C and installed in 1992 by Work Order 92-49415-00 as a 2.55 inch long plug with a 0.800 inch tapered end, fabricated from bar stock to fit the existing valve socket weld cavity. All other details described in the original (REV 0) event notification remain as stated. The BFN Corrective Action PER# 172551 will be amended to include this additional information.
"The NRC [Resident Inspector] has been notified of this event notification revision."
Notified R2DO (Lesser).