ENS 45107
ENS Event | |
|---|---|
18:30 May 31, 2009 | |
| Title | Degraded Condition Due to Discovery of Pressure Boundary Leakage |
| Event Description | {{#Wiki_filter: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). }}[[Event description::Description::{{#Regex_clear: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). }}| ]] |
| Where | |
|---|---|
Browns Ferry  Alabama (NRC Region 2) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded | |
| Time - Person (Reporting Time:+-0.85 h-0.0354 days <br />-0.00506 weeks <br />-0.00116 months <br />) | |
| Opened: | William Baker 17:39 May 31, 2009 |
| NRC Officer: | Howie Crouch |
| Last Updated: | Jun 3, 2009 |
| 45107 - NRC Website
Loading map... {"minzoom":false,"maxzoom":false,"mappingservice":"leaflet","width":"350px","height":"250px","centre":false,"title":"","label":"","icon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"zoom":6,"defzoom":14,"layers":["OpenStreetMap"],"image layers":[],"overlays":[],"resizable":false,"fullscreen":false,"scrollwheelzoom":true,"cluster":false,"clustermaxzoom":20,"clusterzoomonclick":true,"clustermaxradius":80,"clusterspiderfy":true,"geojson":"","clicktarget":"","imageLayers":[],"locations":[{"text":"\u003Cb\u003E\u003Cdiv class=\"mw-parser-output\"\u003E\u003Cp\u003E\u003Ca class=\"mw-selflink selflink\"\u003EENS 45107\u003C/a\u003E - \u003Ca href=\"/Browns_Ferry\" title=\"Browns Ferry\"\u003EBrowns Ferry\u003C/a\u003E\n\u003C/p\u003E\u003C/div\u003E\u003C/b\u003E\u003Cdiv class=\"mw-parser-output\"\u003E\u003Cp\u003EDegraded Condition Due to Discovery of Pressure Boundary Leakage\n\u003C/p\u003E\u003C/div\u003E","title":"ENS 45107 - Browns Ferry\n","link":"","lat":34.70418333333333,"lon":-87.11844722222222,"icon":"/w/images/c/ce/Tennessee_Valley_Authority_icon.png"}],"imageoverlays":null} | |
Unit 2 | |
|---|---|
| Reactor critical | Not Critical |
| Scram | No |
| Before | Cold Shutdown (0 %) |
| After | Cold Shutdown (0 %) |
WEEKMONTHYEARENS 564112023-03-15T03:57:00015 March 2023 03:57:00
[Table view]10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Reactor Coolant System (RCS) Boundary Degraded Condition ENS 563712023-02-18T10:39:00018 February 2023 10:39:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Degraded Condition ENS 562572022-12-03T16:00:0003 December 2022 16:00:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Degraded Condition Discovered on Shutdown Cooling Test Line ENS 557062022-01-16T05:20:00016 January 2022 05:20:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Degraded Condition Discovered on Shutdown Cooling Test Line ENS 539592019-03-26T15:30:00026 March 2019 15:30:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Report of Degraded Condition Due to Leaking Valve ENS 451072009-05-31T18:30:00031 May 2009 18:30:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Degraded Condition Due to Discovery of Pressure Boundary Leakage ENS 446802008-11-23T18:00:00023 November 2008 18:00:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Pressure Boundary Leak ENS 436592007-09-22T17:45:00022 September 2007 17:45:00 10 CFR 50.72(b)(3)(ii)(A), Seriously Degraded Pressure Boundary Leakage Discovered During Drywell Inspection 2023-03-15T03:57:00 | |