05000425/LER-2005-003
| Docket Number(S) | |
| Event date: | 12-10-2005 |
|---|---|
| Report date: | 3-0-5000 |
| 4252005003R00 - NRC Website | |
A) REQUIREMENT FOR REPORT This report is required per 10 CFR 50.73 (a)(2)(i)(A), because a unit shutdown to Mode 5 (Cold Shutdown) was completed on December 10, 2005, to comply with Technical Specification 3.4.13.a., following the discovery of Reactor Coolant System (RCS) pressure boundary leakage.
B) UNIT STATUS AT TIME OF EVENT Prior to the start of this event, Unit 2 was in Mode 1 (Power Operations) at 100% rated thermal power.
Other than that described herein, there was no inoperable equipment that contributed to the occurrence of this event.
C) DESCRIPTION OF EVENT On December 8, 2005, at 1450 EST, control room operators entered abnormal operating procedure 18004-C, "RCS Leakage," after containment radiation monitors indicated an increase in radioactivity in the containment building atmosphere. An RCS leakage calculation determined unidentified leakage to be 0.254 gpm. An inspection team entered the containment building and found no leakage outside the bioshield wall. A robotic camera was sent inside the bioshield wall and on December 9, 2005, at 0412 EST, the camera observed water running down the inside of the bioshield wall. Unit 2 was placed in Mode 3 (Hot Standby) at 1313 EST on December 9, 2005. At 1544 EST, an inspection team found leakage at an ASME Class 1, 3/4" socket welded piping connection inside the bioshield wall, above the RCS Loop 2 cold leg. Because the connection was non-isolable, this constituted pressure boundary leakage that required Unit 2 to be placed in Mode 5 (Cold Shutdown). This was accomplished on December 10, 2005, at 1019 EST. The NRC Operations Center was notified of this event on December 9, 2005, at 1811 EST.
The leak occurred in a 3/4" socket weld between a 0.250" flow restrictor and a 3/4" ANSI 6000# half coupling. The 3/4" half-coupling is drilled for a 0.375" flow opening. This 3/4" piping is part of a bypass line used to equalize pressure on either side of the 12" Residual Heat Removal (RHR) Loop Suction Valve, 21-1V-8701B. The weld defect was linear, approximately 1/4" to V2" long and about 1/8" from the toe of the weld nearest the half-coupling.
D) CAUSE OF EVENT The probable cause of the defect is a lack of fusion, a weld fabrication flaw that occurred when this weld was installed in October 2002 as part of a design change. The photographs of the flaw identified it to be located at an apparent start/stop point of a small bore 3/4" socket weld. The start/stop point in a weld is a common location for lack of fusion type defects. Socket welds are inspected by surface examination methods. Therefore, when the weld was originally inspected there were no flaws open to the surface. However, a lack of fusion type flaw at the weld root may have existed since installation, resulting in a shortened effective weld throat. Such a root anomaly could have propagated to the surface as the result of normal operating conditions that would not have affected a sound weld.
On December 16, 2005, an interview was conducted with the welder who affected the removal and replacement of the subject weld. During the removal of the defective weld by grinding, the welder described what he characterized as a lack of fusion at the weld root and observed little or no weld metal penetration at the root of the weld.
E) ANALYSIS OF EVENT Operators properly responded to determine the source of the leakage by initiating a plant shutdown and then initiating a unit cooldown to cold shutdown when it was determined to involve the RCS pressure boundary. During this event, the maximum total unidentified leakage was calculated to be 0.477 gpm.
Had a complete rupture of the 3/4" connection occurred, coolant would have been lost through the 3/8" opening of the half-coupling. However, the Chemical Volume and Control System (CVCS) is fully capable of making-up this quantity of coolant loss, plus any additional leakage through the 2HV-8701B valve as allowed by Technical Specifications, and the RCS would have retained its full volume and pressure. Based on these considerations, there was no adverse effect on plant safety or on the health and safety of the public as a result of this event.
This event does not represent a safety system functional failure.
F) CORRECTIVE ACTIONS 1)The welded connection was repaired, and the unit was re-started and returned to power operations.
2) Similar welded connections made as part of the design change that was implemented on Unit 2 in 2002 were inspected and found to be acceptable prior to the unit restart.
3) A similar design change was implemented on Unit 1. The non-isolable socket welded connections associated with this design change in Unit 1 will be inspected no later than the end of the next refueling outage, which is planned for Fall 2006.
G) ADDITIONAL INFORMATION 1) Failed Components:
None 2) Previous Similar Events:
There have been no previous similar events in the last three years.
3) Energy Industry Identification System Codes:
Reactor Coolant System — AB Chemical Volume and Control System — CB Plant Effluent Radiation Monitoring System — IL Residual Heat Removal System — BP