05000366/LER-2007-004
| Docket Number(S) | |
| Event date: | 03-09-2007 |
|---|---|
| Report date: | 05-03-2007 |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) |
| 3662007004R00 - NRC Website | |
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor Energy Industry Identification System codes appear in the text as (EIIS Code XX).
DESCRIPTION OF EVENT
On March 9, 2007, at 1230 EST, Unit 2 was in the Cold Shutdown mode. At that time, a leak was identified in a one-inch socket weld elbow, located below the condensing chamber on the "D" Main Steam Line (EIIS Code SB). The TS definition of pressure boundary leakage is "leakage through a non-isolable fault in the reactor coolant system." Due to its location, the leak met this definition.
Based upon inspection of the weld and adjacent area, it was determined that the leak existed when the Unit was in mode 1. The TS allows no pressure boundary leakage in mode 1. A section of piping was removed and sent to an independent laboratory for analysis. Based on this analysis, the piping was replaced using high cycle fatigue resistant welds (2 to 1 welds).
CAUSE OF EVENT
The cause of the leak is failure of the socket weld, which was caused by high cycle fatigue.
REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT
This report is required, per 10 CFR 50.73(a)(2)(i)(B), because a condition existed which was prohibited by the plant's TS. The Unit 2 TS allows no pressure boundary leakage in mode 1. The long-term increasing trend in drywell floor drain sump in-leakage, the discovery of a leak at a one- inch socket weld elbow located below the condensing chamber on the "D" Main Steam Line, and indication of spray impingement on adjacent components, led to the determination that a leak in the pressure boundary had existed for longer than allowed by the TS. Therefore, the plant was in a condition prohibited by the Unit 2 TS.
The reactor coolant system (RCS) includes systems and components that contain or transport the coolant to or from the reactor core. The pressure-retaining components of the RCS and the portions of connecting systems out to and including the isolation valves define the reactor coolant pressure boundary. During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration. Limits on RCS operational leakage are required to ensure appropriate action is taken before the integrity of the reactor coolant pressure boundary is compromised. The TS delineate the limits on the specific types of leakage.
The unidentified leakage flow limit allows time for corrective action to be taken before the reactor coolant pressure boundary can be compromised significantly. The five gallons per minute (gpm) limit is a small fraction of the calculated flow from a critical crack in the primary system piping. A critical crack is one large enough to propagate rapidly, ultimately leading to failure of the affected component. Crack behavior from experimental programs shows that leakage rates of hundreds of gallons per minute will precede crack instability (see Unit 2 Final Safety Analysis Report, section 5.2.7.5, "Nuclear System Leakage Detection and Leakage Rate Limits," and Unit 2 TS Bases B 3.4.4, "RCS Operational Leakage").
In this event, unidentified leakage into the drywell had increased linearly from approximately 0.02 gpm at the beginning of March 2006 to approximately 0.13 gpm at the beginning of the refueling outage in February 2007. During outage activities, a leak was identified and investigated.
This leak was determined to meet the TS definition of pressure boundary leakage.
At the time the unit was shut down, the unidentified leakage rate was less than three percent of the TS limit of five gpm. The size of the crack was much smaller than the "critical crack" (on which the TS limit is based) as evidenced by the low leakage rate. Therefore, at the time it was discovered and corrective action taken, the crack was not unstable and would not have resulted in catastrophic failure of the line. However, a worst-case instantaneous and complete severing of the one-inch line, due to the presence of a crack, would not result in a significant loss of reactor coolant or present any challenge to core cooling.
A rupture of this one-inch steam line does not result in a significant decrease in water inventory within the vessel. In addition, even if the inventory loss were completely water, the break would still be bounded by both the Loss of Coolant Accident analysis and the Feedwater Line break analysis. This proposed leak is less than 10 percent of the rated capacity of the High Pressure Coolant Injection (EIIS Code BJ) system, which is sized to provide adequate coolant make-up for pipe breaks up to four inches, and approximates the rated capacity of the Reactor Core Isolation Cooling (EIIS Code BN) system. It should be noted that the calculation assumed only liquid flows out of the resulting opening; in reality, a combination of liquid and vapor would flow from the break area. The actual, two-phase flow rate would be lower than that resulting from liquid only.
Consequently, either system would have been capable of indefinitely maintaining normal reactor water level. Additionally, a leak of several hundred gpm would have adequately been accommodated by the feedwater system (EIIS Code SJ), which has a flow rate capacity margin at rated conditions of at least 10 percent (over 2000 gpm). Therefore, any one of three diverse and independent high pressure injection systems could have provided sufficient make-up flow to maintain water level well above the top of the active fuel.
Based upon the preceding analysis, it is concluded that this event had no adverse impact on nuclear safety. This analysis is applicable to all operating conditions under which the crack might have propagated to line failure.
CORRECTIVE ACTIONS
The failed weld and adjacent welds were replaced with high cycle fatigue resistant welds.
Inspection of the three other similar lines was performed. No additional leaks were identified.
ADDITIONAL INFORMATION
Other Systems Affected: No systems were affected by this event other than those which have already been discussed in this report.
Failed Components Information:
Master Parts List Number: 2B21-Socket Weld�EIIS System Code: SB Manufacturer: N/A Reportable to EPIX: No Model Number: N/A Root Cause Code: B Type: N/A EIIS Component Code: N/A Manufacturer Code: N/A Commitment Information: This report does not create any permanent licensing commitments.
Previous Similar Events: No License Event Reports have been reported in the past two years in which a failure of the reactor pressure boundary has occurred.