05000366/LER-2007-004, Regarding Leak in Reactor Pressure Boundary Due to Failure of a Socket Weld
| ML071230404 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/03/2007 |
| From: | Madison D Southern Nuclear Operating Co |
| To: | Document Control Desk, Plant Licensing Branch III-2 |
| References | |
| NL-07-0956 LER 07-004-00 | |
| Download: ML071230404 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
| 3662007004R00 - NRC Website | |
text
D. L Madiron (hair)
Sodhrrn birr V i President - Hatch Oprrtinfl I=-
Plant W i n I. Hatch 11028 H a h Parkway, North Baxley. Georgia 3151 3 Tel912.537.5859 Fax 91 2.366.2077 May 3,2007 Docket No.:
50-366 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 Licensee Event Report 2-2007-004 Leak in Reactor Pressure Boundary Due to Failure of a Socket Weld Ladies and Gentlemen:
In accordance with the requirements of 10 CFR 50.73(a)(2)(i), Southern Nuclear Operating Company (SNC) is submitting the enclosed Licensee Event Report concerning a reactor pressure boundary leak that resulted from a failed socket weld.
This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, Vice President - Hatch Edwin I. Hatch Nuclear Plant 1 1028 Hatch Parkway North Baxley, GA 31513 Enclosure: LER 2-2007-004 cc:
Southern Nuclear Operating Commy Mr. J. T. Gasser, Executive Vice President Mr. D. H. Jones, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear ~ e d a t o ~
Commission Dr. W. D. Travers, Regional Administrator Mr. R. E. Martin, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch
LICENSEE EVENT REPORT (LER)
(See reverse for required number of 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. The Technical Specification (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.
The cause of the leak is failure of the socket weld due to high cycle fatigue.
Corrective actions for this event included replacing the failed weld and adjacent welds with high cycle fatigue resistant welds and inspection of similar piping to confirm no other leaks.
NRC FORM 366 (52004)
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.
(If more space is required, use additional copies of NRC Form 366.4)
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: 2B2 1 -Socket Weld EIIS System Code: SB Manufacturer: NIA Reportable to EPIX: No Model Number: NIA Root Cause Code: B Type: NIA EIIS Component Code: NIA Manufacturer Code: NIA 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.
tC FORM 36644 (1-2001)