Non-destructive and Failure Evaluation of Tubing from a Retired Steam Generator (NUREG/CR-6924)
On this page:
Download complete document
Manuscript Completed: August 2006
Date Published: March 2007
D.S. Kupperman, J.Y. Park, S. Majumdar, S. Bakhtiari
K. Kasza, W.J. Shack
Argonne National Laboratory
9700 South Cass Avenue
Argonne, IL 60439
E. Reichelt, NRC Project Manager
Division of Fuel, Engineering and Radiological Research
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
NRC Job Code Y6588
Stress corrosion cracks (SCC) in tubes removed from a nuclear reactor retired steam generator (SG) have been examined. Eddy current (EC) nondestructive evaluation (NDE), pressure testing, and metallurgical sectioning or fractography were carried out on six tube support plate SCCs and four tube sheet SCCs. The deposits on the top of the tube sheet in the vicinity of the cracking were analyzed. Failure pressures and leak rates for the tube support plate SCCs were evaluated using finite element analysis and were compared to experimental results. Crack profiles used in the analysis were estimated using eddy current examination and fractography results.
The data obtained from the pressure tests were compared with the industry probability of leakage database and correlation. The results were generally consistent with the probability of leakage values presented by industry, i.e., tubes that would not be expected to leak based on measured eddy current voltages and the industry correlation did not leak.
The best predictions of the SCC profile and crack depth were obtained from the use of a multiparameter algorithm applied to rotating pancake coil data. Very little correlation between +Point amplitude and maximum depth of SCC was found for the tubes that were examined.
Page Last Reviewed/Updated Thursday, March 25, 2021