Pb-assisted stress-corrosion cracking (PbSCC) of Alloy 600 steam-generator tubing in high-temperature-water service and laboratory tests were studied by analytical transmission electron microscopy of cross-sectioned samples. Examinations of pulled tubes from many pressurized water reactors revealed lead in cracks from 11 of 17 samples. Comparisons of the degraded intergranular structures with ones produced in simple laboratory tests with PbO in near-neutral AVT water showed that the PbSCC characteristics in service tubing could be reproduced without complex chemistries and heat-flow conditions that can occur during plant operation. Observations of intergranular and transgranular cracks promoted by Pb in the test samples also provided new insights into the mechanisms of PbSCC in mill-annealed and thermally treated Alloy 600.
Revised: May 19, 2011 |
Published: August 15, 2005
Citation
Thomas L., and S.M. Bruemmer. 2005.Observations and insights into Pb-assisted stress corrosion cracking of alloy 600 steam generator tubes. In 12th International Conference on Environmental Degradation of Materials in Nuclear Systems-Water Reactors, edited by TR Allen, PJ King and L Nelson, 1143-1154. Warrendale, Pennsylvania:TMS.PNNL-SA-45801.