| ||Format||Pages||Price|| |
|PDF (576K)||30||$25||  ADD TO CART|
|Complete Source PDF (20M)||787||$195||  ADD TO CART|
The in-reactor oxidation of zirconium alloys, the structural material in almost all water-cooled and water-moderated reactors, can be life-limiting. In September 1989 the International Atomic Energy Agency (IAEA), in cooperation with the U.S. Department of Energy (DOE), arranged a workshop on zirconium-alloy corrosion. This paper reviews the primary findings of the workshop, considers some previously reported work to provide perspective, and summarizes the state of the technology. During the 1980s, significant progress was made towards correlating microstructural features and irradiation-induced changes with corrosion resistance. The development and use of sophisticated microcharacterization tools contributed greatly to these advancements. Also, significant progress toward understanding chemistry effects has been made. During the 1990s, the actual mechanism of corrosion should become well enough understood to support optimization of both alloy composition and microstructure. This will provide utilities with the ability to minimize fuel-cycle costs through burnup optimization.
zirconium, Zircaloy, corrosion
Manager, Cladding Development, Westinghouse Bettis Atomic Power Laboratory, West Mifflin, PA
Program Manager, Extended Burnup, Washington, DC,