Article ID | Journal | Published Year | Pages | File Type |
---|---|---|---|---|
1567266 | Journal of Nuclear Materials | 2010 | 6 Pages |
Abstract
The search for a new cladding material is part of the research studies carried out at CEA to develop a sodium-cooled fast reactor meeting the expectations of the Generation IV International Forum. In this study, the tensile properties of a ferritic oxide dispersion strengthened steel produced by hot extrusion at CEA have been evaluated. They prove the studied alloy to be as resistant as and more ductile than the other nano-reinforced alloys of literature. The effects of the strain rate and temperature on the total plastic strain of the material remind of diffusion phenomena. Intergranular damage and intergranular decohesion are clearly highlighted.
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Authors
A. Steckmeyer, M. Praud, B. Fournier, J. Malaplate, J. Garnier, J.L. Béchade, I. Tournié, A. Tancray, A. Bougault, P. Bonnaillie,