Article ID | Journal | Published Year | Pages | File Type |
---|---|---|---|---|
7963204 | Journal of Nuclear Materials | 2018 | 9 Pages |
Abstract
Understanding micromechanics of plastic deformation of existing materials is essential for improving their properties further and/or developing advanced materials for much more severe load bearing applications. The objective of the present work was to understand micromechanics of plastic deformation of Zircaloy-4, a zirconium-based alloy used as fuel cladding and channel (in BWRs) material in nuclear reactors. The Zircaloy-4 in recrystallized (at 973â¯K for 4â¯h) condition was subjected to uniaxial tensile testing at a constant cross-head velocity at temperatures in the range 293â¯K-1073â¯K and repeated stress relaxation tests at 293â¯K, 573â¯K, and 773â¯K. The minimum in the total elongation was indicative of dynamic strain aging phenomenon in this alloy in the intermediate temperature regime. The yield stress of the alloy was separated into effective and athermal components and the transition from thermally activated dislocation glide to athermal regime took place at around 673â¯K with the athermal stress estimated to be 115â¯MPa. The activation volume was found to be in the range of 40â¯b3 to 160â¯b3. The activation volume values and the data analyses using the solid-solution models in literature indicated dislocation-solute interaction to be a potential deformation mechanism in thermally activated regime. The activation energy calculated at 573â¯K was very close to that found for diffusivity of oxygen in α-Zr that was suggestive of dislocations-oxygen interaction during plastic deformation. This type of information may be helpful in alloy design in selecting different elements to control the deformation behavior of the material and impart desired mechanical properties in those materials for specific applications.
Keywords
Related Topics
Physical Sciences and Engineering
Energy
Nuclear Energy and Engineering
Authors
N. Kumar, A. Alomari, K.L. Murty,